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| {{#Wiki_filter:1Q/2000 Inspection Findings - Clinton Page 1 of 8 Clinton Initiating Events Significance: Aug 21, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow procedures associated with feed water level control system surveillance testing. | | {{#Wiki_filter:}} |
| Human performance and corrective action deficiencies contributed to a Non-Cited Violation of Technical Specification 5.4.1 for failing to follow procedures. This led to the unplanned automatic reactor shutdown on July 24, 2001. The finding was of very low safety significance because no complications occurred during the unplanned automatic reactor shut down and the finding did not increase the likelihood of mitigation equipment being unavailable.
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| Inspection Report# : 2001010(pdf)
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| Significance: Dec 31, 2000 Identified By: NRC Item Type: FIN Finding Operators did not adequately control reactor vessel inventory after a reactor scram which resulted in the motor driven reactor feedwater pump tripping on high reactor vessel water level.
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| During operator response to a reactor scram on December 18, 2000, operators did not adequately control reactor vessel inventory prior to the motor driven reactor feedwater pump tripping on high reactor vessel water level. The inspectors reviewed this issue using the significance determination process for a transient with a loss of feedwater and determined this was of very low safety significance because all other reactor vessel level control systems were operable and functioned as designed.
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| Inspection Report# : 2000020(pdf)
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| Significance: Dec 31, 2000 Identified By: NRC Item Type: NCV NonCited Violation Operators failed to adequately control reactor vessel water level and pressure following the automatic reactor scram which resulted in a second automatic scram.
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| Operators failed to adequately control reactor vessel water level and pressure, while attempting to open the main steam isolation valves following the automatic reactor scram on December 18, 2000. This resulted in an automatic scram signal due to low reactor vessel water level. The inspectors reviewed this issue using the significance determination process for a transient with a loss of feedwater and determined this was of very low safety significance because the event occurred while the reactor was shut down and all control rods were already fully inserted.
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| Inspection Report# : 2000020(pdf)
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| Significance: Dec 31, 2000 Identified By: NRC Item Type: NCV NonCited Violation Operators failed to adequately evaluate an alarming moisture separator drain tank level annunciator that resulted in a turbine trip.
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| During plant restart following refueling outage 7, operators did not adequately evaluate an alarming moisture separator drain tank level annunciator.
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| As a result, high water level in the moisture separator drain tank caused a turbine trip with the reactor at approximately 25% power. The inspectors reviewed this issue using the significance determination process for a transient. Since only the initiating event cornerstone is affected and associated assumptions have no other impact than slightly increasing the likelihood of an uncomplicated reactor trip, the finding is considered to be of very low safety significance.
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| Inspection Report# : 2000020(pdf)
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| Significance: Nov 16, 2000 Identified By: NRC Item Type: NCV NonCited Violation An alternate rod insertion system initiation and a manual reactor scram occurred with the reactor shutdown as a result of an inadequate
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| 1Q/2000 Inspection Findings - Clinton Page 2 of 8 maintenance procedure.
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| During replacement of power supplies for the alternate rod insertion (ARI) system, maintenance personnel failed to fully evaluate the impacts that re-energizing the power supplies had on the ARI initiation logic. While re-energizing the power supplies, the initiation logic sensed an ARI signal (low reactor water level). This caused the vent and drain valves to close and the scram discharge volume to fill with water. Plant operators inserted a manual scram signal before the automatic high scram discharge volume set point was reached. One Non-Cited Violation was identified for having an inadequate maintenance procedure to control this activity. This finding was evaluated using the shutdown significance determination process contained in Appendix G of IMC 0609 and was determined to have very low risk significance because it did not impact any of the five shutdown safety functions identified by NUMARC 91-06.
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| Inspection Report# : 2000017(pdf)
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| Significance: Sep 30, 2000 Identified By: NRC Item Type: FIN Finding Human performance errors and an inadequate procedure resulted in exceeding the allowed outage time for the emergency reserve auxiliary transformer static VAR compensator.
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| Human performance errors and the failure to develop an adequate procedure for the emergency reserve auxiliary transformer static VAR (Volt Ampere Reactive) compensator (ERAT-SVC) surveillance test resulted in several delays during the test. These delays caused the work to not be completed within the allowed outage time. Therefore, a request for Enforcement Discretion was presented to the NRC which was formally granted on September 20, 2000 (NOED 00-6-011). The safety significance of this finding was very low because all other emergency core cooling system trains (automatic depressurization system, low pressure core spray, and low pressure core injection), emergency diesel generators, and the reactor core isolation cooling system were operable.
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| Inspection Report# : 2000015(pdf)
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| Significance: May 17, 2000 Identified By: Self Disclosing Item Type: FIN Finding Manual reactor shutdown A labeling discrepancy contributed to the improper isolation of a protective relay for the 4.16kV Bus 1B Reserve Feed Breaker. As a result, during functional testing, the relay actuated and caused the bus to be de-energized which ultimately resulted in a manual reactor shut down. This issue was determined to be of very low risk significance due to remaining mitigation capability and recovery potential.
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| Inspection Report# : 2000008(pdf)
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| Mitigating Systems Significance: N/A Dec 29, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW 10 CFR 55.59(c)(5) REQUIREMENTS FOR RETAINING LICENSED OPERATOR REQUALIFICATION PROGRAM RECORDS The inspectors identified a Non-Cited Violation wherein the facility licensee had failed to follow the Code of Federal Regulations (CFR) Title 10, Part 55.59(c)(5), Records, requirements by failing to systematically retain all of the original or authenticated copies of the original evaluation documents during the year 2000 annual NRC examination. The finding was of very low safety significance because although the records were not the original or authenticated copies of the original, records did exist in computerized clerically transcribed documents. The computer records had not been signed, and there was no indication that they had been verified correct by the original authors. The unauthenticated documents did provide information that licensed operators, for the most part, had participated and were evaluated during the year 2000 NRC annual requalification examination. However, the inspectors determined that the finding was more than minor. Specifically, the inspectors identified at least one instance in which the transcribed information appeared to be incorrect or missing. The records failure had credible impact on safety, in that, it negatively impacted on the intent of the licensed operator requalification examination process which, in part, is to maintain a high level of confidence that licensed operators continue to possess the requisite knowledge and abilities needed to safely perform licensed duties. In addition, inadequate records keeping adversely affects the NRC's ability to regulate.
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| Inspection Report# : 2001015(pdf)
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| Significance: Jan 26, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct longstanding Reactor Core Isolation Cooling (RCIC) System valve degradation
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| 1Q/2000 Inspection Findings - Clinton Page 3 of 8 Corrective actions for a longstanding deficiency with the Reactor Core Isolation Cooling (RCIC) system steam bypass valve were not effective in stopping the leakage past the valve. This finding was determined to be a Non-Cited Violation of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action." This finding was determined to have very low risk significance because the degraded condition of the valve did not affect the operability of the RCIC system.
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| Inspection Report# : 2001002(pdf)
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| Significance: Jan 26, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow Condition Report process for Shutdown Service Water (SX) pipe wall thinning Corrective actions were not implemented to replace a portion of the shutdown service water (SX) system piping after pipe wall thinning was identified. The failure to take the specified corrective actions by the committed due date or to properly reevaluate the degraded condition was determined to be a Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, "Procedures." This finding was determined to have very low safety significance because the SX system remained operable and capable of performing its' safety function.
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| Inspection Report# : 2001002(pdf)
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| Significance: Sep 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation Procedural requirements were not followed when unexpected equipment response was encountered.
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| Maintenance personnel failed to appropriately follow procedure instructions during testing of the Division III emergency diesel generator room fire detection system. These actions led to the emergency diesel generator being rendered inoperable. The procedure violation was treated as a Non-Cited Violation. This issue was of very low safety significance since the other divisional emergency diesel generators and all emergency core cooling systems were operable at the time of discovery.
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| Inspection Report# : 2000015(pdf)
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| Significance: Jun 16, 2000 Identified By: NRC Item Type: NCV NonCited Violation The licensee failed to ensure that appropriate post-modification testing was specified and accomplished for the Division I and Division III EDG output breaker circuitry modifications The licensee failed to ensure that the appropriate post-modification testing (PMT) was specified in the Division I and Division III emergency diesel generator (EDG) output breaker circuitry modification packages and that the post-modification tests were correctly accomplished. This was required to demonstrate through component and functional testing that the modified (rewired) portions of the Division I and Division III EDG output breaker circuitry were adequately installed to accomplish the intent of the plant design changes.
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| Inspection Report# : 2000012(pdf)
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| Barrier Integrity Significance: Nov 16, 2000 Identified By: NRC Item Type: NCV NonCited Violation Secondary containment was inoperable for 6 minutes during fuel movements when interlock doors were opened.
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| Secondary containment was inoperable for 6 minutes during fuel movements when secondary containment interlock doors were inadvertently opened to move scaffolding. The inoperability was discovered when operators in the control room received an alarm indicating a loss of secondary containment vacuum. One Non-Cited Violation was identified for violating Technical Specification 3.6.4.1 which requires secondary containment operability during fuel moves. This finding was evaluated using the shutdown significance determination process contained in Appendix G of IMC 0609 and was determined to have very low risk significance because it did not meet the criteria for findings requiring a phase 2 significance evaluation.
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| Inspection Report# : 2000017(pdf)
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| Significance: Nov 14, 2000
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| 1Q/2000 Inspection Findings - Clinton Page 4 of 8 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform radiographic examinations of Class 2 welds.
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| The inspectors identified a Non-Cited Violation for the failure to perform radiographic examinations of Class 2 welds in compliance with applicable American Society of Mechanical Engineers (ASME) Code requirements. During installation testing of the 1999 Feedwater Keep Fill FW-39 modification, five radiographic examinations had recorded geometric unsharpness values which exceeded Section III and Section V ASME Code limits. Radiographic geometric unsharpness values are used to ensure that the film is of adequate quality to see defects. In addition, inspectors identified that three examinations did not meet Section V Code requirements for documentation of radiographic technique variables which can affect the image quality of the film. The safety significance of this issue was considered very low at this time, based on the absence of adverse consequences, the presence of other image quality indicators, and because the issue did not involve the system isolation valves. The failure to comply with ASME Code radiographic examination requirements could result in the failure to detect flaws within reactor coolant boundary piping, and was considered a Non-Cited Violation of 10 CFR Part 50.55a, "Codes and Standards".
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| Inspection Report# : 2000019(pdf)
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| Emergency Preparedness Significance: Aug 21, 2001 Identified By: NRC Item Type: NCV NonCited Violation Violation of 10 CFR 50.54(q) re. SCBA qualifications A Non-Cited Violation of 10 CFR 50.54(q) was identified by the NRC associated with the failure to maintain personnel qualifications for self contained breathing apparatus in accordance with the Clinton Power Station Emergency Plan. The finding was of very low safety significance because the licensee maintained an adequate number of qualified personnel to maintain minimum coverage of the required positions identified in the Emergency Plan.
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| Inspection Report# : 2001010(pdf)
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| Significance: Jun 08, 2001 Identified By: NRC Item Type: VIO Violation Supplemental Inspection -- Failure to correct self-identified defficiencies disclosed through control room communications drills This supplemental inspection was performed by the NRC to assess the licensee's evaluation associated with inaccuracies in the reporting of the Drill and Exercise Performance (DEP) performance indicator and with the performance deficiencies that resulted in a White DEP performance indicator (fourth quarter 1999 through the fourth quarter 2000). During the inspection, performed in accordance with NRC Inspection Procedure 95001, the inspector concluded that the licensee performed an adequate evaluation to determine the causes of both issues. In the case of the performance indicator errors, the licensee performed a root cause evaluation which identified a personnel error that was compounded by the lack of self-checking and verification. In addition, the licensee identified contributing causes that included the failure to provide adequate training to the emergency preparedness staff and the failure to provide adequate procedural guidance to the performance indicator data stewards and verifiers, which also applied to performance indicators in other cornerstones. The inspector concluded that the scope of corrective actions planned and implemented by the licensee appeared to address the identified causes. However, the inspector observed an additional discrepancy in the recently completed performance indicator evaluation related to drill and exercise participation. In addition, the licensee identified an error in its evaluation of one of the other emergency preparedness performance indicators that was not detected during its evaluation. These observations demonstrated weaknesses in the licensee's corrective actions and extent of condition review. The errors in the licensee's reporting of the DEP performance indicator was significant, in that the error resulted in a change of color, (i.e., Green-to-White). Consequently, a violation of 10 CFR 50.9 of more than minor safety significance was identified. Since the inaccurate reporting occurred during the period that the NRC's Enforcement Policy afforded discretion for the non-willful submittal of inaccurate performance indicator information, the NRC is exercising enforcement discretion and not citing the violation. In the case of the White DEP performance indicator, the inspector concluded that the licensee adequately assessed the deficiencies that led to the performance issues. Based on its review, the licensee attributed the White performance indicator to the high failure rate of control room communicator drills (i.e., job performance measures). The licensee identified two apparent causes for the high failure rate: (1) weaknesses in formal training; and (2) failure to meet emergency preparedness management expectations concerning the identification and correction of drill deficiencies. The inspector reviewed the licensee's corrective actions and determined that they addressed the causes identified. As a result of the licensee's immediate corrective actions, the licensee's performance returned the performance indicator to the Green band. The inspector and the licensee concluded that the high failure rate of the control room communicators resulted, in part, from inadequate corrective actions for self-identified deficiencies. Specifically, the licensee control room communicator drills were a portion of an overall annual evaluation of non-licensed operators, which included non-emergency preparedness functions. Generally, the failure of the communications segment of the evaluation did not result in a total failure of the annual evaluation. Therefore, the licensee's remedial actions were limited and were not effective in correcting the deficiencies and preventing similar failures from occurring, as required by 10 CFR 50.47(b)(14). By letter dated 08/22/01, the NRC concluded that a violation of 10 CFR 50.47(b)(14) had occurred and using the NRC's significance determination process, determined that the finding was white.
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| Inspection Report# : 2001009(pdf)
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| 1Q/2000 Inspection Findings - Clinton Page 5 of 8 Significance: Feb 23, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to follow emergency plan for on-shift staffing For an approximate 2-month time period, the licensee failed to meet one of the minimum on-shift emergency response organization (ERO) staffing requirements contained in Table 2-1 of the licensee's emergency plan.
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| Inspection Report# : 2001003(pdf)
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| Significance: N/A Apr 28, 2000 Identified By: NRC Item Type: FIN Finding Emergency Preparedness Performance Indicator Verification Alert and Notification System, Drill & Exercise Participation, and Drill & Exercise performance indicators: The inspectors verified that the licensee had acceptably gathered information and reported these three performance indicators, which were in the green band, with the following minor exception. The inspectors identified a discrepancy with the licensee's initial assessment of the Drill and Exercise Performance (DEP) indicator related to the number of performance opportunities associated with a General Emergency declaration during a drill or an exercise. The licensee initially assumed that only three performance opportunities would exist rather than four as provided in NEI 99-02, but later recognized that they had misinterpreted the guidance. This did not affect the DEP performance indicator which was in the green band.
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| Inspection Report# : 2000009(pdf)
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| Occupational Radiation Safety Significance: Oct 08, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an Adequate Survey to identify and to post a High Radiation Area A finding and associated Non-Cited Violation was identified concerning the failure to perform an adequate radiological survey, as required by 10 CFR 20.1501. Although the licensee identified this issue, the licensee did not thoroughly evaluate the cause(s) of the unanticipated radiological conditions and associated problems in the monitoring of radioactive waste activities, which have resulted in previous, similar incidents. The finding was of very low safety significance because the area radiation levels and the licensee's additional administrative barriers would have limited the potential for an individual inadvertently entering the area and receiving a radiation exposure in excess of regulatory limits.
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| Inspection Report# : 2001015(pdf)
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| Significance: Aug 21, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to maintain administrative control of high radiation area keys as required by Technical Specification 5.7.2 Technical Specification 5.7.2 requires, in part, that doors to high radiation areas in which an individual could receive a deep dose equivalent greater than or equal to 1000 millirem in one hour (at 30 centimeters) shall be provided with locked or continuously guarded doors to prevent unauthorized entry and that the keys to such doors shall be administratively controlled. During October 29 - 31, 2001, the licensee failed to maintain administrative control of a key that controlled five access points to high radiation areas specified above (i.e., lost the key and failed to perform required key inventories to identify its loss), as described in CR No. 2-00-11-016. Since the inspector concluded that sufficient barriers remained to prevent an unauthorized individual from entering the affected areas and receiving an overexposure, the inspector concluded that the incident was of very low safety significance. The licensee also reported the incident to the NRC as an occurrence for the Occupational Exposure Control Effectiveness performance indicator. This is being treated as a Non-Cited Violation.
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| Inspection Report# : 2001010(pdf)
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| Significance: SL-IV Jul 27, 2001 Identified By: NRC Item Type: NCV NonCited Violation Misuse of Radioactive Material to Alarm a PCM Radiation protection technician used contaminated material to alarm a portal contamination monitor (PCM), while an individual was performing a contamination survey. Based on the licensee's investigation, the contamination was not placed on the individual, and the individual successfully monitored through an additional PCM. This incident will be reviewed by the NRC for potential enforcement actions. Update: On July 27, 2001, the NRC identified and forwarded to the licensee (by letter) a Non-Cited Violation of the Clinton Station Facility Operating License associated with the deliberate misuse of radioactive material by a junior contract radiation protection technician. On October 20, 2000, the technician misused
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| 1Q/2000 Inspection Findings - Clinton Page 6 of 8 radioactive material to cause an erroneous alarm on a PCM, as another individual was performing a contamination survey. The licensee identified the incident, entered the incident into its corrective action program, and implemented immediate corrective actions. Since the violation was determined to be willful, the NRC did not assign a significance to the violation using the NRC's Significance Determination Process. In accordance with the NRC Enforcement Policy, the NRC determined that the incident constituted a Severity Level IV violation of the Clinton Power Station Facility Operating License. Further, the NRC determined that the violation met the criteria necessary to disposition the violation as a Non-Cited Violation (Section VI.A.1.d of the NRC Enforcement Policy).
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| Inspection Report# : 2001010(pdf)
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| Inspection Report# : 2000018(pdf)
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| Significance: Oct 25, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Three individuals entered a HRA in violation of Technical Specification 5.7.1 On October 25, 2000, three individuals entered the B residual heat removal heat exchanger room (a posted high radiation area); however, the individuals were not working under a radiation work permit that allowed entry into the high radiation area and did not satisfy either of the three entry conditions of Technical Specification 5.7.1.
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| Inspection Report# : 2000018(pdf)
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| Public Radiation Safety Significance: Dec 08, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Inadvertent Release of Radioactive Material to Unrestricted Area During September 2000, the licensee conducted a survey of tools, equipment, etc. outside of the restricted area (protected area and owner controlled area) and identified low-level contaminated materials that were not under constant surveillance or control. The failure to maintain contstant surveillance and control of the material was a violation of 10 CFR 20.1802 and was characterized as a Non-Cited Violation. Based on the licensee's conservative annual dose assessment (about 1.56 millirem) and the inability to define the origin of each of the items, the inspector concluded that the issue constituted one occurrence/event per the NRC Significance Determination Process (Green).
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| Inspection Report# : 2000021(pdf)
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| Physical Protection Miscellaneous Significance: Feb 17, 2002 Identified By: NRC Item Type: NCV NonCited Violation Non-cited violation of T.S. 5.4.1 for an inadequate surveillance procedure.
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| Inspection Report# : 2001016(pdf)
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| Significance: Feb 17, 2002 Identified By: NRC Item Type: FIN Finding A temporary modification on the "A" RR FCV control cirucuitry.
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| On December 14, 2001, the licensee installed a temporary modification on the "A" RR FCV control circuitry. The T-mod was installed to assist the
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| 1Q/2000 Inspection Findings - Clinton Page 7 of 8 operators in manually controlling the "A" RR FCV because the reliability of the normal control circuitry was in question. During the implementation portion of the T-mod installations, the "A" RR FCV unexpectedly moved from 94 percent open to 102 percent open at which point the protective position circuitry locked the valve at the 102 percent position. Recator power was observed to go from 94 percent to 98 percent during this unexpected valve movement. Following this unexpected FCV movement, opeerations personnel ordered the T-mod to be removed and operators then proceeded to manually shut down the reactor without any further movements of the "A" RR FCV.
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| Inspection Report# : 2001016(pdf)
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| Significance: SL-IV Aug 18, 2001 Identified By: NRC Item Type: VIO Violation Falsification of Test Records by Licensee Employee SL IV - On July 2, 2001, by separate letter, NRC issued a Severity Level IV violation of 10 CFR 50.9 for a deliberate falsification by a plant test engineer. Following investigation by the Office of Investigations, NRC determined that, on October 20, 2000, a test engineer forged another employee's signature on two test package cover sheets on by forging another employee's signature without his prior concurrence, in violation of Clinton established plant protocol and procedure.
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| Inspection Report# : 2001010(pdf)
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| Significance: SL-IV Apr 06, 2001 Identified By: NRC Item Type: VIO Violation Violation of 10 CFR 50.7 "Employee Protection" On April 6, 2001, the NRC issued the licensee a Severity Level IV Violation of 10 CFR 50.7. The NRC concluded that the licensee took adverse employment actions against an employee in the licensee's Nuclear Training Department (i.e., unfavorable 1999 performance review), in part, as a result of the employee's engagement in protected activities. In addition, the NRC learned that several training personnel may be reluctant to discuss department issues within the nuclear training department.
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| Inspection Report# : 2001010(pdf)
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| Inspection Report# : 2001007(pdf)
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| Significance: N/A Jan 26, 2001 Identified By: NRC Item Type: FIN Finding Assessment of Problem Identification and Resolution Performance The team identified that the licensee appropriately entered significant plant issues into the corrective action process by initiating condition reports.
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| Some less significant conditions adverse to quality were evaluated and corrected outside the established process. The trending program was not fully effective as a problem identification tool. Quality Assurance audits and self-assessments reviewed varied in quality. Identified issues were generally evaluated properly, although in several cases the corrective action process did not work effectively to either evaluate or prioritize issues.
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| Current station performance issues including human performance, corrective action program, surveillance testing, and labeling indicate that long term corrective actions previously taken in these areas as restart and post-restart initiatives have not been fully effective to support sustained improvement. Corrective actions were not always fully effective or timely for some individual equipment issues and the effectiveness review process (CARE) did not always identify ineffective corrective actions. The licensee had recently recognized similar deficiencies in corrective action program implementation but had not yet fully developed or completed the corrective actions to improve these areas. The inspectors did not find any reluctance by the station employees to raise safety issues.
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| Inspection Report# : 2001002(pdf)
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| Significance: N/A Dec 31, 2000 Identified By: NRC Item Type: FIN Finding Recent human performance issues have occurred which are associated with operator performance and knowledge based deficiencies that have affected plant operations and responses to transient conditions.
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| Recent human performance issues have occurred which are associated with operator performance and knowledge based deficiencies that have affected plant operations and responses to transient conditions. While the risk of the individual events was very low, the failure of operators to adequately control level parameters indicated a declining trend in this area. These issues could not be easily evaluated by present risk analysis methods because failures to follow procedures and maintaining management expectations were not modeled in the Clinton Individual Plant Evaluation. Therefore, the finding is characterized as having no color.
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| Inspection Report# : 2000020(pdf)
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| Significance: N/A Nov 14, 2000 Identified By: NRC Item Type: FIN Finding Three procedures were not written in compliance with the applicable ASME Code.
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| The inspectors reviewed three special process procedures, and identified areas where all three procedures were not written in compliance with the applicable ASME Code. The procedure deficiencies had the potential to affect the ASME Code compliance of weld fabrication and nondestructive examination used on safety-related components and piping. The inspectors noted that each of the ASME Code problems identified contained elements of human performance deficiencies. The human performance aspects, while not always being the root cause of the problem, were significant contributors leading to procedure deficiencies. While the risk of the individual examples was very low, the number of deficiencies
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| 1Q/2000 Inspection Findings - Clinton Page 8 of 8 indicated a problem with incorporation of applicable ASME Code requirements into special process procedures.
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| Inspection Report# : 2000019(pdf)
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| Significance: N/A Sep 30, 2000 Identified By: NRC Item Type: FIN Finding Recent events affecting plant operations contained elements of human performance deficiencies.
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| NO COLOR. The inspectors noted that several recent events which have affected plant operations and the operability of safety-related components or other components important to safety contained elements of human performance deficiencies. The human performance aspects, while not always being the root cause of the problem, were significant contributors leading to the events. While the risk of the individual events was very low, the number of maintenance-related incidents indicated a problem exists with the control, review, and performance of maintenance activities.
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| Inspection Report# : 2000015(pdf)
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| Significance: N/A May 20, 2000 Identified By: NRC Item Type: FIN Finding Inaccurate historical data for the Safety System Functional Failure Indicator No Color. The licensee identified a failure to submit accurate information to the NRC. The inaccurate information involved the historical data submittal for the Safety System Functional Failure Performance Indicator. The error resulted in a response band color change from Green to White for the first quarter 1999 Performance Indicator. The NRC exercised Enforcement Discretion pursuant to Section VII.B.6 of the Enforcement Policy and did not cite the violation.
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| Inspection Report# : 2000008(pdf)
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| Last modified : April 01, 2002
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| 2Q/2000 Inspection Findings - Clinton Page 1 of 8 Clinton Initiating Events Significance: May 17, 2000 Identified By: Self Disclosing Item Type: FIN Finding Manual reactor shutdown A labeling discrepancy contributed to the improper isolation of a protective relay for the 4.16kV Bus 1B Reserve Feed Breaker. As a result, during functional testing, the relay actuated and caused the bus to be de-energized which ultimately resulted in a manual reactor shut down. This issue was determined to be of very low risk significance due to remaining mitigation capability and recovery potential.
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| Inspection Report# : 2000008(pdf)
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| Significance: Aug 21, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow procedures associated with feed water level control system surveillance testing.
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| Human performance and corrective action deficiencies contributed to a Non-Cited Violation of Technical Specification 5.4.1 for failing to follow procedures. This led to the unplanned automatic reactor shutdown on July 24, 2001. The finding was of very low safety significance because no complications occurred during the unplanned automatic reactor shut down and the finding did not increase the likelihood of mitigation equipment being unavailable.
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| Inspection Report# : 2001010(pdf)
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| Significance: Dec 31, 2000 Identified By: NRC Item Type: NCV NonCited Violation Operators failed to adequately evaluate an alarming moisture separator drain tank level annunciator that resulted in a turbine trip.
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| During plant restart following refueling outage 7, operators did not adequately evaluate an alarming moisture separator drain tank level annunciator.
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| As a result, high water level in the moisture separator drain tank caused a turbine trip with the reactor at approximately 25% power. The inspectors reviewed this issue using the significance determination process for a transient. Since only the initiating event cornerstone is affected and associated assumptions have no other impact than slightly increasing the likelihood of an uncomplicated reactor trip, the finding is considered to be of very low safety significance.
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| Inspection Report# : 2000020(pdf)
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| Significance: Dec 31, 2000 Identified By: NRC Item Type: NCV NonCited Violation Operators failed to adequately control reactor vessel water level and pressure following the automatic reactor scram which resulted in a second automatic scram.
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| Operators failed to adequately control reactor vessel water level and pressure, while attempting to open the main steam isolation valves following the automatic reactor scram on December 18, 2000. This resulted in an automatic scram signal due to low reactor vessel water level. The inspectors reviewed this issue using the significance determination process for a transient with a loss of feedwater and determined this was of very low safety significance because the event occurred while the reactor was shut down and all control rods were already fully inserted.
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| Inspection Report# : 2000020(pdf)
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| Significance: Dec 31, 2000 Identified By: NRC Item Type: FIN Finding Operators did not adequately control reactor vessel inventory after a reactor scram which resulted in the motor driven reactor feedwater pump tripping on high reactor vessel water level.
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| During operator response to a reactor scram on December 18, 2000, operators did not adequately control reactor vessel inventory prior to the
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| 2Q/2000 Inspection Findings - Clinton Page 2 of 8 motor driven reactor feedwater pump tripping on high reactor vessel water level. The inspectors reviewed this issue using the significance determination process for a transient with a loss of feedwater and determined this was of very low safety significance because all other reactor vessel level control systems were operable and functioned as designed.
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| Inspection Report# : 2000020(pdf)
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| Significance: Nov 16, 2000 Identified By: NRC Item Type: NCV NonCited Violation An alternate rod insertion system initiation and a manual reactor scram occurred with the reactor shutdown as a result of an inadequate maintenance procedure.
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| During replacement of power supplies for the alternate rod insertion (ARI) system, maintenance personnel failed to fully evaluate the impacts that re-energizing the power supplies had on the ARI initiation logic. While re-energizing the power supplies, the initiation logic sensed an ARI signal (low reactor water level). This caused the vent and drain valves to close and the scram discharge volume to fill with water. Plant operators inserted a manual scram signal before the automatic high scram discharge volume set point was reached. One Non-Cited Violation was identified for having an inadequate maintenance procedure to control this activity. This finding was evaluated using the shutdown significance determination process contained in Appendix G of IMC 0609 and was determined to have very low risk significance because it did not impact any of the five shutdown safety functions identified by NUMARC 91-06.
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| Inspection Report# : 2000017(pdf)
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| Significance: Sep 30, 2000 Identified By: NRC Item Type: FIN Finding Human performance errors and an inadequate procedure resulted in exceeding the allowed outage time for the emergency reserve auxiliary transformer static VAR compensator.
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| Human performance errors and the failure to develop an adequate procedure for the emergency reserve auxiliary transformer static VAR (Volt Ampere Reactive) compensator (ERAT-SVC) surveillance test resulted in several delays during the test. These delays caused the work to not be completed within the allowed outage time. Therefore, a request for Enforcement Discretion was presented to the NRC which was formally granted on September 20, 2000 (NOED 00-6-011). The safety significance of this finding was very low because all other emergency core cooling system trains (automatic depressurization system, low pressure core spray, and low pressure core injection), emergency diesel generators, and the reactor core isolation cooling system were operable.
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| Inspection Report# : 2000015(pdf)
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| Mitigating Systems Significance: Jun 16, 2000 Identified By: NRC Item Type: NCV NonCited Violation The licensee failed to ensure that appropriate post-modification testing was specified and accomplished for the Division I and Division III EDG output breaker circuitry modifications The licensee failed to ensure that the appropriate post-modification testing (PMT) was specified in the Division I and Division III emergency diesel generator (EDG) output breaker circuitry modification packages and that the post-modification tests were correctly accomplished. This was required to demonstrate through component and functional testing that the modified (rewired) portions of the Division I and Division III EDG output breaker circuitry were adequately installed to accomplish the intent of the plant design changes.
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| Inspection Report# : 2000012(pdf)
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| Significance: N/A Dec 29, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW 10 CFR 55.59(c)(5) REQUIREMENTS FOR RETAINING LICENSED OPERATOR REQUALIFICATION PROGRAM RECORDS The inspectors identified a Non-Cited Violation wherein the facility licensee had failed to follow the Code of Federal Regulations (CFR) Title 10, Part 55.59(c)(5), Records, requirements by failing to systematically retain all of the original or authenticated copies of the original evaluation documents during the year 2000 annual NRC examination. The finding was of very low safety significance because although the records were not the original or authenticated copies of the original, records did exist in computerized clerically transcribed documents. The computer records had not been signed, and there was no indication that they had been verified correct by the original authors. The unauthenticated documents did provide information that licensed operators, for the most part, had participated and were evaluated during the year 2000 NRC annual requalification
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| 2Q/2000 Inspection Findings - Clinton Page 3 of 8 examination. However, the inspectors determined that the finding was more than minor. Specifically, the inspectors identified at least one instance in which the transcribed information appeared to be incorrect or missing. The records failure had credible impact on safety, in that, it negatively impacted on the intent of the licensed operator requalification examination process which, in part, is to maintain a high level of confidence that licensed operators continue to possess the requisite knowledge and abilities needed to safely perform licensed duties. In addition, inadequate records keeping adversely affects the NRC's ability to regulate.
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| Inspection Report# : 2001015(pdf)
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| Significance: Jan 26, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow Condition Report process for Shutdown Service Water (SX) pipe wall thinning Corrective actions were not implemented to replace a portion of the shutdown service water (SX) system piping after pipe wall thinning was identified. The failure to take the specified corrective actions by the committed due date or to properly reevaluate the degraded condition was determined to be a Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, "Procedures." This finding was determined to have very low safety significance because the SX system remained operable and capable of performing its' safety function.
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| Inspection Report# : 2001002(pdf)
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| Significance: Jan 26, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct longstanding Reactor Core Isolation Cooling (RCIC) System valve degradation Corrective actions for a longstanding deficiency with the Reactor Core Isolation Cooling (RCIC) system steam bypass valve were not effective in stopping the leakage past the valve. This finding was determined to be a Non-Cited Violation of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action." This finding was determined to have very low risk significance because the degraded condition of the valve did not affect the operability of the RCIC system.
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| Inspection Report# : 2001002(pdf)
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| Significance: Sep 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation Procedural requirements were not followed when unexpected equipment response was encountered.
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| Maintenance personnel failed to appropriately follow procedure instructions during testing of the Division III emergency diesel generator room fire detection system. These actions led to the emergency diesel generator being rendered inoperable. The procedure violation was treated as a Non-Cited Violation. This issue was of very low safety significance since the other divisional emergency diesel generators and all emergency core cooling systems were operable at the time of discovery.
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| Inspection Report# : 2000015(pdf)
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| Barrier Integrity Significance: Nov 16, 2000 Identified By: NRC Item Type: NCV NonCited Violation Secondary containment was inoperable for 6 minutes during fuel movements when interlock doors were opened.
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| Secondary containment was inoperable for 6 minutes during fuel movements when secondary containment interlock doors were inadvertently opened to move scaffolding. The inoperability was discovered when operators in the control room received an alarm indicating a loss of secondary containment vacuum. One Non-Cited Violation was identified for violating Technical Specification 3.6.4.1 which requires secondary containment operability during fuel moves. This finding was evaluated using the shutdown significance determination process contained in Appendix G of IMC 0609 and was determined to have very low risk significance because it did not meet the criteria for findings requiring a phase 2 significance evaluation.
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| Inspection Report# : 2000017(pdf)
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| Significance: Nov 14, 2000
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| 2Q/2000 Inspection Findings - Clinton Page 4 of 8 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform radiographic examinations of Class 2 welds.
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| The inspectors identified a Non-Cited Violation for the failure to perform radiographic examinations of Class 2 welds in compliance with applicable American Society of Mechanical Engineers (ASME) Code requirements. During installation testing of the 1999 Feedwater Keep Fill FW-39 modification, five radiographic examinations had recorded geometric unsharpness values which exceeded Section III and Section V ASME Code limits. Radiographic geometric unsharpness values are used to ensure that the film is of adequate quality to see defects. In addition, inspectors identified that three examinations did not meet Section V Code requirements for documentation of radiographic technique variables which can affect the image quality of the film. The safety significance of this issue was considered very low at this time, based on the absence of adverse consequences, the presence of other image quality indicators, and because the issue did not involve the system isolation valves. The failure to comply with ASME Code radiographic examination requirements could result in the failure to detect flaws within reactor coolant boundary piping, and was considered a Non-Cited Violation of 10 CFR Part 50.55a, "Codes and Standards".
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| Inspection Report# : 2000019(pdf)
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| Emergency Preparedness Significance: N/A Apr 28, 2000 Identified By: NRC Item Type: FIN Finding Emergency Preparedness Performance Indicator Verification Alert and Notification System, Drill & Exercise Participation, and Drill & Exercise performance indicators: The inspectors verified that the licensee had acceptably gathered information and reported these three performance indicators, which were in the green band, with the following minor exception. The inspectors identified a discrepancy with the licensee's initial assessment of the Drill and Exercise Performance (DEP) indicator related to the number of performance opportunities associated with a General Emergency declaration during a drill or an exercise. The licensee initially assumed that only three performance opportunities would exist rather than four as provided in NEI 99-02, but later recognized that they had misinterpreted the guidance. This did not affect the DEP performance indicator which was in the green band.
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| Inspection Report# : 2000009(pdf)
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| Significance: Aug 21, 2001 Identified By: NRC Item Type: NCV NonCited Violation Violation of 10 CFR 50.54(q) re. SCBA qualifications A Non-Cited Violation of 10 CFR 50.54(q) was identified by the NRC associated with the failure to maintain personnel qualifications for self contained breathing apparatus in accordance with the Clinton Power Station Emergency Plan. The finding was of very low safety significance because the licensee maintained an adequate number of qualified personnel to maintain minimum coverage of the required positions identified in the Emergency Plan.
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| Inspection Report# : 2001010(pdf)
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| Significance: Jun 08, 2001 Identified By: NRC Item Type: VIO Violation Supplemental Inspection -- Failure to correct self-identified defficiencies disclosed through control room communications drills This supplemental inspection was performed by the NRC to assess the licensee's evaluation associated with inaccuracies in the reporting of the Drill and Exercise Performance (DEP) performance indicator and with the performance deficiencies that resulted in a White DEP performance indicator (fourth quarter 1999 through the fourth quarter 2000). During the inspection, performed in accordance with NRC Inspection Procedure 95001, the inspector concluded that the licensee performed an adequate evaluation to determine the causes of both issues. In the case of the performance indicator errors, the licensee performed a root cause evaluation which identified a personnel error that was compounded by the lack of self-checking and verification. In addition, the licensee identified contributing causes that included the failure to provide adequate training to the emergency preparedness staff and the failure to provide adequate procedural guidance to the performance indicator data stewards and verifiers, which also applied to performance indicators in other cornerstones. The inspector concluded that the scope of corrective actions planned and implemented by the licensee appeared to address the identified causes. However, the inspector observed an additional discrepancy in the recently completed performance indicator evaluation related to drill and exercise participation. In addition, the licensee identified an error in its evaluation of one of the other emergency preparedness performance indicators that was not detected during its evaluation. These observations demonstrated weaknesses in the licensee's corrective actions and extent of condition review. The errors in the licensee's reporting of the DEP performance indicator was significant, in that the error resulted in a change of color, (i.e., Green-to-White). Consequently, a violation of 10 CFR 50.9 of more than minor safety significance was identified. Since the inaccurate reporting occurred during the period that the NRC's Enforcement Policy afforded discretion for the non-willful submittal of inaccurate performance indicator information, the NRC is exercising enforcement discretion and not citing the violation. In the case of the White DEP performance indicator, the inspector concluded that the licensee adequately assessed the deficiencies that led to the performance issues. Based on its review, the licensee attributed the White performance indicator to the high failure rate of control
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| 2Q/2000 Inspection Findings - Clinton Page 5 of 8 room communicator drills (i.e., job performance measures). The licensee identified two apparent causes for the high failure rate: (1) weaknesses in formal training; and (2) failure to meet emergency preparedness management expectations concerning the identification and correction of drill deficiencies. The inspector reviewed the licensee's corrective actions and determined that they addressed the causes identified. As a result of the licensee's immediate corrective actions, the licensee's performance returned the performance indicator to the Green band. The inspector and the licensee concluded that the high failure rate of the control room communicators resulted, in part, from inadequate corrective actions for self-identified deficiencies. Specifically, the licensee control room communicator drills were a portion of an overall annual evaluation of non-licensed operators, which included non-emergency preparedness functions. Generally, the failure of the communications segment of the evaluation did not result in a total failure of the annual evaluation. Therefore, the licensee's remedial actions were limited and were not effective in correcting the deficiencies and preventing similar failures from occurring, as required by 10 CFR 50.47(b)(14). By letter dated 08/22/01, the NRC concluded that a violation of 10 CFR 50.47(b)(14) had occurred and using the NRC's significance determination process, determined that the finding was white.
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| Inspection Report# : 2001009(pdf)
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| Significance: Feb 23, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to follow emergency plan for on-shift staffing For an approximate 2-month time period, the licensee failed to meet one of the minimum on-shift emergency response organization (ERO) staffing requirements contained in Table 2-1 of the licensee's emergency plan.
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| Inspection Report# : 2001003(pdf)
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| Occupational Radiation Safety Significance: Oct 08, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an Adequate Survey to identify and to post a High Radiation Area A finding and associated Non-Cited Violation was identified concerning the failure to perform an adequate radiological survey, as required by 10 CFR 20.1501. Although the licensee identified this issue, the licensee did not thoroughly evaluate the cause(s) of the unanticipated radiological conditions and associated problems in the monitoring of radioactive waste activities, which have resulted in previous, similar incidents. The finding was of very low safety significance because the area radiation levels and the licensee's additional administrative barriers would have limited the potential for an individual inadvertently entering the area and receiving a radiation exposure in excess of regulatory limits.
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| Inspection Report# : 2001015(pdf)
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| Significance: Aug 21, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to maintain administrative control of high radiation area keys as required by Technical Specification 5.7.2 Technical Specification 5.7.2 requires, in part, that doors to high radiation areas in which an individual could receive a deep dose equivalent greater than or equal to 1000 millirem in one hour (at 30 centimeters) shall be provided with locked or continuously guarded doors to prevent unauthorized entry and that the keys to such doors shall be administratively controlled. During October 29 - 31, 2001, the licensee failed to maintain administrative control of a key that controlled five access points to high radiation areas specified above (i.e., lost the key and failed to perform required key inventories to identify its loss), as described in CR No. 2-00-11-016. Since the inspector concluded that sufficient barriers remained to prevent an unauthorized individual from entering the affected areas and receiving an overexposure, the inspector concluded that the incident was of very low safety significance. The licensee also reported the incident to the NRC as an occurrence for the Occupational Exposure Control Effectiveness performance indicator. This is being treated as a Non-Cited Violation.
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| Inspection Report# : 2001010(pdf)
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| Significance: SL-IV Jul 27, 2001 Identified By: NRC Item Type: NCV NonCited Violation Misuse of Radioactive Material to Alarm a PCM Radiation protection technician used contaminated material to alarm a portal contamination monitor (PCM), while an individual was performing a contamination survey. Based on the licensee's investigation, the contamination was not placed on the individual, and the individual successfully monitored through an additional PCM. This incident will be reviewed by the NRC for potential enforcement actions. Update: On July 27, 2001, the NRC identified and forwarded to the licensee (by letter) a Non-Cited Violation of the Clinton Station Facility Operating License associated with the deliberate misuse of radioactive material by a junior contract radiation protection technician. On October 20, 2000, the technician misused
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| 2Q/2000 Inspection Findings - Clinton Page 6 of 8 radioactive material to cause an erroneous alarm on a PCM, as another individual was performing a contamination survey. The licensee identified the incident, entered the incident into its corrective action program, and implemented immediate corrective actions. Since the violation was determined to be willful, the NRC did not assign a significance to the violation using the NRC's Significance Determination Process. In accordance with the NRC Enforcement Policy, the NRC determined that the incident constituted a Severity Level IV violation of the Clinton Power Station Facility Operating License. Further, the NRC determined that the violation met the criteria necessary to disposition the violation as a Non-Cited Violation (Section VI.A.1.d of the NRC Enforcement Policy).
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| Inspection Report# : 2001010(pdf)
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| Inspection Report# : 2000018(pdf)
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| Significance: Oct 25, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Three individuals entered a HRA in violation of Technical Specification 5.7.1 On October 25, 2000, three individuals entered the B residual heat removal heat exchanger room (a posted high radiation area); however, the individuals were not working under a radiation work permit that allowed entry into the high radiation area and did not satisfy either of the three entry conditions of Technical Specification 5.7.1.
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| Inspection Report# : 2000018(pdf)
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| Public Radiation Safety Significance: Dec 08, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Inadvertent Release of Radioactive Material to Unrestricted Area During September 2000, the licensee conducted a survey of tools, equipment, etc. outside of the restricted area (protected area and owner controlled area) and identified low-level contaminated materials that were not under constant surveillance or control. The failure to maintain contstant surveillance and control of the material was a violation of 10 CFR 20.1802 and was characterized as a Non-Cited Violation. Based on the licensee's conservative annual dose assessment (about 1.56 millirem) and the inability to define the origin of each of the items, the inspector concluded that the issue constituted one occurrence/event per the NRC Significance Determination Process (Green).
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| Inspection Report# : 2000021(pdf)
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| Physical Protection Miscellaneous Significance: N/A May 20, 2000 Identified By: NRC Item Type: FIN Finding Inaccurate historical data for the Safety System Functional Failure Indicator No Color. The licensee identified a failure to submit accurate information to the NRC. The inaccurate information involved the historical data submittal for the Safety System Functional Failure Performance Indicator. The error resulted in a response band color change from Green to White for the first quarter 1999 Performance Indicator. The NRC exercised Enforcement Discretion pursuant to Section VII.B.6 of the Enforcement Policy and did not cite the violation.
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| Inspection Report# : 2000008(pdf)
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| Significance: Feb 17, 2002 Identified By: NRC Item Type: FIN Finding A temporary modification on the "A" RR FCV control cirucuitry.
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| 2Q/2000 Inspection Findings - Clinton Page 7 of 8 On December 14, 2001, the licensee installed a temporary modification on the "A" RR FCV control circuitry. The T-mod was installed to assist the operators in manually controlling the "A" RR FCV because the reliability of the normal control circuitry was in question. During the implementation portion of the T-mod installations, the "A" RR FCV unexpectedly moved from 94 percent open to 102 percent open at which point the protective position circuitry locked the valve at the 102 percent position. Recator power was observed to go from 94 percent to 98 percent during this unexpected valve movement. Following this unexpected FCV movement, opeerations personnel ordered the T-mod to be removed and operators then proceeded to manually shut down the reactor without any further movements of the "A" RR FCV.
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| Inspection Report# : 2001016(pdf)
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| Significance: Feb 17, 2002 Identified By: NRC Item Type: NCV NonCited Violation Non-cited violation of T.S. 5.4.1 for an inadequate surveillance procedure.
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| Inspection Report# : 2001016(pdf)
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| Significance: SL-IV Aug 18, 2001 Identified By: NRC Item Type: VIO Violation Falsification of Test Records by Licensee Employee SL IV - On July 2, 2001, by separate letter, NRC issued a Severity Level IV violation of 10 CFR 50.9 for a deliberate falsification by a plant test engineer. Following investigation by the Office of Investigations, NRC determined that, on October 20, 2000, a test engineer forged another employee's signature on two test package cover sheets on by forging another employee's signature without his prior concurrence, in violation of Clinton established plant protocol and procedure.
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| Inspection Report# : 2001010(pdf)
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| Significance: SL-IV Apr 06, 2001 Identified By: NRC Item Type: VIO Violation Violation of 10 CFR 50.7 "Employee Protection" On April 6, 2001, the NRC issued the licensee a Severity Level IV Violation of 10 CFR 50.7. The NRC concluded that the licensee took adverse employment actions against an employee in the licensee's Nuclear Training Department (i.e., unfavorable 1999 performance review), in part, as a result of the employee's engagement in protected activities. In addition, the NRC learned that several training personnel may be reluctant to discuss department issues within the nuclear training department.
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| Inspection Report# : 2001007(pdf)
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| Inspection Report# : 2001010(pdf)
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| Significance: N/A Jan 26, 2001 Identified By: NRC Item Type: FIN Finding Assessment of Problem Identification and Resolution Performance The team identified that the licensee appropriately entered significant plant issues into the corrective action process by initiating condition reports.
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| Some less significant conditions adverse to quality were evaluated and corrected outside the established process. The trending program was not fully effective as a problem identification tool. Quality Assurance audits and self-assessments reviewed varied in quality. Identified issues were generally evaluated properly, although in several cases the corrective action process did not work effectively to either evaluate or prioritize issues.
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| Current station performance issues including human performance, corrective action program, surveillance testing, and labeling indicate that long term corrective actions previously taken in these areas as restart and post-restart initiatives have not been fully effective to support sustained improvement. Corrective actions were not always fully effective or timely for some individual equipment issues and the effectiveness review process (CARE) did not always identify ineffective corrective actions. The licensee had recently recognized similar deficiencies in corrective action program implementation but had not yet fully developed or completed the corrective actions to improve these areas. The inspectors did not find any reluctance by the station employees to raise safety issues.
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| Inspection Report# : 2001002(pdf)
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| Significance: N/A Dec 31, 2000 Identified By: NRC Item Type: FIN Finding Recent human performance issues have occurred which are associated with operator performance and knowledge based deficiencies that have affected plant operations and responses to transient conditions.
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| Recent human performance issues have occurred which are associated with operator performance and knowledge based deficiencies that have affected plant operations and responses to transient conditions. While the risk of the individual events was very low, the failure of operators to adequately control level parameters indicated a declining trend in this area. These issues could not be easily evaluated by present risk analysis methods because failures to follow procedures and maintaining management expectations were not modeled in the Clinton Individual Plant Evaluation. Therefore, the finding is characterized as having no color.
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| Inspection Report# : 2000020(pdf)
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| 2Q/2000 Inspection Findings - Clinton Page 8 of 8 Significance: N/A Nov 14, 2000 Identified By: NRC Item Type: FIN Finding Three procedures were not written in compliance with the applicable ASME Code.
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| The inspectors reviewed three special process procedures, and identified areas where all three procedures were not written in compliance with the applicable ASME Code. The procedure deficiencies had the potential to affect the ASME Code compliance of weld fabrication and nondestructive examination used on safety-related components and piping. The inspectors noted that each of the ASME Code problems identified contained elements of human performance deficiencies. The human performance aspects, while not always being the root cause of the problem, were significant contributors leading to procedure deficiencies. While the risk of the individual examples was very low, the number of deficiencies indicated a problem with incorporation of applicable ASME Code requirements into special process procedures.
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| Inspection Report# : 2000019(pdf)
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| Significance: N/A Sep 30, 2000 Identified By: NRC Item Type: FIN Finding Recent events affecting plant operations contained elements of human performance deficiencies.
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| NO COLOR. The inspectors noted that several recent events which have affected plant operations and the operability of safety-related components or other components important to safety contained elements of human performance deficiencies. The human performance aspects, while not always being the root cause of the problem, were significant contributors leading to the events. While the risk of the individual events was very low, the number of maintenance-related incidents indicated a problem exists with the control, review, and performance of maintenance activities.
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| Inspection Report# : 2000015(pdf)
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| Last modified : April 01, 2002
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| 3Q/2000 Inspection Findings - Clinton Page 1 of 8 Clinton Initiating Events Significance: Sep 30, 2000 Identified By: NRC Item Type: FIN Finding Human performance errors and an inadequate procedure resulted in exceeding the allowed outage time for the emergency reserve auxiliary transformer static VAR compensator.
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| Human performance errors and the failure to develop an adequate procedure for the emergency reserve auxiliary transformer static VAR (Volt Ampere Reactive) compensator (ERAT-SVC) surveillance test resulted in several delays during the test. These delays caused the work to not be completed within the allowed outage time. Therefore, a request for Enforcement Discretion was presented to the NRC which was formally granted on September 20, 2000 (NOED 00-6-011). The safety significance of this finding was very low because all other emergency core cooling system trains (automatic depressurization system, low pressure core spray, and low pressure core injection), emergency diesel generators, and the reactor core isolation cooling system were operable.
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| Inspection Report# : 2000015(pdf)
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| Significance: May 17, 2000 Identified By: Self Disclosing Item Type: FIN Finding Manual reactor shutdown A labeling discrepancy contributed to the improper isolation of a protective relay for the 4.16kV Bus 1B Reserve Feed Breaker. As a result, during functional testing, the relay actuated and caused the bus to be de-energized which ultimately resulted in a manual reactor shut down. This issue was determined to be of very low risk significance due to remaining mitigation capability and recovery potential.
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| Inspection Report# : 2000008(pdf)
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| Significance: Aug 21, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow procedures associated with feed water level control system surveillance testing.
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| Human performance and corrective action deficiencies contributed to a Non-Cited Violation of Technical Specification 5.4.1 for failing to follow procedures. This led to the unplanned automatic reactor shutdown on July 24, 2001. The finding was of very low safety significance because no complications occurred during the unplanned automatic reactor shut down and the finding did not increase the likelihood of mitigation equipment being unavailable.
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| Inspection Report# : 2001010(pdf)
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| Significance: Dec 31, 2000 Identified By: NRC Item Type: FIN Finding Operators did not adequately control reactor vessel inventory after a reactor scram which resulted in the motor driven reactor feedwater pump tripping on high reactor vessel water level.
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| During operator response to a reactor scram on December 18, 2000, operators did not adequately control reactor vessel inventory prior to the motor driven reactor feedwater pump tripping on high reactor vessel water level. The inspectors reviewed this issue using the significance determination process for a transient with a loss of feedwater and determined this was of very low safety significance because all other reactor vessel level control systems were operable and functioned as designed.
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| Inspection Report# : 2000020(pdf)
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| Significance: Dec 31, 2000 Identified By: NRC Item Type: NCV NonCited Violation Operators failed to adequately control reactor vessel water level and pressure following the automatic reactor scram which resulted in a
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| 3Q/2000 Inspection Findings - Clinton Page 2 of 8 second automatic scram.
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| Operators failed to adequately control reactor vessel water level and pressure, while attempting to open the main steam isolation valves following the automatic reactor scram on December 18, 2000. This resulted in an automatic scram signal due to low reactor vessel water level. The inspectors reviewed this issue using the significance determination process for a transient with a loss of feedwater and determined this was of very low safety significance because the event occurred while the reactor was shut down and all control rods were already fully inserted.
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| Inspection Report# : 2000020(pdf)
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| Significance: Dec 31, 2000 Identified By: NRC Item Type: NCV NonCited Violation Operators failed to adequately evaluate an alarming moisture separator drain tank level annunciator that resulted in a turbine trip.
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| During plant restart following refueling outage 7, operators did not adequately evaluate an alarming moisture separator drain tank level annunciator.
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| As a result, high water level in the moisture separator drain tank caused a turbine trip with the reactor at approximately 25% power. The inspectors reviewed this issue using the significance determination process for a transient. Since only the initiating event cornerstone is affected and associated assumptions have no other impact than slightly increasing the likelihood of an uncomplicated reactor trip, the finding is considered to be of very low safety significance.
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| Inspection Report# : 2000020(pdf)
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| Significance: Nov 16, 2000 Identified By: NRC Item Type: NCV NonCited Violation An alternate rod insertion system initiation and a manual reactor scram occurred with the reactor shutdown as a result of an inadequate maintenance procedure.
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| During replacement of power supplies for the alternate rod insertion (ARI) system, maintenance personnel failed to fully evaluate the impacts that re-energizing the power supplies had on the ARI initiation logic. While re-energizing the power supplies, the initiation logic sensed an ARI signal (low reactor water level). This caused the vent and drain valves to close and the scram discharge volume to fill with water. Plant operators inserted a manual scram signal before the automatic high scram discharge volume set point was reached. One Non-Cited Violation was identified for having an inadequate maintenance procedure to control this activity. This finding was evaluated using the shutdown significance determination process contained in Appendix G of IMC 0609 and was determined to have very low risk significance because it did not impact any of the five shutdown safety functions identified by NUMARC 91-06.
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| Inspection Report# : 2000017(pdf)
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| Mitigating Systems Significance: Sep 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation Procedural requirements were not followed when unexpected equipment response was encountered.
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| Maintenance personnel failed to appropriately follow procedure instructions during testing of the Division III emergency diesel generator room fire detection system. These actions led to the emergency diesel generator being rendered inoperable. The procedure violation was treated as a Non-Cited Violation. This issue was of very low safety significance since the other divisional emergency diesel generators and all emergency core cooling systems were operable at the time of discovery.
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| Inspection Report# : 2000015(pdf)
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| Significance: Jun 16, 2000 Identified By: NRC Item Type: NCV NonCited Violation The licensee failed to ensure that appropriate post-modification testing was specified and accomplished for the Division I and Division III EDG output breaker circuitry modifications The licensee failed to ensure that the appropriate post-modification testing (PMT) was specified in the Division I and Division III emergency diesel generator (EDG) output breaker circuitry modification packages and that the post-modification tests were correctly accomplished. This was required to demonstrate through component and functional testing that the modified (rewired) portions of the Division I and Division III EDG output breaker circuitry were adequately installed to accomplish the intent of the plant design changes.
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| Inspection Report# : 2000012(pdf)
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| 3Q/2000 Inspection Findings - Clinton Page 3 of 8 Significance: N/A Dec 29, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW 10 CFR 55.59(c)(5) REQUIREMENTS FOR RETAINING LICENSED OPERATOR REQUALIFICATION PROGRAM RECORDS The inspectors identified a Non-Cited Violation wherein the facility licensee had failed to follow the Code of Federal Regulations (CFR) Title 10, Part 55.59(c)(5), Records, requirements by failing to systematically retain all of the original or authenticated copies of the original evaluation documents during the year 2000 annual NRC examination. The finding was of very low safety significance because although the records were not the original or authenticated copies of the original, records did exist in computerized clerically transcribed documents. The computer records had not been signed, and there was no indication that they had been verified correct by the original authors. The unauthenticated documents did provide information that licensed operators, for the most part, had participated and were evaluated during the year 2000 NRC annual requalification examination. However, the inspectors determined that the finding was more than minor. Specifically, the inspectors identified at least one instance in which the transcribed information appeared to be incorrect or missing. The records failure had credible impact on safety, in that, it negatively impacted on the intent of the licensed operator requalification examination process which, in part, is to maintain a high level of confidence that licensed operators continue to possess the requisite knowledge and abilities needed to safely perform licensed duties. In addition, inadequate records keeping adversely affects the NRC's ability to regulate.
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| Inspection Report# : 2001015(pdf)
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| Significance: Jan 26, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow Condition Report process for Shutdown Service Water (SX) pipe wall thinning Corrective actions were not implemented to replace a portion of the shutdown service water (SX) system piping after pipe wall thinning was identified. The failure to take the specified corrective actions by the committed due date or to properly reevaluate the degraded condition was determined to be a Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, "Procedures." This finding was determined to have very low safety significance because the SX system remained operable and capable of performing its' safety function.
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| Inspection Report# : 2001002(pdf)
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| Significance: Jan 26, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct longstanding Reactor Core Isolation Cooling (RCIC) System valve degradation Corrective actions for a longstanding deficiency with the Reactor Core Isolation Cooling (RCIC) system steam bypass valve were not effective in stopping the leakage past the valve. This finding was determined to be a Non-Cited Violation of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action." This finding was determined to have very low risk significance because the degraded condition of the valve did not affect the operability of the RCIC system.
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| Inspection Report# : 2001002(pdf)
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| Barrier Integrity Significance: Nov 16, 2000 Identified By: NRC Item Type: NCV NonCited Violation Secondary containment was inoperable for 6 minutes during fuel movements when interlock doors were opened.
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| Secondary containment was inoperable for 6 minutes during fuel movements when secondary containment interlock doors were inadvertently opened to move scaffolding. The inoperability was discovered when operators in the control room received an alarm indicating a loss of secondary containment vacuum. One Non-Cited Violation was identified for violating Technical Specification 3.6.4.1 which requires secondary containment operability during fuel moves. This finding was evaluated using the shutdown significance determination process contained in Appendix G of IMC 0609 and was determined to have very low risk significance because it did not meet the criteria for findings requiring a phase 2 significance evaluation.
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| Inspection Report# : 2000017(pdf)
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| Significance: Nov 14, 2000 Identified By: NRC
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| 3Q/2000 Inspection Findings - Clinton Page 4 of 8 Item Type: NCV NonCited Violation Failure to perform radiographic examinations of Class 2 welds.
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| The inspectors identified a Non-Cited Violation for the failure to perform radiographic examinations of Class 2 welds in compliance with applicable American Society of Mechanical Engineers (ASME) Code requirements. During installation testing of the 1999 Feedwater Keep Fill FW-39 modification, five radiographic examinations had recorded geometric unsharpness values which exceeded Section III and Section V ASME Code limits. Radiographic geometric unsharpness values are used to ensure that the film is of adequate quality to see defects. In addition, inspectors identified that three examinations did not meet Section V Code requirements for documentation of radiographic technique variables which can affect the image quality of the film. The safety significance of this issue was considered very low at this time, based on the absence of adverse consequences, the presence of other image quality indicators, and because the issue did not involve the system isolation valves. The failure to comply with ASME Code radiographic examination requirements could result in the failure to detect flaws within reactor coolant boundary piping, and was considered a Non-Cited Violation of 10 CFR Part 50.55a, "Codes and Standards".
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| Inspection Report# : 2000019(pdf)
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| Emergency Preparedness Significance: N/A Apr 28, 2000 Identified By: NRC Item Type: FIN Finding Emergency Preparedness Performance Indicator Verification Alert and Notification System, Drill & Exercise Participation, and Drill & Exercise performance indicators: The inspectors verified that the licensee had acceptably gathered information and reported these three performance indicators, which were in the green band, with the following minor exception. The inspectors identified a discrepancy with the licensee's initial assessment of the Drill and Exercise Performance (DEP) indicator related to the number of performance opportunities associated with a General Emergency declaration during a drill or an exercise. The licensee initially assumed that only three performance opportunities would exist rather than four as provided in NEI 99-02, but later recognized that they had misinterpreted the guidance. This did not affect the DEP performance indicator which was in the green band.
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| Inspection Report# : 2000009(pdf)
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| Significance: Aug 21, 2001 Identified By: NRC Item Type: NCV NonCited Violation Violation of 10 CFR 50.54(q) re. SCBA qualifications A Non-Cited Violation of 10 CFR 50.54(q) was identified by the NRC associated with the failure to maintain personnel qualifications for self contained breathing apparatus in accordance with the Clinton Power Station Emergency Plan. The finding was of very low safety significance because the licensee maintained an adequate number of qualified personnel to maintain minimum coverage of the required positions identified in the Emergency Plan.
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| Inspection Report# : 2001010(pdf)
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| Significance: Jun 08, 2001 Identified By: NRC Item Type: VIO Violation Supplemental Inspection -- Failure to correct self-identified defficiencies disclosed through control room communications drills This supplemental inspection was performed by the NRC to assess the licensee's evaluation associated with inaccuracies in the reporting of the Drill and Exercise Performance (DEP) performance indicator and with the performance deficiencies that resulted in a White DEP performance indicator (fourth quarter 1999 through the fourth quarter 2000). During the inspection, performed in accordance with NRC Inspection Procedure 95001, the inspector concluded that the licensee performed an adequate evaluation to determine the causes of both issues. In the case of the performance indicator errors, the licensee performed a root cause evaluation which identified a personnel error that was compounded by the lack of self-checking and verification. In addition, the licensee identified contributing causes that included the failure to provide adequate training to the emergency preparedness staff and the failure to provide adequate procedural guidance to the performance indicator data stewards and verifiers, which also applied to performance indicators in other cornerstones. The inspector concluded that the scope of corrective actions planned and implemented by the licensee appeared to address the identified causes. However, the inspector observed an additional discrepancy in the recently completed performance indicator evaluation related to drill and exercise participation. In addition, the licensee identified an error in its evaluation of one of the other emergency preparedness performance indicators that was not detected during its evaluation. These observations demonstrated weaknesses in the licensee's corrective actions and extent of condition review. The errors in the licensee's reporting of the DEP performance indicator was significant, in that the error resulted in a change of color, (i.e., Green-to-White). Consequently, a violation of 10 CFR 50.9 of more than minor safety significance was identified. Since the inaccurate reporting occurred during the period that the NRC's Enforcement Policy afforded discretion for the non-willful submittal of inaccurate performance indicator information, the NRC is exercising enforcement discretion and not citing the violation. In the case of the White DEP performance indicator, the inspector concluded that the licensee adequately assessed the deficiencies that led to the performance issues. Based on its review, the licensee attributed the White performance indicator to the high failure rate of control room communicator drills (i.e., job performance measures). The licensee identified two apparent causes for the high failure rate: (1) weaknesses in
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| 3Q/2000 Inspection Findings - Clinton Page 5 of 8 formal training; and (2) failure to meet emergency preparedness management expectations concerning the identification and correction of drill deficiencies. The inspector reviewed the licensee's corrective actions and determined that they addressed the causes identified. As a result of the licensee's immediate corrective actions, the licensee's performance returned the performance indicator to the Green band. The inspector and the licensee concluded that the high failure rate of the control room communicators resulted, in part, from inadequate corrective actions for self-identified deficiencies. Specifically, the licensee control room communicator drills were a portion of an overall annual evaluation of non-licensed operators, which included non-emergency preparedness functions. Generally, the failure of the communications segment of the evaluation did not result in a total failure of the annual evaluation. Therefore, the licensee's remedial actions were limited and were not effective in correcting the deficiencies and preventing similar failures from occurring, as required by 10 CFR 50.47(b)(14). By letter dated 08/22/01, the NRC concluded that a violation of 10 CFR 50.47(b)(14) had occurred and using the NRC's significance determination process, determined that the finding was white.
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| Inspection Report# : 2001009(pdf)
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| Significance: Feb 23, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to follow emergency plan for on-shift staffing For an approximate 2-month time period, the licensee failed to meet one of the minimum on-shift emergency response organization (ERO) staffing requirements contained in Table 2-1 of the licensee's emergency plan.
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| Inspection Report# : 2001003(pdf)
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| Occupational Radiation Safety Significance: Oct 08, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an Adequate Survey to identify and to post a High Radiation Area A finding and associated Non-Cited Violation was identified concerning the failure to perform an adequate radiological survey, as required by 10 CFR 20.1501. Although the licensee identified this issue, the licensee did not thoroughly evaluate the cause(s) of the unanticipated radiological conditions and associated problems in the monitoring of radioactive waste activities, which have resulted in previous, similar incidents. The finding was of very low safety significance because the area radiation levels and the licensee's additional administrative barriers would have limited the potential for an individual inadvertently entering the area and receiving a radiation exposure in excess of regulatory limits.
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| Inspection Report# : 2001015(pdf)
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| Significance: Aug 21, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to maintain administrative control of high radiation area keys as required by Technical Specification 5.7.2 Technical Specification 5.7.2 requires, in part, that doors to high radiation areas in which an individual could receive a deep dose equivalent greater than or equal to 1000 millirem in one hour (at 30 centimeters) shall be provided with locked or continuously guarded doors to prevent unauthorized entry and that the keys to such doors shall be administratively controlled. During October 29 - 31, 2001, the licensee failed to maintain administrative control of a key that controlled five access points to high radiation areas specified above (i.e., lost the key and failed to perform required key inventories to identify its loss), as described in CR No. 2-00-11-016. Since the inspector concluded that sufficient barriers remained to prevent an unauthorized individual from entering the affected areas and receiving an overexposure, the inspector concluded that the incident was of very low safety significance. The licensee also reported the incident to the NRC as an occurrence for the Occupational Exposure Control Effectiveness performance indicator. This is being treated as a Non-Cited Violation.
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| Inspection Report# : 2001010(pdf)
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| Significance: SL-IV Jul 27, 2001 Identified By: NRC Item Type: NCV NonCited Violation Misuse of Radioactive Material to Alarm a PCM Radiation protection technician used contaminated material to alarm a portal contamination monitor (PCM), while an individual was performing a contamination survey. Based on the licensee's investigation, the contamination was not placed on the individual, and the individual successfully monitored through an additional PCM. This incident will be reviewed by the NRC for potential enforcement actions. Update: On July 27, 2001, the NRC identified and forwarded to the licensee (by letter) a Non-Cited Violation of the Clinton Station Facility Operating License associated with the deliberate misuse of radioactive material by a junior contract radiation protection technician. On October 20, 2000, the technician misused radioactive material to cause an erroneous alarm on a PCM, as another individual was performing a contamination survey. The licensee identified
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| 3Q/2000 Inspection Findings - Clinton Page 6 of 8 the incident, entered the incident into its corrective action program, and implemented immediate corrective actions. Since the violation was determined to be willful, the NRC did not assign a significance to the violation using the NRC's Significance Determination Process. In accordance with the NRC Enforcement Policy, the NRC determined that the incident constituted a Severity Level IV violation of the Clinton Power Station Facility Operating License. Further, the NRC determined that the violation met the criteria necessary to disposition the violation as a Non-Cited Violation (Section VI.A.1.d of the NRC Enforcement Policy).
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| Inspection Report# : 2000018(pdf)
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| Inspection Report# : 2001010(pdf)
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| Significance: Oct 25, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Three individuals entered a HRA in violation of Technical Specification 5.7.1 On October 25, 2000, three individuals entered the B residual heat removal heat exchanger room (a posted high radiation area); however, the individuals were not working under a radiation work permit that allowed entry into the high radiation area and did not satisfy either of the three entry conditions of Technical Specification 5.7.1.
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| Inspection Report# : 2000018(pdf)
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| Public Radiation Safety Significance: Dec 08, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Inadvertent Release of Radioactive Material to Unrestricted Area During September 2000, the licensee conducted a survey of tools, equipment, etc. outside of the restricted area (protected area and owner controlled area) and identified low-level contaminated materials that were not under constant surveillance or control. The failure to maintain contstant surveillance and control of the material was a violation of 10 CFR 20.1802 and was characterized as a Non-Cited Violation. Based on the licensee's conservative annual dose assessment (about 1.56 millirem) and the inability to define the origin of each of the items, the inspector concluded that the issue constituted one occurrence/event per the NRC Significance Determination Process (Green).
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| Inspection Report# : 2000021(pdf)
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| Physical Protection Miscellaneous Significance: N/A Sep 30, 2000 Identified By: NRC Item Type: FIN Finding Recent events affecting plant operations contained elements of human performance deficiencies.
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| NO COLOR. The inspectors noted that several recent events which have affected plant operations and the operability of safety-related components or other components important to safety contained elements of human performance deficiencies. The human performance aspects, while not always being the root cause of the problem, were significant contributors leading to the events. While the risk of the individual events was very low, the number of maintenance-related incidents indicated a problem exists with the control, review, and performance of maintenance activities.
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| Inspection Report# : 2000015(pdf)
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| Significance: N/A May 20, 2000 Identified By: NRC Item Type: FIN Finding Inaccurate historical data for the Safety System Functional Failure Indicator No Color. The licensee identified a failure to submit accurate information to the NRC. The inaccurate information involved the historical data submittal for the Safety System Functional Failure Performance Indicator. The error resulted in a response band color change from Green to White for the first quarter 1999 Performance Indicator. The NRC exercised Enforcement Discretion pursuant to Section VII.B.6 of the Enforcement Policy and did not cite the violation.
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| 3Q/2000 Inspection Findings - Clinton Page 7 of 8 Inspection Report# : 2000008(pdf)
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| Significance: Feb 17, 2002 Identified By: NRC Item Type: FIN Finding A temporary modification on the "A" RR FCV control cirucuitry.
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| On December 14, 2001, the licensee installed a temporary modification on the "A" RR FCV control circuitry. The T-mod was installed to assist the operators in manually controlling the "A" RR FCV because the reliability of the normal control circuitry was in question. During the implementation portion of the T-mod installations, the "A" RR FCV unexpectedly moved from 94 percent open to 102 percent open at which point the protective position circuitry locked the valve at the 102 percent position. Recator power was observed to go from 94 percent to 98 percent during this unexpected valve movement. Following this unexpected FCV movement, opeerations personnel ordered the T-mod to be removed and operators then proceeded to manually shut down the reactor without any further movements of the "A" RR FCV.
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| Inspection Report# : 2001016(pdf)
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| Significance: Feb 17, 2002 Identified By: NRC Item Type: NCV NonCited Violation Non-cited violation of T.S. 5.4.1 for an inadequate surveillance procedure.
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| Inspection Report# : 2001016(pdf)
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| Significance: SL-IV Aug 18, 2001 Identified By: NRC Item Type: VIO Violation Falsification of Test Records by Licensee Employee SL IV - On July 2, 2001, by separate letter, NRC issued a Severity Level IV violation of 10 CFR 50.9 for a deliberate falsification by a plant test engineer. Following investigation by the Office of Investigations, NRC determined that, on October 20, 2000, a test engineer forged another employee's signature on two test package cover sheets on by forging another employee's signature without his prior concurrence, in violation of Clinton established plant protocol and procedure.
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| Inspection Report# : 2001010(pdf)
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| Significance: SL-IV Apr 06, 2001 Identified By: NRC Item Type: VIO Violation Violation of 10 CFR 50.7 "Employee Protection" On April 6, 2001, the NRC issued the licensee a Severity Level IV Violation of 10 CFR 50.7. The NRC concluded that the licensee took adverse employment actions against an employee in the licensee's Nuclear Training Department (i.e., unfavorable 1999 performance review), in part, as a result of the employee's engagement in protected activities. In addition, the NRC learned that several training personnel may be reluctant to discuss department issues within the nuclear training department.
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| Inspection Report# : 2001007(pdf)
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| Inspection Report# : 2001010(pdf)
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| Significance: N/A Jan 26, 2001 Identified By: NRC Item Type: FIN Finding Assessment of Problem Identification and Resolution Performance The team identified that the licensee appropriately entered significant plant issues into the corrective action process by initiating condition reports.
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| Some less significant conditions adverse to quality were evaluated and corrected outside the established process. The trending program was not fully effective as a problem identification tool. Quality Assurance audits and self-assessments reviewed varied in quality. Identified issues were generally evaluated properly, although in several cases the corrective action process did not work effectively to either evaluate or prioritize issues.
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| Current station performance issues including human performance, corrective action program, surveillance testing, and labeling indicate that long term corrective actions previously taken in these areas as restart and post-restart initiatives have not been fully effective to support sustained improvement. Corrective actions were not always fully effective or timely for some individual equipment issues and the effectiveness review process (CARE) did not always identify ineffective corrective actions. The licensee had recently recognized similar deficiencies in corrective action program implementation but had not yet fully developed or completed the corrective actions to improve these areas. The inspectors did not find any reluctance by the station employees to raise safety issues.
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| Inspection Report# : 2001002(pdf)
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| Significance: N/A Dec 31, 2000 Identified By: NRC Item Type: FIN Finding
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| 3Q/2000 Inspection Findings - Clinton Page 8 of 8 Recent human performance issues have occurred which are associated with operator performance and knowledge based deficiencies that have affected plant operations and responses to transient conditions.
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| Recent human performance issues have occurred which are associated with operator performance and knowledge based deficiencies that have affected plant operations and responses to transient conditions. While the risk of the individual events was very low, the failure of operators to adequately control level parameters indicated a declining trend in this area. These issues could not be easily evaluated by present risk analysis methods because failures to follow procedures and maintaining management expectations were not modeled in the Clinton Individual Plant Evaluation. Therefore, the finding is characterized as having no color.
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| Inspection Report# : 2000020(pdf)
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| Significance: N/A Nov 14, 2000 Identified By: NRC Item Type: FIN Finding Three procedures were not written in compliance with the applicable ASME Code.
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| The inspectors reviewed three special process procedures, and identified areas where all three procedures were not written in compliance with the applicable ASME Code. The procedure deficiencies had the potential to affect the ASME Code compliance of weld fabrication and nondestructive examination used on safety-related components and piping. The inspectors noted that each of the ASME Code problems identified contained elements of human performance deficiencies. The human performance aspects, while not always being the root cause of the problem, were significant contributors leading to procedure deficiencies. While the risk of the individual examples was very low, the number of deficiencies indicated a problem with incorporation of applicable ASME Code requirements into special process procedures.
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| Inspection Report# : 2000019(pdf)
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| Last modified : March 29, 2002
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| 4Q/2000 Inspection Findings - Clinton Page 1 of 8 Clinton Initiating Events Significance: Dec 31, 2000 Identified By: NRC Item Type: NCV NonCited Violation Operators failed to adequately evaluate an alarming moisture separator drain tank level annunciator that resulted in a turbine trip.
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| During plant restart following refueling outage 7, operators did not adequately evaluate an alarming moisture separator drain tank level annunciator.
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| As a result, high water level in the moisture separator drain tank caused a turbine trip with the reactor at approximately 25% power. The inspectors reviewed this issue using the significance determination process for a transient. Since only the initiating event cornerstone is affected and associated assumptions have no other impact than slightly increasing the likelihood of an uncomplicated reactor trip, the finding is considered to be of very low safety significance.
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| Inspection Report# : 2000020(pdf)
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| Significance: Dec 31, 2000 Identified By: NRC Item Type: NCV NonCited Violation Operators failed to adequately control reactor vessel water level and pressure following the automatic reactor scram which resulted in a second automatic scram.
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| Operators failed to adequately control reactor vessel water level and pressure, while attempting to open the main steam isolation valves following the automatic reactor scram on December 18, 2000. This resulted in an automatic scram signal due to low reactor vessel water level. The inspectors reviewed this issue using the significance determination process for a transient with a loss of feedwater and determined this was of very low safety significance because the event occurred while the reactor was shut down and all control rods were already fully inserted.
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| Inspection Report# : 2000020(pdf)
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| Significance: Dec 31, 2000 Identified By: NRC Item Type: FIN Finding Operators did not adequately control reactor vessel inventory after a reactor scram which resulted in the motor driven reactor feedwater pump tripping on high reactor vessel water level.
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| During operator response to a reactor scram on December 18, 2000, operators did not adequately control reactor vessel inventory prior to the motor driven reactor feedwater pump tripping on high reactor vessel water level. The inspectors reviewed this issue using the significance determination process for a transient with a loss of feedwater and determined this was of very low safety significance because all other reactor vessel level control systems were operable and functioned as designed.
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| Inspection Report# : 2000020(pdf)
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| Significance: Nov 16, 2000 Identified By: NRC Item Type: NCV NonCited Violation An alternate rod insertion system initiation and a manual reactor scram occurred with the reactor shutdown as a result of an inadequate maintenance procedure.
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| During replacement of power supplies for the alternate rod insertion (ARI) system, maintenance personnel failed to fully evaluate the impacts that re-energizing the power supplies had on the ARI initiation logic. While re-energizing the power supplies, the initiation logic sensed an ARI signal (low reactor water level). This caused the vent and drain valves to close and the scram discharge volume to fill with water. Plant operators inserted a manual scram signal before the automatic high scram discharge volume set point was reached. One Non-Cited Violation was identified for having an inadequate maintenance procedure to control this activity. This finding was evaluated using the shutdown significance determination process contained in Appendix G of IMC 0609 and was determined to have very low risk significance because it did not impact any of the five shutdown safety functions identified by NUMARC 91-06.
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| Inspection Report# : 2000017(pdf)
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| 4Q/2000 Inspection Findings - Clinton Page 2 of 8 Significance: Sep 30, 2000 Identified By: NRC Item Type: FIN Finding Human performance errors and an inadequate procedure resulted in exceeding the allowed outage time for the emergency reserve auxiliary transformer static VAR compensator.
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| Human performance errors and the failure to develop an adequate procedure for the emergency reserve auxiliary transformer static VAR (Volt Ampere Reactive) compensator (ERAT-SVC) surveillance test resulted in several delays during the test. These delays caused the work to not be completed within the allowed outage time. Therefore, a request for Enforcement Discretion was presented to the NRC which was formally granted on September 20, 2000 (NOED 00-6-011). The safety significance of this finding was very low because all other emergency core cooling system trains (automatic depressurization system, low pressure core spray, and low pressure core injection), emergency diesel generators, and the reactor core isolation cooling system were operable.
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| Inspection Report# : 2000015(pdf)
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| Significance: May 17, 2000 Identified By: Self Disclosing Item Type: FIN Finding Manual reactor shutdown A labeling discrepancy contributed to the improper isolation of a protective relay for the 4.16kV Bus 1B Reserve Feed Breaker. As a result, during functional testing, the relay actuated and caused the bus to be de-energized which ultimately resulted in a manual reactor shut down. This issue was determined to be of very low risk significance due to remaining mitigation capability and recovery potential.
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| Inspection Report# : 2000008(pdf)
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| Significance: Aug 21, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow procedures associated with feed water level control system surveillance testing.
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| Human performance and corrective action deficiencies contributed to a Non-Cited Violation of Technical Specification 5.4.1 for failing to follow procedures. This led to the unplanned automatic reactor shutdown on July 24, 2001. The finding was of very low safety significance because no complications occurred during the unplanned automatic reactor shut down and the finding did not increase the likelihood of mitigation equipment being unavailable.
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| Inspection Report# : 2001010(pdf)
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| Mitigating Systems Significance: Sep 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation Procedural requirements were not followed when unexpected equipment response was encountered.
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| Maintenance personnel failed to appropriately follow procedure instructions during testing of the Division III emergency diesel generator room fire detection system. These actions led to the emergency diesel generator being rendered inoperable. The procedure violation was treated as a Non-Cited Violation. This issue was of very low safety significance since the other divisional emergency diesel generators and all emergency core cooling systems were operable at the time of discovery.
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| Inspection Report# : 2000015(pdf)
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| Significance: Jun 16, 2000 Identified By: NRC Item Type: NCV NonCited Violation The licensee failed to ensure that appropriate post-modification testing was specified and accomplished for the Division I and Division III EDG output breaker circuitry modifications The licensee failed to ensure that the appropriate post-modification testing (PMT) was specified in the Division I and Division III emergency diesel generator (EDG) output breaker circuitry modification packages and that the post-modification tests were correctly accomplished. This was required to demonstrate through component and functional testing that the modified (rewired) portions of the Division I and Division III EDG output breaker
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| 4Q/2000 Inspection Findings - Clinton Page 3 of 8 circuitry were adequately installed to accomplish the intent of the plant design changes.
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| Inspection Report# : 2000012(pdf)
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| Significance: N/A Dec 29, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW 10 CFR 55.59(c)(5) REQUIREMENTS FOR RETAINING LICENSED OPERATOR REQUALIFICATION PROGRAM RECORDS The inspectors identified a Non-Cited Violation wherein the facility licensee had failed to follow the Code of Federal Regulations (CFR) Title 10, Part 55.59(c)(5), Records, requirements by failing to systematically retain all of the original or authenticated copies of the original evaluation documents during the year 2000 annual NRC examination. The finding was of very low safety significance because although the records were not the original or authenticated copies of the original, records did exist in computerized clerically transcribed documents. The computer records had not been signed, and there was no indication that they had been verified correct by the original authors. The unauthenticated documents did provide information that licensed operators, for the most part, had participated and were evaluated during the year 2000 NRC annual requalification examination. However, the inspectors determined that the finding was more than minor. Specifically, the inspectors identified at least one instance in which the transcribed information appeared to be incorrect or missing. The records failure had credible impact on safety, in that, it negatively impacted on the intent of the licensed operator requalification examination process which, in part, is to maintain a high level of confidence that licensed operators continue to possess the requisite knowledge and abilities needed to safely perform licensed duties. In addition, inadequate records keeping adversely affects the NRC's ability to regulate.
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| Inspection Report# : 2001015(pdf)
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| Significance: Jan 26, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct longstanding Reactor Core Isolation Cooling (RCIC) System valve degradation Corrective actions for a longstanding deficiency with the Reactor Core Isolation Cooling (RCIC) system steam bypass valve were not effective in stopping the leakage past the valve. This finding was determined to be a Non-Cited Violation of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action." This finding was determined to have very low risk significance because the degraded condition of the valve did not affect the operability of the RCIC system.
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| Inspection Report# : 2001002(pdf)
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| Significance: Jan 26, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow Condition Report process for Shutdown Service Water (SX) pipe wall thinning Corrective actions were not implemented to replace a portion of the shutdown service water (SX) system piping after pipe wall thinning was identified. The failure to take the specified corrective actions by the committed due date or to properly reevaluate the degraded condition was determined to be a Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, "Procedures." This finding was determined to have very low safety significance because the SX system remained operable and capable of performing its' safety function.
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| Inspection Report# : 2001002(pdf)
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| Barrier Integrity Significance: Nov 16, 2000 Identified By: NRC Item Type: NCV NonCited Violation Secondary containment was inoperable for 6 minutes during fuel movements when interlock doors were opened.
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| Secondary containment was inoperable for 6 minutes during fuel movements when secondary containment interlock doors were inadvertently opened to move scaffolding. The inoperability was discovered when operators in the control room received an alarm indicating a loss of secondary containment vacuum. One Non-Cited Violation was identified for violating Technical Specification 3.6.4.1 which requires secondary containment operability during fuel moves. This finding was evaluated using the shutdown significance determination process contained in Appendix G of IMC 0609 and was determined to have very low risk significance because it did not meet the criteria for findings requiring a phase 2 significance evaluation.
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| Inspection Report# : 2000017(pdf)
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| 4Q/2000 Inspection Findings - Clinton Page 4 of 8 Significance: Nov 14, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform radiographic examinations of Class 2 welds.
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| The inspectors identified a Non-Cited Violation for the failure to perform radiographic examinations of Class 2 welds in compliance with applicable American Society of Mechanical Engineers (ASME) Code requirements. During installation testing of the 1999 Feedwater Keep Fill FW-39 modification, five radiographic examinations had recorded geometric unsharpness values which exceeded Section III and Section V ASME Code limits. Radiographic geometric unsharpness values are used to ensure that the film is of adequate quality to see defects. In addition, inspectors identified that three examinations did not meet Section V Code requirements for documentation of radiographic technique variables which can affect the image quality of the film. The safety significance of this issue was considered very low at this time, based on the absence of adverse consequences, the presence of other image quality indicators, and because the issue did not involve the system isolation valves. The failure to comply with ASME Code radiographic examination requirements could result in the failure to detect flaws within reactor coolant boundary piping, and was considered a Non-Cited Violation of 10 CFR Part 50.55a, "Codes and Standards".
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| Inspection Report# : 2000019(pdf)
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| Emergency Preparedness Significance: N/A Apr 28, 2000 Identified By: NRC Item Type: FIN Finding Emergency Preparedness Performance Indicator Verification Alert and Notification System, Drill & Exercise Participation, and Drill & Exercise performance indicators: The inspectors verified that the licensee had acceptably gathered information and reported these three performance indicators, which were in the green band, with the following minor exception. The inspectors identified a discrepancy with the licensee's initial assessment of the Drill and Exercise Performance (DEP) indicator related to the number of performance opportunities associated with a General Emergency declaration during a drill or an exercise. The licensee initially assumed that only three performance opportunities would exist rather than four as provided in NEI 99-02, but later recognized that they had misinterpreted the guidance. This did not affect the DEP performance indicator which was in the green band.
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| Inspection Report# : 2000009(pdf)
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| Significance: Aug 21, 2001 Identified By: NRC Item Type: NCV NonCited Violation Violation of 10 CFR 50.54(q) re. SCBA qualifications A Non-Cited Violation of 10 CFR 50.54(q) was identified by the NRC associated with the failure to maintain personnel qualifications for self contained breathing apparatus in accordance with the Clinton Power Station Emergency Plan. The finding was of very low safety significance because the licensee maintained an adequate number of qualified personnel to maintain minimum coverage of the required positions identified in the Emergency Plan.
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| Inspection Report# : 2001010(pdf)
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| Significance: Jun 08, 2001 Identified By: NRC Item Type: VIO Violation Supplemental Inspection -- Failure to correct self-identified defficiencies disclosed through control room communications drills This supplemental inspection was performed by the NRC to assess the licensee's evaluation associated with inaccuracies in the reporting of the Drill and Exercise Performance (DEP) performance indicator and with the performance deficiencies that resulted in a White DEP performance indicator (fourth quarter 1999 through the fourth quarter 2000). During the inspection, performed in accordance with NRC Inspection Procedure 95001, the inspector concluded that the licensee performed an adequate evaluation to determine the causes of both issues. In the case of the performance indicator errors, the licensee performed a root cause evaluation which identified a personnel error that was compounded by the lack of self-checking and verification. In addition, the licensee identified contributing causes that included the failure to provide adequate training to the emergency preparedness staff and the failure to provide adequate procedural guidance to the performance indicator data stewards and verifiers, which also applied to performance indicators in other cornerstones. The inspector concluded that the scope of corrective actions planned and implemented by the licensee appeared to address the identified causes. However, the inspector observed an additional discrepancy in the recently completed performance indicator evaluation related to drill and exercise participation. In addition, the licensee identified an error in its evaluation of one of the other emergency preparedness performance indicators that was not detected during its evaluation. These observations demonstrated weaknesses in the licensee's corrective actions and extent of condition review. The errors in the licensee's reporting of the DEP performance indicator was significant, in that the error resulted in a change of color, (i.e., Green-to-White). Consequently, a violation of 10 CFR 50.9 of more than minor safety significance was identified. Since the inaccurate reporting occurred during the period that the NRC's Enforcement Policy afforded
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| 4Q/2000 Inspection Findings - Clinton Page 5 of 8 discretion for the non-willful submittal of inaccurate performance indicator information, the NRC is exercising enforcement discretion and not citing the violation. In the case of the White DEP performance indicator, the inspector concluded that the licensee adequately assessed the deficiencies that led to the performance issues. Based on its review, the licensee attributed the White performance indicator to the high failure rate of control room communicator drills (i.e., job performance measures). The licensee identified two apparent causes for the high failure rate: (1) weaknesses in formal training; and (2) failure to meet emergency preparedness management expectations concerning the identification and correction of drill deficiencies. The inspector reviewed the licensee's corrective actions and determined that they addressed the causes identified. As a result of the licensee's immediate corrective actions, the licensee's performance returned the performance indicator to the Green band. The inspector and the licensee concluded that the high failure rate of the control room communicators resulted, in part, from inadequate corrective actions for self-identified deficiencies. Specifically, the licensee control room communicator drills were a portion of an overall annual evaluation of non-licensed operators, which included non-emergency preparedness functions. Generally, the failure of the communications segment of the evaluation did not result in a total failure of the annual evaluation. Therefore, the licensee's remedial actions were limited and were not effective in correcting the deficiencies and preventing similar failures from occurring, as required by 10 CFR 50.47(b)(14). By letter dated 08/22/01, the NRC concluded that a violation of 10 CFR 50.47(b)(14) had occurred and using the NRC's significance determination process, determined that the finding was white.
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| Inspection Report# : 2001009(pdf)
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| Significance: Feb 23, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to follow emergency plan for on-shift staffing For an approximate 2-month time period, the licensee failed to meet one of the minimum on-shift emergency response organization (ERO) staffing requirements contained in Table 2-1 of the licensee's emergency plan.
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| Inspection Report# : 2001003(pdf)
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| Occupational Radiation Safety Significance: Oct 25, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Three individuals entered a HRA in violation of Technical Specification 5.7.1 On October 25, 2000, three individuals entered the B residual heat removal heat exchanger room (a posted high radiation area); however, the individuals were not working under a radiation work permit that allowed entry into the high radiation area and did not satisfy either of the three entry conditions of Technical Specification 5.7.1.
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| Inspection Report# : 2000018(pdf)
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| Significance: Oct 08, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an Adequate Survey to identify and to post a High Radiation Area A finding and associated Non-Cited Violation was identified concerning the failure to perform an adequate radiological survey, as required by 10 CFR 20.1501. Although the licensee identified this issue, the licensee did not thoroughly evaluate the cause(s) of the unanticipated radiological conditions and associated problems in the monitoring of radioactive waste activities, which have resulted in previous, similar incidents. The finding was of very low safety significance because the area radiation levels and the licensee's additional administrative barriers would have limited the potential for an individual inadvertently entering the area and receiving a radiation exposure in excess of regulatory limits.
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| Inspection Report# : 2001015(pdf)
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| Significance: Aug 21, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to maintain administrative control of high radiation area keys as required by Technical Specification 5.7.2 Technical Specification 5.7.2 requires, in part, that doors to high radiation areas in which an individual could receive a deep dose equivalent greater than or equal to 1000 millirem in one hour (at 30 centimeters) shall be provided with locked or continuously guarded doors to prevent unauthorized entry and that the keys to such doors shall be administratively controlled. During October 29 - 31, 2001, the licensee failed to maintain administrative control of a key that controlled five access points to high radiation areas specified above (i.e., lost the key and failed to perform required key inventories to identify its loss), as described in CR No. 2-00-11-016. Since the inspector concluded that sufficient barriers remained to
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| 4Q/2000 Inspection Findings - Clinton Page 6 of 8 prevent an unauthorized individual from entering the affected areas and receiving an overexposure, the inspector concluded that the incident was of very low safety significance. The licensee also reported the incident to the NRC as an occurrence for the Occupational Exposure Control Effectiveness performance indicator. This is being treated as a Non-Cited Violation.
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| Inspection Report# : 2001010(pdf)
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| Significance: SL-IV Jul 27, 2001 Identified By: NRC Item Type: NCV NonCited Violation Misuse of Radioactive Material to Alarm a PCM Radiation protection technician used contaminated material to alarm a portal contamination monitor (PCM), while an individual was performing a contamination survey. Based on the licensee's investigation, the contamination was not placed on the individual, and the individual successfully monitored through an additional PCM. This incident will be reviewed by the NRC for potential enforcement actions. Update: On July 27, 2001, the NRC identified and forwarded to the licensee (by letter) a Non-Cited Violation of the Clinton Station Facility Operating License associated with the deliberate misuse of radioactive material by a junior contract radiation protection technician. On October 20, 2000, the technician misused radioactive material to cause an erroneous alarm on a PCM, as another individual was performing a contamination survey. The licensee identified the incident, entered the incident into its corrective action program, and implemented immediate corrective actions. Since the violation was determined to be willful, the NRC did not assign a significance to the violation using the NRC's Significance Determination Process. In accordance with the NRC Enforcement Policy, the NRC determined that the incident constituted a Severity Level IV violation of the Clinton Power Station Facility Operating License. Further, the NRC determined that the violation met the criteria necessary to disposition the violation as a Non-Cited Violation (Section VI.A.1.d of the NRC Enforcement Policy).
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| Inspection Report# : 2001010(pdf)
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| Inspection Report# : 2000018(pdf)
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| Public Radiation Safety Significance: Dec 08, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Inadvertent Release of Radioactive Material to Unrestricted Area During September 2000, the licensee conducted a survey of tools, equipment, etc. outside of the restricted area (protected area and owner controlled area) and identified low-level contaminated materials that were not under constant surveillance or control. The failure to maintain contstant surveillance and control of the material was a violation of 10 CFR 20.1802 and was characterized as a Non-Cited Violation. Based on the licensee's conservative annual dose assessment (about 1.56 millirem) and the inability to define the origin of each of the items, the inspector concluded that the issue constituted one occurrence/event per the NRC Significance Determination Process (Green).
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| Inspection Report# : 2000021(pdf)
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| Physical Protection Miscellaneous Significance: N/A Dec 31, 2000 Identified By: NRC Item Type: FIN Finding Recent human performance issues have occurred which are associated with operator performance and knowledge based deficiencies that have affected plant operations and responses to transient conditions.
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| Recent human performance issues have occurred which are associated with operator performance and knowledge based deficiencies that have affected plant operations and responses to transient conditions. While the risk of the individual events was very low, the failure of operators to adequately control level parameters indicated a declining trend in this area. These issues could not be easily evaluated by present risk analysis methods because failures to follow procedures and maintaining management expectations were not modeled in the Clinton Individual Plant Evaluation. Therefore, the finding is characterized as having no color.
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| Inspection Report# : 2000020(pdf)
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| Significance: N/A Nov 14, 2000 Identified By: NRC
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| 4Q/2000 Inspection Findings - Clinton Page 7 of 8 Item Type: FIN Finding Three procedures were not written in compliance with the applicable ASME Code.
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| The inspectors reviewed three special process procedures, and identified areas where all three procedures were not written in compliance with the applicable ASME Code. The procedure deficiencies had the potential to affect the ASME Code compliance of weld fabrication and nondestructive examination used on safety-related components and piping. The inspectors noted that each of the ASME Code problems identified contained elements of human performance deficiencies. The human performance aspects, while not always being the root cause of the problem, were significant contributors leading to procedure deficiencies. While the risk of the individual examples was very low, the number of deficiencies indicated a problem with incorporation of applicable ASME Code requirements into special process procedures.
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| Inspection Report# : 2000019(pdf)
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| Significance: N/A Sep 30, 2000 Identified By: NRC Item Type: FIN Finding Recent events affecting plant operations contained elements of human performance deficiencies.
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| NO COLOR. The inspectors noted that several recent events which have affected plant operations and the operability of safety-related components or other components important to safety contained elements of human performance deficiencies. The human performance aspects, while not always being the root cause of the problem, were significant contributors leading to the events. While the risk of the individual events was very low, the number of maintenance-related incidents indicated a problem exists with the control, review, and performance of maintenance activities.
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| Inspection Report# : 2000015(pdf)
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| Significance: N/A May 20, 2000 Identified By: NRC Item Type: FIN Finding Inaccurate historical data for the Safety System Functional Failure Indicator No Color. The licensee identified a failure to submit accurate information to the NRC. The inaccurate information involved the historical data submittal for the Safety System Functional Failure Performance Indicator. The error resulted in a response band color change from Green to White for the first quarter 1999 Performance Indicator. The NRC exercised Enforcement Discretion pursuant to Section VII.B.6 of the Enforcement Policy and did not cite the violation.
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| Inspection Report# : 2000008(pdf)
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| Significance: Feb 17, 2002 Identified By: NRC Item Type: NCV NonCited Violation Non-cited violation of T.S. 5.4.1 for an inadequate surveillance procedure.
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| Inspection Report# : 2001016(pdf)
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| Significance: Feb 17, 2002 Identified By: NRC Item Type: FIN Finding A temporary modification on the "A" RR FCV control cirucuitry.
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| On December 14, 2001, the licensee installed a temporary modification on the "A" RR FCV control circuitry. The T-mod was installed to assist the operators in manually controlling the "A" RR FCV because the reliability of the normal control circuitry was in question. During the implementation portion of the T-mod installations, the "A" RR FCV unexpectedly moved from 94 percent open to 102 percent open at which point the protective position circuitry locked the valve at the 102 percent position. Recator power was observed to go from 94 percent to 98 percent during this unexpected valve movement. Following this unexpected FCV movement, opeerations personnel ordered the T-mod to be removed and operators then proceeded to manually shut down the reactor without any further movements of the "A" RR FCV.
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| Inspection Report# : 2001016(pdf)
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| Significance: SL-IV Aug 18, 2001 Identified By: NRC Item Type: VIO Violation Falsification of Test Records by Licensee Employee SL IV - On July 2, 2001, by separate letter, NRC issued a Severity Level IV violation of 10 CFR 50.9 for a deliberate falsification by a plant test engineer. Following investigation by the Office of Investigations, NRC determined that, on October 20, 2000, a test engineer forged another employee's signature on two test package cover sheets on by forging another employee's signature without his prior concurrence, in violation of Clinton established plant protocol and procedure.
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| Inspection Report# : 2001010(pdf)
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| Significance: SL-IV Apr 06, 2001 Identified By: NRC
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| 4Q/2000 Inspection Findings - Clinton Page 8 of 8 Item Type: VIO Violation Violation of 10 CFR 50.7 "Employee Protection" On April 6, 2001, the NRC issued the licensee a Severity Level IV Violation of 10 CFR 50.7. The NRC concluded that the licensee took adverse employment actions against an employee in the licensee's Nuclear Training Department (i.e., unfavorable 1999 performance review), in part, as a result of the employee's engagement in protected activities. In addition, the NRC learned that several training personnel may be reluctant to discuss department issues within the nuclear training department.
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| Inspection Report# : 2001007(pdf)
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| Inspection Report# : 2001010(pdf)
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| Significance: N/A Jan 26, 2001 Identified By: NRC Item Type: FIN Finding Assessment of Problem Identification and Resolution Performance The team identified that the licensee appropriately entered significant plant issues into the corrective action process by initiating condition reports.
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| Some less significant conditions adverse to quality were evaluated and corrected outside the established process. The trending program was not fully effective as a problem identification tool. Quality Assurance audits and self-assessments reviewed varied in quality. Identified issues were generally evaluated properly, although in several cases the corrective action process did not work effectively to either evaluate or prioritize issues.
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| Current station performance issues including human performance, corrective action program, surveillance testing, and labeling indicate that long term corrective actions previously taken in these areas as restart and post-restart initiatives have not been fully effective to support sustained improvement. Corrective actions were not always fully effective or timely for some individual equipment issues and the effectiveness review process (CARE) did not always identify ineffective corrective actions. The licensee had recently recognized similar deficiencies in corrective action program implementation but had not yet fully developed or completed the corrective actions to improve these areas. The inspectors did not find any reluctance by the station employees to raise safety issues.
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| Inspection Report# : 2001002(pdf)
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| Last modified : March 28, 2002
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| 1Q/2001 Inspection Findings - Clinton Page 1 of 8 Clinton Initiating Events Significance: Dec 31, 2000 Identified By: NRC Item Type: NCV NonCited Violation Operators failed to adequately control reactor vessel water level and pressure following the automatic reactor scram which resulted in a second automatic scram.
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| Operators failed to adequately control reactor vessel water level and pressure, while attempting to open the main steam isolation valves following the automatic reactor scram on December 18, 2000. This resulted in an automatic scram signal due to low reactor vessel water level. The inspectors reviewed this issue using the significance determination process for a transient with a loss of feedwater and determined this was of very low safety significance because the event occurred while the reactor was shut down and all control rods were already fully inserted.
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| Inspection Report# : 2000020(pdf)
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| Significance: Dec 31, 2000 Identified By: NRC Item Type: NCV NonCited Violation Operators failed to adequately evaluate an alarming moisture separator drain tank level annunciator that resulted in a turbine trip.
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| During plant restart following refueling outage 7, operators did not adequately evaluate an alarming moisture separator drain tank level annunciator.
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| As a result, high water level in the moisture separator drain tank caused a turbine trip with the reactor at approximately 25% power. The inspectors reviewed this issue using the significance determination process for a transient. Since only the initiating event cornerstone is affected and associated assumptions have no other impact than slightly increasing the likelihood of an uncomplicated reactor trip, the finding is considered to be of very low safety significance.
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| Inspection Report# : 2000020(pdf)
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| Significance: Dec 31, 2000 Identified By: NRC Item Type: FIN Finding Operators did not adequately control reactor vessel inventory after a reactor scram which resulted in the motor driven reactor feedwater pump tripping on high reactor vessel water level.
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| During operator response to a reactor scram on December 18, 2000, operators did not adequately control reactor vessel inventory prior to the motor driven reactor feedwater pump tripping on high reactor vessel water level. The inspectors reviewed this issue using the significance determination process for a transient with a loss of feedwater and determined this was of very low safety significance because all other reactor vessel level control systems were operable and functioned as designed.
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| Inspection Report# : 2000020(pdf)
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| Significance: Nov 16, 2000 Identified By: NRC Item Type: NCV NonCited Violation An alternate rod insertion system initiation and a manual reactor scram occurred with the reactor shutdown as a result of an inadequate maintenance procedure.
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| During replacement of power supplies for the alternate rod insertion (ARI) system, maintenance personnel failed to fully evaluate the impacts that re-energizing the power supplies had on the ARI initiation logic. While re-energizing the power supplies, the initiation logic sensed an ARI signal (low reactor water level). This caused the vent and drain valves to close and the scram discharge volume to fill with water. Plant operators inserted a manual scram signal before the automatic high scram discharge volume set point was reached. One Non-Cited Violation was identified for having an inadequate maintenance procedure to control this activity. This finding was evaluated using the shutdown significance determination process contained in Appendix G of IMC 0609 and was determined to have very low risk significance because it did not impact any of the five shutdown safety functions identified by NUMARC 91-06.
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| Inspection Report# : 2000017(pdf)
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| 1Q/2001 Inspection Findings - Clinton Page 2 of 8 Significance: Sep 30, 2000 Identified By: NRC Item Type: FIN Finding Human performance errors and an inadequate procedure resulted in exceeding the allowed outage time for the emergency reserve auxiliary transformer static VAR compensator.
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| Human performance errors and the failure to develop an adequate procedure for the emergency reserve auxiliary transformer static VAR (Volt Ampere Reactive) compensator (ERAT-SVC) surveillance test resulted in several delays during the test. These delays caused the work to not be completed within the allowed outage time. Therefore, a request for Enforcement Discretion was presented to the NRC which was formally granted on September 20, 2000 (NOED 00-6-011). The safety significance of this finding was very low because all other emergency core cooling system trains (automatic depressurization system, low pressure core spray, and low pressure core injection), emergency diesel generators, and the reactor core isolation cooling system were operable.
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| Inspection Report# : 2000015(pdf)
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| Significance: May 17, 2000 Identified By: Self Disclosing Item Type: FIN Finding Manual reactor shutdown A labeling discrepancy contributed to the improper isolation of a protective relay for the 4.16kV Bus 1B Reserve Feed Breaker. As a result, during functional testing, the relay actuated and caused the bus to be de-energized which ultimately resulted in a manual reactor shut down. This issue was determined to be of very low risk significance due to remaining mitigation capability and recovery potential.
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| Inspection Report# : 2000008(pdf)
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| Significance: Aug 21, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow procedures associated with feed water level control system surveillance testing.
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| Human performance and corrective action deficiencies contributed to a Non-Cited Violation of Technical Specification 5.4.1 for failing to follow procedures. This led to the unplanned automatic reactor shutdown on July 24, 2001. The finding was of very low safety significance because no complications occurred during the unplanned automatic reactor shut down and the finding did not increase the likelihood of mitigation equipment being unavailable.
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| Inspection Report# : 2001010(pdf)
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| Mitigating Systems Significance: Jan 26, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct longstanding Reactor Core Isolation Cooling (RCIC) System valve degradation Corrective actions for a longstanding deficiency with the Reactor Core Isolation Cooling (RCIC) system steam bypass valve were not effective in stopping the leakage past the valve. This finding was determined to be a Non-Cited Violation of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action." This finding was determined to have very low risk significance because the degraded condition of the valve did not affect the operability of the RCIC system.
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| Inspection Report# : 2001002(pdf)
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| Significance: Jan 26, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow Condition Report process for Shutdown Service Water (SX) pipe wall thinning Corrective actions were not implemented to replace a portion of the shutdown service water (SX) system piping after pipe wall thinning was identified. The failure to take the specified corrective actions by the committed due date or to properly reevaluate the degraded condition was determined to be a Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, "Procedures." This finding was determined to have very low safety significance because the SX system remained operable and capable of performing its' safety function.
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| 1Q/2001 Inspection Findings - Clinton Page 3 of 8 Inspection Report# : 2001002(pdf)
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| Significance: Sep 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation Procedural requirements were not followed when unexpected equipment response was encountered.
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| Maintenance personnel failed to appropriately follow procedure instructions during testing of the Division III emergency diesel generator room fire detection system. These actions led to the emergency diesel generator being rendered inoperable. The procedure violation was treated as a Non-Cited Violation. This issue was of very low safety significance since the other divisional emergency diesel generators and all emergency core cooling systems were operable at the time of discovery.
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| Inspection Report# : 2000015(pdf)
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| Significance: Jun 16, 2000 Identified By: NRC Item Type: NCV NonCited Violation The licensee failed to ensure that appropriate post-modification testing was specified and accomplished for the Division I and Division III EDG output breaker circuitry modifications The licensee failed to ensure that the appropriate post-modification testing (PMT) was specified in the Division I and Division III emergency diesel generator (EDG) output breaker circuitry modification packages and that the post-modification tests were correctly accomplished. This was required to demonstrate through component and functional testing that the modified (rewired) portions of the Division I and Division III EDG output breaker circuitry were adequately installed to accomplish the intent of the plant design changes.
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| Inspection Report# : 2000012(pdf)
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| Significance: N/A Dec 29, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW 10 CFR 55.59(c)(5) REQUIREMENTS FOR RETAINING LICENSED OPERATOR REQUALIFICATION PROGRAM RECORDS The inspectors identified a Non-Cited Violation wherein the facility licensee had failed to follow the Code of Federal Regulations (CFR) Title 10, Part 55.59(c)(5), Records, requirements by failing to systematically retain all of the original or authenticated copies of the original evaluation documents during the year 2000 annual NRC examination. The finding was of very low safety significance because although the records were not the original or authenticated copies of the original, records did exist in computerized clerically transcribed documents. The computer records had not been signed, and there was no indication that they had been verified correct by the original authors. The unauthenticated documents did provide information that licensed operators, for the most part, had participated and were evaluated during the year 2000 NRC annual requalification examination. However, the inspectors determined that the finding was more than minor. Specifically, the inspectors identified at least one instance in which the transcribed information appeared to be incorrect or missing. The records failure had credible impact on safety, in that, it negatively impacted on the intent of the licensed operator requalification examination process which, in part, is to maintain a high level of confidence that licensed operators continue to possess the requisite knowledge and abilities needed to safely perform licensed duties. In addition, inadequate records keeping adversely affects the NRC's ability to regulate.
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| Inspection Report# : 2001015(pdf)
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| Barrier Integrity Significance: Nov 16, 2000 Identified By: NRC Item Type: NCV NonCited Violation Secondary containment was inoperable for 6 minutes during fuel movements when interlock doors were opened.
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| Secondary containment was inoperable for 6 minutes during fuel movements when secondary containment interlock doors were inadvertently opened to move scaffolding. The inoperability was discovered when operators in the control room received an alarm indicating a loss of secondary containment vacuum. One Non-Cited Violation was identified for violating Technical Specification 3.6.4.1 which requires secondary containment operability during fuel moves. This finding was evaluated using the shutdown significance determination process contained in Appendix G of IMC 0609 and was determined to have very low risk significance because it did not meet the criteria for findings requiring a phase 2 significance evaluation.
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| Inspection Report# : 2000017(pdf)
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| 1Q/2001 Inspection Findings - Clinton Page 4 of 8 Significance: Nov 14, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform radiographic examinations of Class 2 welds.
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| The inspectors identified a Non-Cited Violation for the failure to perform radiographic examinations of Class 2 welds in compliance with applicable American Society of Mechanical Engineers (ASME) Code requirements. During installation testing of the 1999 Feedwater Keep Fill FW-39 modification, five radiographic examinations had recorded geometric unsharpness values which exceeded Section III and Section V ASME Code limits. Radiographic geometric unsharpness values are used to ensure that the film is of adequate quality to see defects. In addition, inspectors identified that three examinations did not meet Section V Code requirements for documentation of radiographic technique variables which can affect the image quality of the film. The safety significance of this issue was considered very low at this time, based on the absence of adverse consequences, the presence of other image quality indicators, and because the issue did not involve the system isolation valves. The failure to comply with ASME Code radiographic examination requirements could result in the failure to detect flaws within reactor coolant boundary piping, and was considered a Non-Cited Violation of 10 CFR Part 50.55a, "Codes and Standards".
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| Inspection Report# : 2000019(pdf)
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| Emergency Preparedness Significance: Feb 23, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to follow emergency plan for on-shift staffing For an approximate 2-month time period, the licensee failed to meet one of the minimum on-shift emergency response organization (ERO) staffing requirements contained in Table 2-1 of the licensee's emergency plan.
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| Inspection Report# : 2001003(pdf)
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| Significance: N/A Apr 28, 2000 Identified By: NRC Item Type: FIN Finding Emergency Preparedness Performance Indicator Verification Alert and Notification System, Drill & Exercise Participation, and Drill & Exercise performance indicators: The inspectors verified that the licensee had acceptably gathered information and reported these three performance indicators, which were in the green band, with the following minor exception. The inspectors identified a discrepancy with the licensee's initial assessment of the Drill and Exercise Performance (DEP) indicator related to the number of performance opportunities associated with a General Emergency declaration during a drill or an exercise. The licensee initially assumed that only three performance opportunities would exist rather than four as provided in NEI 99-02, but later recognized that they had misinterpreted the guidance. This did not affect the DEP performance indicator which was in the green band.
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| Inspection Report# : 2000009(pdf)
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| Significance: Aug 21, 2001 Identified By: NRC Item Type: NCV NonCited Violation Violation of 10 CFR 50.54(q) re. SCBA qualifications A Non-Cited Violation of 10 CFR 50.54(q) was identified by the NRC associated with the failure to maintain personnel qualifications for self contained breathing apparatus in accordance with the Clinton Power Station Emergency Plan. The finding was of very low safety significance because the licensee maintained an adequate number of qualified personnel to maintain minimum coverage of the required positions identified in the Emergency Plan.
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| Inspection Report# : 2001010(pdf)
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| Significance: Jun 08, 2001 Identified By: NRC Item Type: VIO Violation Supplemental Inspection -- Failure to correct self-identified defficiencies disclosed through control room communications drills This supplemental inspection was performed by the NRC to assess the licensee's evaluation associated with inaccuracies in the reporting of the Drill and Exercise Performance (DEP) performance indicator and with the performance deficiencies that resulted in a White DEP performance indicator (fourth quarter 1999 through the fourth quarter 2000). During the inspection, performed in accordance with NRC Inspection Procedure
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| 1Q/2001 Inspection Findings - Clinton Page 5 of 8 95001, the inspector concluded that the licensee performed an adequate evaluation to determine the causes of both issues. In the case of the performance indicator errors, the licensee performed a root cause evaluation which identified a personnel error that was compounded by the lack of self-checking and verification. In addition, the licensee identified contributing causes that included the failure to provide adequate training to the emergency preparedness staff and the failure to provide adequate procedural guidance to the performance indicator data stewards and verifiers, which also applied to performance indicators in other cornerstones. The inspector concluded that the scope of corrective actions planned and implemented by the licensee appeared to address the identified causes. However, the inspector observed an additional discrepancy in the recently completed performance indicator evaluation related to drill and exercise participation. In addition, the licensee identified an error in its evaluation of one of the other emergency preparedness performance indicators that was not detected during its evaluation. These observations demonstrated weaknesses in the licensee's corrective actions and extent of condition review. The errors in the licensee's reporting of the DEP performance indicator was significant, in that the error resulted in a change of color, (i.e., Green-to-White). Consequently, a violation of 10 CFR 50.9 of more than minor safety significance was identified. Since the inaccurate reporting occurred during the period that the NRC's Enforcement Policy afforded discretion for the non-willful submittal of inaccurate performance indicator information, the NRC is exercising enforcement discretion and not citing the violation. In the case of the White DEP performance indicator, the inspector concluded that the licensee adequately assessed the deficiencies that led to the performance issues. Based on its review, the licensee attributed the White performance indicator to the high failure rate of control room communicator drills (i.e., job performance measures). The licensee identified two apparent causes for the high failure rate: (1) weaknesses in formal training; and (2) failure to meet emergency preparedness management expectations concerning the identification and correction of drill deficiencies. The inspector reviewed the licensee's corrective actions and determined that they addressed the causes identified. As a result of the licensee's immediate corrective actions, the licensee's performance returned the performance indicator to the Green band. The inspector and the licensee concluded that the high failure rate of the control room communicators resulted, in part, from inadequate corrective actions for self-identified deficiencies. Specifically, the licensee control room communicator drills were a portion of an overall annual evaluation of non-licensed operators, which included non-emergency preparedness functions. Generally, the failure of the communications segment of the evaluation did not result in a total failure of the annual evaluation. Therefore, the licensee's remedial actions were limited and were not effective in correcting the deficiencies and preventing similar failures from occurring, as required by 10 CFR 50.47(b)(14). By letter dated 08/22/01, the NRC concluded that a violation of 10 CFR 50.47(b)(14) had occurred and using the NRC's significance determination process, determined that the finding was white.
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| Inspection Report# : 2001009(pdf)
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| Occupational Radiation Safety Significance: Oct 25, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Three individuals entered a HRA in violation of Technical Specification 5.7.1 On October 25, 2000, three individuals entered the B residual heat removal heat exchanger room (a posted high radiation area); however, the individuals were not working under a radiation work permit that allowed entry into the high radiation area and did not satisfy either of the three entry conditions of Technical Specification 5.7.1.
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| Inspection Report# : 2000018(pdf)
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| Significance: Oct 08, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an Adequate Survey to identify and to post a High Radiation Area A finding and associated Non-Cited Violation was identified concerning the failure to perform an adequate radiological survey, as required by 10 CFR 20.1501. Although the licensee identified this issue, the licensee did not thoroughly evaluate the cause(s) of the unanticipated radiological conditions and associated problems in the monitoring of radioactive waste activities, which have resulted in previous, similar incidents. The finding was of very low safety significance because the area radiation levels and the licensee's additional administrative barriers would have limited the potential for an individual inadvertently entering the area and receiving a radiation exposure in excess of regulatory limits.
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| Inspection Report# : 2001015(pdf)
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| Significance: Aug 21, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to maintain administrative control of high radiation area keys as required by Technical Specification 5.7.2 Technical Specification 5.7.2 requires, in part, that doors to high radiation areas in which an individual could receive a deep dose equivalent greater than or equal to 1000 millirem in one hour (at 30 centimeters) shall be provided with locked or continuously guarded doors to prevent unauthorized entry and that the keys to such doors shall be administratively controlled. During October 29 - 31, 2001, the licensee failed to maintain administrative control of a key that controlled five access points to high radiation areas specified above (i.e., lost the key and failed to perform required key inventories to identify its loss), as described in CR No. 2-00-11-016. Since the inspector concluded that sufficient barriers remained to
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| 1Q/2001 Inspection Findings - Clinton Page 6 of 8 prevent an unauthorized individual from entering the affected areas and receiving an overexposure, the inspector concluded that the incident was of very low safety significance. The licensee also reported the incident to the NRC as an occurrence for the Occupational Exposure Control Effectiveness performance indicator. This is being treated as a Non-Cited Violation.
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| Inspection Report# : 2001010(pdf)
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| Significance: SL-IV Jul 27, 2001 Identified By: NRC Item Type: NCV NonCited Violation Misuse of Radioactive Material to Alarm a PCM Radiation protection technician used contaminated material to alarm a portal contamination monitor (PCM), while an individual was performing a contamination survey. Based on the licensee's investigation, the contamination was not placed on the individual, and the individual successfully monitored through an additional PCM. This incident will be reviewed by the NRC for potential enforcement actions. Update: On July 27, 2001, the NRC identified and forwarded to the licensee (by letter) a Non-Cited Violation of the Clinton Station Facility Operating License associated with the deliberate misuse of radioactive material by a junior contract radiation protection technician. On October 20, 2000, the technician misused radioactive material to cause an erroneous alarm on a PCM, as another individual was performing a contamination survey. The licensee identified the incident, entered the incident into its corrective action program, and implemented immediate corrective actions. Since the violation was determined to be willful, the NRC did not assign a significance to the violation using the NRC's Significance Determination Process. In accordance with the NRC Enforcement Policy, the NRC determined that the incident constituted a Severity Level IV violation of the Clinton Power Station Facility Operating License. Further, the NRC determined that the violation met the criteria necessary to disposition the violation as a Non-Cited Violation (Section VI.A.1.d of the NRC Enforcement Policy).
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| Inspection Report# : 2001010(pdf)
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| Inspection Report# : 2000018(pdf)
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| Public Radiation Safety Significance: Dec 08, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Inadvertent Release of Radioactive Material to Unrestricted Area During September 2000, the licensee conducted a survey of tools, equipment, etc. outside of the restricted area (protected area and owner controlled area) and identified low-level contaminated materials that were not under constant surveillance or control. The failure to maintain contstant surveillance and control of the material was a violation of 10 CFR 20.1802 and was characterized as a Non-Cited Violation. Based on the licensee's conservative annual dose assessment (about 1.56 millirem) and the inability to define the origin of each of the items, the inspector concluded that the issue constituted one occurrence/event per the NRC Significance Determination Process (Green).
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| Inspection Report# : 2000021(pdf)
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| Physical Protection Miscellaneous Significance: N/A Jan 26, 2001 Identified By: NRC Item Type: FIN Finding Assessment of Problem Identification and Resolution Performance The team identified that the licensee appropriately entered significant plant issues into the corrective action process by initiating condition reports.
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| Some less significant conditions adverse to quality were evaluated and corrected outside the established process. The trending program was not fully effective as a problem identification tool. Quality Assurance audits and self-assessments reviewed varied in quality. Identified issues were generally evaluated properly, although in several cases the corrective action process did not work effectively to either evaluate or prioritize issues.
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| Current station performance issues including human performance, corrective action program, surveillance testing, and labeling indicate that long term corrective actions previously taken in these areas as restart and post-restart initiatives have not been fully effective to support sustained improvement. Corrective actions were not always fully effective or timely for some individual equipment issues and the effectiveness review process (CARE) did not always identify ineffective corrective actions. The licensee had recently recognized similar deficiencies in corrective action program implementation but had not yet fully developed or completed the corrective actions to improve these areas. The inspectors did not find any reluctance by the station employees to raise safety issues.
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| 1Q/2001 Inspection Findings - Clinton Page 7 of 8 Inspection Report# : 2001002(pdf)
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| Significance: N/A Dec 31, 2000 Identified By: NRC Item Type: FIN Finding Recent human performance issues have occurred which are associated with operator performance and knowledge based deficiencies that have affected plant operations and responses to transient conditions.
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| Recent human performance issues have occurred which are associated with operator performance and knowledge based deficiencies that have affected plant operations and responses to transient conditions. While the risk of the individual events was very low, the failure of operators to adequately control level parameters indicated a declining trend in this area. These issues could not be easily evaluated by present risk analysis methods because failures to follow procedures and maintaining management expectations were not modeled in the Clinton Individual Plant Evaluation. Therefore, the finding is characterized as having no color.
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| Inspection Report# : 2000020(pdf)
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| Significance: N/A Nov 14, 2000 Identified By: NRC Item Type: FIN Finding Three procedures were not written in compliance with the applicable ASME Code.
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| The inspectors reviewed three special process procedures, and identified areas where all three procedures were not written in compliance with the applicable ASME Code. The procedure deficiencies had the potential to affect the ASME Code compliance of weld fabrication and nondestructive examination used on safety-related components and piping. The inspectors noted that each of the ASME Code problems identified contained elements of human performance deficiencies. The human performance aspects, while not always being the root cause of the problem, were significant contributors leading to procedure deficiencies. While the risk of the individual examples was very low, the number of deficiencies indicated a problem with incorporation of applicable ASME Code requirements into special process procedures.
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| Inspection Report# : 2000019(pdf)
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| Significance: N/A Sep 30, 2000 Identified By: NRC Item Type: FIN Finding Recent events affecting plant operations contained elements of human performance deficiencies.
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| NO COLOR. The inspectors noted that several recent events which have affected plant operations and the operability of safety-related components or other components important to safety contained elements of human performance deficiencies. The human performance aspects, while not always being the root cause of the problem, were significant contributors leading to the events. While the risk of the individual events was very low, the number of maintenance-related incidents indicated a problem exists with the control, review, and performance of maintenance activities.
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| Inspection Report# : 2000015(pdf)
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| Significance: N/A May 20, 2000 Identified By: NRC Item Type: FIN Finding Inaccurate historical data for the Safety System Functional Failure Indicator No Color. The licensee identified a failure to submit accurate information to the NRC. The inaccurate information involved the historical data submittal for the Safety System Functional Failure Performance Indicator. The error resulted in a response band color change from Green to White for the first quarter 1999 Performance Indicator. The NRC exercised Enforcement Discretion pursuant to Section VII.B.6 of the Enforcement Policy and did not cite the violation.
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| Inspection Report# : 2000008(pdf)
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| Significance: Feb 17, 2002 Identified By: NRC Item Type: NCV NonCited Violation Non-cited violation of T.S. 5.4.1 for an inadequate surveillance procedure.
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| Inspection Report# : 2001016(pdf)
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| Significance: Feb 17, 2002 Identified By: NRC Item Type: FIN Finding A temporary modification on the "A" RR FCV control cirucuitry.
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| On December 14, 2001, the licensee installed a temporary modification on the "A" RR FCV control circuitry. The T-mod was installed to assist the operators in manually controlling the "A" RR FCV because the reliability of the normal control circuitry was in question. During the implementation portion of the T-mod installations, the "A" RR FCV unexpectedly moved from 94 percent open to 102 percent open at which point the protective position circuitry locked the valve at the 102 percent position. Recator power was observed to go from 94 percent to 98 percent during this
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| 1Q/2001 Inspection Findings - Clinton Page 8 of 8 unexpected valve movement. Following this unexpected FCV movement, opeerations personnel ordered the T-mod to be removed and operators then proceeded to manually shut down the reactor without any further movements of the "A" RR FCV.
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| Inspection Report# : 2001016(pdf)
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| Significance: SL-IV Aug 18, 2001 Identified By: NRC Item Type: VIO Violation Falsification of Test Records by Licensee Employee SL IV - On July 2, 2001, by separate letter, NRC issued a Severity Level IV violation of 10 CFR 50.9 for a deliberate falsification by a plant test engineer. Following investigation by the Office of Investigations, NRC determined that, on October 20, 2000, a test engineer forged another employee's signature on two test package cover sheets on by forging another employee's signature without his prior concurrence, in violation of Clinton established plant protocol and procedure.
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| Inspection Report# : 2001010(pdf)
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| Significance: SL-IV Apr 06, 2001 Identified By: NRC Item Type: VIO Violation Violation of 10 CFR 50.7 "Employee Protection" On April 6, 2001, the NRC issued the licensee a Severity Level IV Violation of 10 CFR 50.7. The NRC concluded that the licensee took adverse employment actions against an employee in the licensee's Nuclear Training Department (i.e., unfavorable 1999 performance review), in part, as a result of the employee's engagement in protected activities. In addition, the NRC learned that several training personnel may be reluctant to discuss department issues within the nuclear training department.
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| Inspection Report# : 2001010(pdf)
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| Inspection Report# : 2001007(pdf)
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| Last modified : March 28, 2002
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| 2Q/2001 Inspection Findings - Clinton Page 1 of 8 Clinton Initiating Events Significance: Dec 31, 2000 Identified By: NRC Item Type: FIN Finding Operators did not adequately control reactor vessel inventory after a reactor scram which resulted in the motor driven reactor feedwater pump tripping on high reactor vessel water level.
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| During operator response to a reactor scram on December 18, 2000, operators did not adequately control reactor vessel inventory prior to the motor driven reactor feedwater pump tripping on high reactor vessel water level. The inspectors reviewed this issue using the significance determination process for a transient with a loss of feedwater and determined this was of very low safety significance because all other reactor vessel level control systems were operable and functioned as designed.
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| Inspection Report# : 2000020(pdf)
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| Significance: Dec 31, 2000 Identified By: NRC Item Type: NCV NonCited Violation Operators failed to adequately control reactor vessel water level and pressure following the automatic reactor scram which resulted in a second automatic scram.
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| Operators failed to adequately control reactor vessel water level and pressure, while attempting to open the main steam isolation valves following the automatic reactor scram on December 18, 2000. This resulted in an automatic scram signal due to low reactor vessel water level. The inspectors reviewed this issue using the significance determination process for a transient with a loss of feedwater and determined this was of very low safety significance because the event occurred while the reactor was shut down and all control rods were already fully inserted.
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| Inspection Report# : 2000020(pdf)
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| Significance: Dec 31, 2000 Identified By: NRC Item Type: NCV NonCited Violation Operators failed to adequately evaluate an alarming moisture separator drain tank level annunciator that resulted in a turbine trip.
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| During plant restart following refueling outage 7, operators did not adequately evaluate an alarming moisture separator drain tank level annunciator.
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| As a result, high water level in the moisture separator drain tank caused a turbine trip with the reactor at approximately 25% power. The inspectors reviewed this issue using the significance determination process for a transient. Since only the initiating event cornerstone is affected and associated assumptions have no other impact than slightly increasing the likelihood of an uncomplicated reactor trip, the finding is considered to be of very low safety significance.
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| Inspection Report# : 2000020(pdf)
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| Significance: Nov 16, 2000 Identified By: NRC Item Type: NCV NonCited Violation An alternate rod insertion system initiation and a manual reactor scram occurred with the reactor shutdown as a result of an inadequate maintenance procedure.
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| During replacement of power supplies for the alternate rod insertion (ARI) system, maintenance personnel failed to fully evaluate the impacts that re-energizing the power supplies had on the ARI initiation logic. While re-energizing the power supplies, the initiation logic sensed an ARI signal (low reactor water level). This caused the vent and drain valves to close and the scram discharge volume to fill with water. Plant operators inserted a manual scram signal before the automatic high scram discharge volume set point was reached. One Non-Cited Violation was identified for having an inadequate maintenance procedure to control this activity. This finding was evaluated using the shutdown significance determination process contained in Appendix G of IMC 0609 and was determined to have very low risk significance because it did not impact any of the five shutdown safety functions identified by NUMARC 91-06.
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| Inspection Report# : 2000017(pdf)
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| 2Q/2001 Inspection Findings - Clinton Page 2 of 8 Significance: Sep 30, 2000 Identified By: NRC Item Type: FIN Finding Human performance errors and an inadequate procedure resulted in exceeding the allowed outage time for the emergency reserve auxiliary transformer static VAR compensator.
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| Human performance errors and the failure to develop an adequate procedure for the emergency reserve auxiliary transformer static VAR (Volt Ampere Reactive) compensator (ERAT-SVC) surveillance test resulted in several delays during the test. These delays caused the work to not be completed within the allowed outage time. Therefore, a request for Enforcement Discretion was presented to the NRC which was formally granted on September 20, 2000 (NOED 00-6-011). The safety significance of this finding was very low because all other emergency core cooling system trains (automatic depressurization system, low pressure core spray, and low pressure core injection), emergency diesel generators, and the reactor core isolation cooling system were operable.
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| Inspection Report# : 2000015(pdf)
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| Significance: Aug 21, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow procedures associated with feed water level control system surveillance testing.
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| Human performance and corrective action deficiencies contributed to a Non-Cited Violation of Technical Specification 5.4.1 for failing to follow procedures. This led to the unplanned automatic reactor shutdown on July 24, 2001. The finding was of very low safety significance because no complications occurred during the unplanned automatic reactor shut down and the finding did not increase the likelihood of mitigation equipment being unavailable.
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| Inspection Report# : 2001010(pdf)
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| Significance: May 17, 2000 Identified By: Self Disclosing Item Type: FIN Finding Manual reactor shutdown A labeling discrepancy contributed to the improper isolation of a protective relay for the 4.16kV Bus 1B Reserve Feed Breaker. As a result, during functional testing, the relay actuated and caused the bus to be de-energized which ultimately resulted in a manual reactor shut down. This issue was determined to be of very low risk significance due to remaining mitigation capability and recovery potential.
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| Inspection Report# : 2000008(pdf)
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| Mitigating Systems Significance: Jan 26, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow Condition Report process for Shutdown Service Water (SX) pipe wall thinning Corrective actions were not implemented to replace a portion of the shutdown service water (SX) system piping after pipe wall thinning was identified. The failure to take the specified corrective actions by the committed due date or to properly reevaluate the degraded condition was determined to be a Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, "Procedures." This finding was determined to have very low safety significance because the SX system remained operable and capable of performing its' safety function.
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| Inspection Report# : 2001002(pdf)
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| Significance: Jan 26, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct longstanding Reactor Core Isolation Cooling (RCIC) System valve degradation Corrective actions for a longstanding deficiency with the Reactor Core Isolation Cooling (RCIC) system steam bypass valve were not effective in stopping the leakage past the valve. This finding was determined to be a Non-Cited Violation of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action." This finding was determined to have very low risk significance because the degraded condition of the valve did not affect the operability of the RCIC system.
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| 2Q/2001 Inspection Findings - Clinton Page 3 of 8 Inspection Report# : 2001002(pdf)
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| Significance: Sep 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation Procedural requirements were not followed when unexpected equipment response was encountered.
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| Maintenance personnel failed to appropriately follow procedure instructions during testing of the Division III emergency diesel generator room fire detection system. These actions led to the emergency diesel generator being rendered inoperable. The procedure violation was treated as a Non-Cited Violation. This issue was of very low safety significance since the other divisional emergency diesel generators and all emergency core cooling systems were operable at the time of discovery.
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| Inspection Report# : 2000015(pdf)
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| Significance: N/A Dec 29, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW 10 CFR 55.59(c)(5) REQUIREMENTS FOR RETAINING LICENSED OPERATOR REQUALIFICATION PROGRAM RECORDS The inspectors identified a Non-Cited Violation wherein the facility licensee had failed to follow the Code of Federal Regulations (CFR) Title 10, Part 55.59(c)(5), Records, requirements by failing to systematically retain all of the original or authenticated copies of the original evaluation documents during the year 2000 annual NRC examination. The finding was of very low safety significance because although the records were not the original or authenticated copies of the original, records did exist in computerized clerically transcribed documents. The computer records had not been signed, and there was no indication that they had been verified correct by the original authors. The unauthenticated documents did provide information that licensed operators, for the most part, had participated and were evaluated during the year 2000 NRC annual requalification examination. However, the inspectors determined that the finding was more than minor. Specifically, the inspectors identified at least one instance in which the transcribed information appeared to be incorrect or missing. The records failure had credible impact on safety, in that, it negatively impacted on the intent of the licensed operator requalification examination process which, in part, is to maintain a high level of confidence that licensed operators continue to possess the requisite knowledge and abilities needed to safely perform licensed duties. In addition, inadequate records keeping adversely affects the NRC's ability to regulate.
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| Inspection Report# : 2001015(pdf)
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| Significance: Jun 16, 2000 Identified By: NRC Item Type: NCV NonCited Violation The licensee failed to ensure that appropriate post-modification testing was specified and accomplished for the Division I and Division III EDG output breaker circuitry modifications The licensee failed to ensure that the appropriate post-modification testing (PMT) was specified in the Division I and Division III emergency diesel generator (EDG) output breaker circuitry modification packages and that the post-modification tests were correctly accomplished. This was required to demonstrate through component and functional testing that the modified (rewired) portions of the Division I and Division III EDG output breaker circuitry were adequately installed to accomplish the intent of the plant design changes.
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| Inspection Report# : 2000012(pdf)
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| Barrier Integrity Significance: Nov 16, 2000 Identified By: NRC Item Type: NCV NonCited Violation Secondary containment was inoperable for 6 minutes during fuel movements when interlock doors were opened.
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| Secondary containment was inoperable for 6 minutes during fuel movements when secondary containment interlock doors were inadvertently opened to move scaffolding. The inoperability was discovered when operators in the control room received an alarm indicating a loss of secondary containment vacuum. One Non-Cited Violation was identified for violating Technical Specification 3.6.4.1 which requires secondary containment operability during fuel moves. This finding was evaluated using the shutdown significance determination process contained in Appendix G of IMC 0609 and was determined to have very low risk significance because it did not meet the criteria for findings requiring a phase 2 significance evaluation.
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| Inspection Report# : 2000017(pdf)
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| 2Q/2001 Inspection Findings - Clinton Page 4 of 8 Significance: Nov 14, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform radiographic examinations of Class 2 welds.
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| The inspectors identified a Non-Cited Violation for the failure to perform radiographic examinations of Class 2 welds in compliance with applicable American Society of Mechanical Engineers (ASME) Code requirements. During installation testing of the 1999 Feedwater Keep Fill FW-39 modification, five radiographic examinations had recorded geometric unsharpness values which exceeded Section III and Section V ASME Code limits. Radiographic geometric unsharpness values are used to ensure that the film is of adequate quality to see defects. In addition, inspectors identified that three examinations did not meet Section V Code requirements for documentation of radiographic technique variables which can affect the image quality of the film. The safety significance of this issue was considered very low at this time, based on the absence of adverse consequences, the presence of other image quality indicators, and because the issue did not involve the system isolation valves. The failure to comply with ASME Code radiographic examination requirements could result in the failure to detect flaws within reactor coolant boundary piping, and was considered a Non-Cited Violation of 10 CFR Part 50.55a, "Codes and Standards".
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| Inspection Report# : 2000019(pdf)
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| Emergency Preparedness Significance: Jun 08, 2001 Identified By: NRC Item Type: VIO Violation Supplemental Inspection -- Failure to correct self-identified defficiencies disclosed through control room communications drills This supplemental inspection was performed by the NRC to assess the licensee's evaluation associated with inaccuracies in the reporting of the Drill and Exercise Performance (DEP) performance indicator and with the performance deficiencies that resulted in a White DEP performance indicator (fourth quarter 1999 through the fourth quarter 2000). During the inspection, performed in accordance with NRC Inspection Procedure 95001, the inspector concluded that the licensee performed an adequate evaluation to determine the causes of both issues. In the case of the performance indicator errors, the licensee performed a root cause evaluation which identified a personnel error that was compounded by the lack of self-checking and verification. In addition, the licensee identified contributing causes that included the failure to provide adequate training to the emergency preparedness staff and the failure to provide adequate procedural guidance to the performance indicator data stewards and verifiers, which also applied to performance indicators in other cornerstones. The inspector concluded that the scope of corrective actions planned and implemented by the licensee appeared to address the identified causes. However, the inspector observed an additional discrepancy in the recently completed performance indicator evaluation related to drill and exercise participation. In addition, the licensee identified an error in its evaluation of one of the other emergency preparedness performance indicators that was not detected during its evaluation. These observations demonstrated weaknesses in the licensee's corrective actions and extent of condition review. The errors in the licensee's reporting of the DEP performance indicator was significant, in that the error resulted in a change of color, (i.e., Green-to-White). Consequently, a violation of 10 CFR 50.9 of more than minor safety significance was identified. Since the inaccurate reporting occurred during the period that the NRC's Enforcement Policy afforded discretion for the non-willful submittal of inaccurate performance indicator information, the NRC is exercising enforcement discretion and not citing the violation. In the case of the White DEP performance indicator, the inspector concluded that the licensee adequately assessed the deficiencies that led to the performance issues. Based on its review, the licensee attributed the White performance indicator to the high failure rate of control room communicator drills (i.e., job performance measures). The licensee identified two apparent causes for the high failure rate: (1) weaknesses in formal training; and (2) failure to meet emergency preparedness management expectations concerning the identification and correction of drill deficiencies. The inspector reviewed the licensee's corrective actions and determined that they addressed the causes identified. As a result of the licensee's immediate corrective actions, the licensee's performance returned the performance indicator to the Green band. The inspector and the licensee concluded that the high failure rate of the control room communicators resulted, in part, from inadequate corrective actions for self-identified deficiencies. Specifically, the licensee control room communicator drills were a portion of an overall annual evaluation of non-licensed operators, which included non-emergency preparedness functions. Generally, the failure of the communications segment of the evaluation did not result in a total failure of the annual evaluation. Therefore, the licensee's remedial actions were limited and were not effective in correcting the deficiencies and preventing similar failures from occurring, as required by 10 CFR 50.47(b)(14). By letter dated 08/22/01, the NRC concluded that a violation of 10 CFR 50.47(b)(14) had occurred and using the NRC's significance determination process, determined that the finding was white.
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| Inspection Report# : 2001009(pdf)
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| Significance: Feb 23, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to follow emergency plan for on-shift staffing For an approximate 2-month time period, the licensee failed to meet one of the minimum on-shift emergency response organization (ERO) staffing requirements contained in Table 2-1 of the licensee's emergency plan.
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| Inspection Report# : 2001003(pdf)
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| 2Q/2001 Inspection Findings - Clinton Page 5 of 8 Significance: Aug 21, 2001 Identified By: NRC Item Type: NCV NonCited Violation Violation of 10 CFR 50.54(q) re. SCBA qualifications A Non-Cited Violation of 10 CFR 50.54(q) was identified by the NRC associated with the failure to maintain personnel qualifications for self contained breathing apparatus in accordance with the Clinton Power Station Emergency Plan. The finding was of very low safety significance because the licensee maintained an adequate number of qualified personnel to maintain minimum coverage of the required positions identified in the Emergency Plan.
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| Inspection Report# : 2001010(pdf)
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| Significance: N/A Apr 28, 2000 Identified By: NRC Item Type: FIN Finding Emergency Preparedness Performance Indicator Verification Alert and Notification System, Drill & Exercise Participation, and Drill & Exercise performance indicators: The inspectors verified that the licensee had acceptably gathered information and reported these three performance indicators, which were in the green band, with the following minor exception. The inspectors identified a discrepancy with the licensee's initial assessment of the Drill and Exercise Performance (DEP) indicator related to the number of performance opportunities associated with a General Emergency declaration during a drill or an exercise. The licensee initially assumed that only three performance opportunities would exist rather than four as provided in NEI 99-02, but later recognized that they had misinterpreted the guidance. This did not affect the DEP performance indicator which was in the green band.
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| Inspection Report# : 2000009(pdf)
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| Occupational Radiation Safety Significance: Oct 25, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Three individuals entered a HRA in violation of Technical Specification 5.7.1 On October 25, 2000, three individuals entered the B residual heat removal heat exchanger room (a posted high radiation area); however, the individuals were not working under a radiation work permit that allowed entry into the high radiation area and did not satisfy either of the three entry conditions of Technical Specification 5.7.1.
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| Inspection Report# : 2000018(pdf)
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| Significance: Oct 08, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an Adequate Survey to identify and to post a High Radiation Area A finding and associated Non-Cited Violation was identified concerning the failure to perform an adequate radiological survey, as required by 10 CFR 20.1501. Although the licensee identified this issue, the licensee did not thoroughly evaluate the cause(s) of the unanticipated radiological conditions and associated problems in the monitoring of radioactive waste activities, which have resulted in previous, similar incidents. The finding was of very low safety significance because the area radiation levels and the licensee's additional administrative barriers would have limited the potential for an individual inadvertently entering the area and receiving a radiation exposure in excess of regulatory limits.
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| Inspection Report# : 2001015(pdf)
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| Significance: Aug 21, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to maintain administrative control of high radiation area keys as required by Technical Specification 5.7.2 Technical Specification 5.7.2 requires, in part, that doors to high radiation areas in which an individual could receive a deep dose equivalent greater than or equal to 1000 millirem in one hour (at 30 centimeters) shall be provided with locked or continuously guarded doors to prevent unauthorized entry and that the keys to such doors shall be administratively controlled. During October 29 - 31, 2001, the licensee failed to maintain administrative control of a key that controlled five access points to high radiation areas specified above (i.e., lost the key and failed to perform required key inventories to identify its loss), as described in CR No. 2-00-11-016. Since the inspector concluded that sufficient barriers remained to prevent an unauthorized individual from entering the affected areas and receiving an overexposure, the inspector concluded that the incident was
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| 2Q/2001 Inspection Findings - Clinton Page 6 of 8 of very low safety significance. The licensee also reported the incident to the NRC as an occurrence for the Occupational Exposure Control Effectiveness performance indicator. This is being treated as a Non-Cited Violation.
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| Inspection Report# : 2001010(pdf)
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| Significance: SL-IV Jul 27, 2001 Identified By: NRC Item Type: NCV NonCited Violation Misuse of Radioactive Material to Alarm a PCM Radiation protection technician used contaminated material to alarm a portal contamination monitor (PCM), while an individual was performing a contamination survey. Based on the licensee's investigation, the contamination was not placed on the individual, and the individual successfully monitored through an additional PCM. This incident will be reviewed by the NRC for potential enforcement actions. Update: On July 27, 2001, the NRC identified and forwarded to the licensee (by letter) a Non-Cited Violation of the Clinton Station Facility Operating License associated with the deliberate misuse of radioactive material by a junior contract radiation protection technician. On October 20, 2000, the technician misused radioactive material to cause an erroneous alarm on a PCM, as another individual was performing a contamination survey. The licensee identified the incident, entered the incident into its corrective action program, and implemented immediate corrective actions. Since the violation was determined to be willful, the NRC did not assign a significance to the violation using the NRC's Significance Determination Process. In accordance with the NRC Enforcement Policy, the NRC determined that the incident constituted a Severity Level IV violation of the Clinton Power Station Facility Operating License. Further, the NRC determined that the violation met the criteria necessary to disposition the violation as a Non-Cited Violation (Section VI.A.1.d of the NRC Enforcement Policy).
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| Inspection Report# : 2000018(pdf)
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| Inspection Report# : 2001010(pdf)
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| Public Radiation Safety Significance: Dec 08, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Inadvertent Release of Radioactive Material to Unrestricted Area During September 2000, the licensee conducted a survey of tools, equipment, etc. outside of the restricted area (protected area and owner controlled area) and identified low-level contaminated materials that were not under constant surveillance or control. The failure to maintain contstant surveillance and control of the material was a violation of 10 CFR 20.1802 and was characterized as a Non-Cited Violation. Based on the licensee's conservative annual dose assessment (about 1.56 millirem) and the inability to define the origin of each of the items, the inspector concluded that the issue constituted one occurrence/event per the NRC Significance Determination Process (Green).
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| Inspection Report# : 2000021(pdf)
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| Physical Protection Miscellaneous Significance: SL-IV Apr 06, 2001 Identified By: NRC Item Type: VIO Violation Violation of 10 CFR 50.7 "Employee Protection" On April 6, 2001, the NRC issued the licensee a Severity Level IV Violation of 10 CFR 50.7. The NRC concluded that the licensee took adverse employment actions against an employee in the licensee's Nuclear Training Department (i.e., unfavorable 1999 performance review), in part, as a result of the employee's engagement in protected activities. In addition, the NRC learned that several training personnel may be reluctant to discuss department issues within the nuclear training department.
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| Inspection Report# : 2001010(pdf)
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| Inspection Report# : 2001007(pdf)
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| Significance: N/A Jan 26, 2001 Identified By: NRC Item Type: FIN Finding Assessment of Problem Identification and Resolution Performance
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| 2Q/2001 Inspection Findings - Clinton Page 7 of 8 The team identified that the licensee appropriately entered significant plant issues into the corrective action process by initiating condition reports.
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| Some less significant conditions adverse to quality were evaluated and corrected outside the established process. The trending program was not fully effective as a problem identification tool. Quality Assurance audits and self-assessments reviewed varied in quality. Identified issues were generally evaluated properly, although in several cases the corrective action process did not work effectively to either evaluate or prioritize issues.
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| Current station performance issues including human performance, corrective action program, surveillance testing, and labeling indicate that long term corrective actions previously taken in these areas as restart and post-restart initiatives have not been fully effective to support sustained improvement. Corrective actions were not always fully effective or timely for some individual equipment issues and the effectiveness review process (CARE) did not always identify ineffective corrective actions. The licensee had recently recognized similar deficiencies in corrective action program implementation but had not yet fully developed or completed the corrective actions to improve these areas. The inspectors did not find any reluctance by the station employees to raise safety issues.
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| Inspection Report# : 2001002(pdf)
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| Significance: N/A Dec 31, 2000 Identified By: NRC Item Type: FIN Finding Recent human performance issues have occurred which are associated with operator performance and knowledge based deficiencies that have affected plant operations and responses to transient conditions.
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| Recent human performance issues have occurred which are associated with operator performance and knowledge based deficiencies that have affected plant operations and responses to transient conditions. While the risk of the individual events was very low, the failure of operators to adequately control level parameters indicated a declining trend in this area. These issues could not be easily evaluated by present risk analysis methods because failures to follow procedures and maintaining management expectations were not modeled in the Clinton Individual Plant Evaluation. Therefore, the finding is characterized as having no color.
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| Inspection Report# : 2000020(pdf)
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| Significance: N/A Nov 14, 2000 Identified By: NRC Item Type: FIN Finding Three procedures were not written in compliance with the applicable ASME Code.
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| The inspectors reviewed three special process procedures, and identified areas where all three procedures were not written in compliance with the applicable ASME Code. The procedure deficiencies had the potential to affect the ASME Code compliance of weld fabrication and nondestructive examination used on safety-related components and piping. The inspectors noted that each of the ASME Code problems identified contained elements of human performance deficiencies. The human performance aspects, while not always being the root cause of the problem, were significant contributors leading to procedure deficiencies. While the risk of the individual examples was very low, the number of deficiencies indicated a problem with incorporation of applicable ASME Code requirements into special process procedures.
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| Inspection Report# : 2000019(pdf)
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| Significance: N/A Sep 30, 2000 Identified By: NRC Item Type: FIN Finding Recent events affecting plant operations contained elements of human performance deficiencies.
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| NO COLOR. The inspectors noted that several recent events which have affected plant operations and the operability of safety-related components or other components important to safety contained elements of human performance deficiencies. The human performance aspects, while not always being the root cause of the problem, were significant contributors leading to the events. While the risk of the individual events was very low, the number of maintenance-related incidents indicated a problem exists with the control, review, and performance of maintenance activities.
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| Inspection Report# : 2000015(pdf)
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| Significance: Feb 17, 2002 Identified By: NRC Item Type: NCV NonCited Violation Non-cited violation of T.S. 5.4.1 for an inadequate surveillance procedure.
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| Inspection Report# : 2001016(pdf)
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| Significance: Feb 17, 2002 Identified By: NRC Item Type: FIN Finding A temporary modification on the "A" RR FCV control cirucuitry.
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| On December 14, 2001, the licensee installed a temporary modification on the "A" RR FCV control circuitry. The T-mod was installed to assist the operators in manually controlling the "A" RR FCV because the reliability of the normal control circuitry was in question. During the implementation portion of the T-mod installations, the "A" RR FCV unexpectedly moved from 94 percent open to 102 percent open at which point the protective position circuitry locked the valve at the 102 percent position. Recator power was observed to go from 94 percent to 98 percent during this
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| 2Q/2001 Inspection Findings - Clinton Page 8 of 8 unexpected valve movement. Following this unexpected FCV movement, opeerations personnel ordered the T-mod to be removed and operators then proceeded to manually shut down the reactor without any further movements of the "A" RR FCV.
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| Inspection Report# : 2001016(pdf)
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| Significance: SL-IV Aug 18, 2001 Identified By: NRC Item Type: VIO Violation Falsification of Test Records by Licensee Employee SL IV - On July 2, 2001, by separate letter, NRC issued a Severity Level IV violation of 10 CFR 50.9 for a deliberate falsification by a plant test engineer. Following investigation by the Office of Investigations, NRC determined that, on October 20, 2000, a test engineer forged another employee's signature on two test package cover sheets on by forging another employee's signature without his prior concurrence, in violation of Clinton established plant protocol and procedure.
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| Inspection Report# : 2001010(pdf)
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| Significance: N/A May 20, 2000 Identified By: NRC Item Type: FIN Finding Inaccurate historical data for the Safety System Functional Failure Indicator No Color. The licensee identified a failure to submit accurate information to the NRC. The inaccurate information involved the historical data submittal for the Safety System Functional Failure Performance Indicator. The error resulted in a response band color change from Green to White for the first quarter 1999 Performance Indicator. The NRC exercised Enforcement Discretion pursuant to Section VII.B.6 of the Enforcement Policy and did not cite the violation.
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| Inspection Report# : 2000008(pdf)
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| Last modified : March 27, 2002
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| 3Q/2001 Inspection Findings - Clinton Page 1 of 8 Clinton Initiating Events Significance: Aug 21, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow procedures associated with feed water level control system surveillance testing.
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| Human performance and corrective action deficiencies contributed to a Non-Cited Violation of Technical Specification 5.4.1 for failing to follow procedures. This led to the unplanned automatic reactor shutdown on July 24, 2001. The finding was of very low safety significance because no complications occurred during the unplanned automatic reactor shut down and the finding did not increase the likelihood of mitigation equipment being unavailable.
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| Inspection Report# : 2001010(pdf)
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| Significance: Dec 31, 2000 Identified By: NRC Item Type: FIN Finding Operators did not adequately control reactor vessel inventory after a reactor scram which resulted in the motor driven reactor feedwater pump tripping on high reactor vessel water level.
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| During operator response to a reactor scram on December 18, 2000, operators did not adequately control reactor vessel inventory prior to the motor driven reactor feedwater pump tripping on high reactor vessel water level. The inspectors reviewed this issue using the significance determination process for a transient with a loss of feedwater and determined this was of very low safety significance because all other reactor vessel level control systems were operable and functioned as designed.
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| Inspection Report# : 2000020(pdf)
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| Significance: Dec 31, 2000 Identified By: NRC Item Type: NCV NonCited Violation Operators failed to adequately evaluate an alarming moisture separator drain tank level annunciator that resulted in a turbine trip.
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| During plant restart following refueling outage 7, operators did not adequately evaluate an alarming moisture separator drain tank level annunciator.
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| As a result, high water level in the moisture separator drain tank caused a turbine trip with the reactor at approximately 25% power. The inspectors reviewed this issue using the significance determination process for a transient. Since only the initiating event cornerstone is affected and associated assumptions have no other impact than slightly increasing the likelihood of an uncomplicated reactor trip, the finding is considered to be of very low safety significance.
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| Inspection Report# : 2000020(pdf)
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| Significance: Dec 31, 2000 Identified By: NRC Item Type: NCV NonCited Violation Operators failed to adequately control reactor vessel water level and pressure following the automatic reactor scram which resulted in a second automatic scram.
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| Operators failed to adequately control reactor vessel water level and pressure, while attempting to open the main steam isolation valves following the automatic reactor scram on December 18, 2000. This resulted in an automatic scram signal due to low reactor vessel water level. The inspectors reviewed this issue using the significance determination process for a transient with a loss of feedwater and determined this was of very low safety significance because the event occurred while the reactor was shut down and all control rods were already fully inserted.
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| Inspection Report# : 2000020(pdf)
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| Significance: Nov 16, 2000 Identified By: NRC Item Type: NCV NonCited Violation An alternate rod insertion system initiation and a manual reactor scram occurred with the reactor shutdown as a result of an inadequate
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| 3Q/2001 Inspection Findings - Clinton Page 2 of 8 maintenance procedure.
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| During replacement of power supplies for the alternate rod insertion (ARI) system, maintenance personnel failed to fully evaluate the impacts that re-energizing the power supplies had on the ARI initiation logic. While re-energizing the power supplies, the initiation logic sensed an ARI signal (low reactor water level). This caused the vent and drain valves to close and the scram discharge volume to fill with water. Plant operators inserted a manual scram signal before the automatic high scram discharge volume set point was reached. One Non-Cited Violation was identified for having an inadequate maintenance procedure to control this activity. This finding was evaluated using the shutdown significance determination process contained in Appendix G of IMC 0609 and was determined to have very low risk significance because it did not impact any of the five shutdown safety functions identified by NUMARC 91-06.
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| Inspection Report# : 2000017(pdf)
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| Significance: Sep 30, 2000 Identified By: NRC Item Type: FIN Finding Human performance errors and an inadequate procedure resulted in exceeding the allowed outage time for the emergency reserve auxiliary transformer static VAR compensator.
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| Human performance errors and the failure to develop an adequate procedure for the emergency reserve auxiliary transformer static VAR (Volt Ampere Reactive) compensator (ERAT-SVC) surveillance test resulted in several delays during the test. These delays caused the work to not be completed within the allowed outage time. Therefore, a request for Enforcement Discretion was presented to the NRC which was formally granted on September 20, 2000 (NOED 00-6-011). The safety significance of this finding was very low because all other emergency core cooling system trains (automatic depressurization system, low pressure core spray, and low pressure core injection), emergency diesel generators, and the reactor core isolation cooling system were operable.
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| Inspection Report# : 2000015(pdf)
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| Significance: May 17, 2000 Identified By: Self Disclosing Item Type: FIN Finding Manual reactor shutdown A labeling discrepancy contributed to the improper isolation of a protective relay for the 4.16kV Bus 1B Reserve Feed Breaker. As a result, during functional testing, the relay actuated and caused the bus to be de-energized which ultimately resulted in a manual reactor shut down. This issue was determined to be of very low risk significance due to remaining mitigation capability and recovery potential.
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| Inspection Report# : 2000008(pdf)
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| Mitigating Systems Significance: Jan 26, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow Condition Report process for Shutdown Service Water (SX) pipe wall thinning Corrective actions were not implemented to replace a portion of the shutdown service water (SX) system piping after pipe wall thinning was identified. The failure to take the specified corrective actions by the committed due date or to properly reevaluate the degraded condition was determined to be a Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, "Procedures." This finding was determined to have very low safety significance because the SX system remained operable and capable of performing its' safety function.
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| Inspection Report# : 2001002(pdf)
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| Significance: Jan 26, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct longstanding Reactor Core Isolation Cooling (RCIC) System valve degradation Corrective actions for a longstanding deficiency with the Reactor Core Isolation Cooling (RCIC) system steam bypass valve were not effective in stopping the leakage past the valve. This finding was determined to be a Non-Cited Violation of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action." This finding was determined to have very low risk significance because the degraded condition of the valve did not affect the operability of the RCIC system.
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| Inspection Report# : 2001002(pdf)
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| 3Q/2001 Inspection Findings - Clinton Page 3 of 8 Significance: N/A Dec 29, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW 10 CFR 55.59(c)(5) REQUIREMENTS FOR RETAINING LICENSED OPERATOR REQUALIFICATION PROGRAM RECORDS The inspectors identified a Non-Cited Violation wherein the facility licensee had failed to follow the Code of Federal Regulations (CFR) Title 10, Part 55.59(c)(5), Records, requirements by failing to systematically retain all of the original or authenticated copies of the original evaluation documents during the year 2000 annual NRC examination. The finding was of very low safety significance because although the records were not the original or authenticated copies of the original, records did exist in computerized clerically transcribed documents. The computer records had not been signed, and there was no indication that they had been verified correct by the original authors. The unauthenticated documents did provide information that licensed operators, for the most part, had participated and were evaluated during the year 2000 NRC annual requalification examination. However, the inspectors determined that the finding was more than minor. Specifically, the inspectors identified at least one instance in which the transcribed information appeared to be incorrect or missing. The records failure had credible impact on safety, in that, it negatively impacted on the intent of the licensed operator requalification examination process which, in part, is to maintain a high level of confidence that licensed operators continue to possess the requisite knowledge and abilities needed to safely perform licensed duties. In addition, inadequate records keeping adversely affects the NRC's ability to regulate.
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| Inspection Report# : 2001015(pdf)
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| Significance: Sep 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation Procedural requirements were not followed when unexpected equipment response was encountered.
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| Maintenance personnel failed to appropriately follow procedure instructions during testing of the Division III emergency diesel generator room fire detection system. These actions led to the emergency diesel generator being rendered inoperable. The procedure violation was treated as a Non-Cited Violation. This issue was of very low safety significance since the other divisional emergency diesel generators and all emergency core cooling systems were operable at the time of discovery.
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| Inspection Report# : 2000015(pdf)
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| Significance: Jun 16, 2000 Identified By: NRC Item Type: NCV NonCited Violation The licensee failed to ensure that appropriate post-modification testing was specified and accomplished for the Division I and Division III EDG output breaker circuitry modifications The licensee failed to ensure that the appropriate post-modification testing (PMT) was specified in the Division I and Division III emergency diesel generator (EDG) output breaker circuitry modification packages and that the post-modification tests were correctly accomplished. This was required to demonstrate through component and functional testing that the modified (rewired) portions of the Division I and Division III EDG output breaker circuitry were adequately installed to accomplish the intent of the plant design changes.
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| Inspection Report# : 2000012(pdf)
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| Barrier Integrity Significance: Nov 16, 2000 Identified By: NRC Item Type: NCV NonCited Violation Secondary containment was inoperable for 6 minutes during fuel movements when interlock doors were opened.
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| Secondary containment was inoperable for 6 minutes during fuel movements when secondary containment interlock doors were inadvertently opened to move scaffolding. The inoperability was discovered when operators in the control room received an alarm indicating a loss of secondary containment vacuum. One Non-Cited Violation was identified for violating Technical Specification 3.6.4.1 which requires secondary containment operability during fuel moves. This finding was evaluated using the shutdown significance determination process contained in Appendix G of IMC 0609 and was determined to have very low risk significance because it did not meet the criteria for findings requiring a phase 2 significance evaluation.
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| Inspection Report# : 2000017(pdf)
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| Significance: Nov 14, 2000
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| 3Q/2001 Inspection Findings - Clinton Page 4 of 8 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform radiographic examinations of Class 2 welds.
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| The inspectors identified a Non-Cited Violation for the failure to perform radiographic examinations of Class 2 welds in compliance with applicable American Society of Mechanical Engineers (ASME) Code requirements. During installation testing of the 1999 Feedwater Keep Fill FW-39 modification, five radiographic examinations had recorded geometric unsharpness values which exceeded Section III and Section V ASME Code limits. Radiographic geometric unsharpness values are used to ensure that the film is of adequate quality to see defects. In addition, inspectors identified that three examinations did not meet Section V Code requirements for documentation of radiographic technique variables which can affect the image quality of the film. The safety significance of this issue was considered very low at this time, based on the absence of adverse consequences, the presence of other image quality indicators, and because the issue did not involve the system isolation valves. The failure to comply with ASME Code radiographic examination requirements could result in the failure to detect flaws within reactor coolant boundary piping, and was considered a Non-Cited Violation of 10 CFR Part 50.55a, "Codes and Standards".
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| Inspection Report# : 2000019(pdf)
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| Emergency Preparedness Significance: Aug 21, 2001 Identified By: NRC Item Type: NCV NonCited Violation Violation of 10 CFR 50.54(q) re. SCBA qualifications A Non-Cited Violation of 10 CFR 50.54(q) was identified by the NRC associated with the failure to maintain personnel qualifications for self contained breathing apparatus in accordance with the Clinton Power Station Emergency Plan. The finding was of very low safety significance because the licensee maintained an adequate number of qualified personnel to maintain minimum coverage of the required positions identified in the Emergency Plan.
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| Inspection Report# : 2001010(pdf)
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| Significance: Jun 08, 2001 Identified By: NRC Item Type: VIO Violation Supplemental Inspection -- Failure to correct self-identified defficiencies disclosed through control room communications drills This supplemental inspection was performed by the NRC to assess the licensee's evaluation associated with inaccuracies in the reporting of the Drill and Exercise Performance (DEP) performance indicator and with the performance deficiencies that resulted in a White DEP performance indicator (fourth quarter 1999 through the fourth quarter 2000). During the inspection, performed in accordance with NRC Inspection Procedure 95001, the inspector concluded that the licensee performed an adequate evaluation to determine the causes of both issues. In the case of the performance indicator errors, the licensee performed a root cause evaluation which identified a personnel error that was compounded by the lack of self-checking and verification. In addition, the licensee identified contributing causes that included the failure to provide adequate training to the emergency preparedness staff and the failure to provide adequate procedural guidance to the performance indicator data stewards and verifiers, which also applied to performance indicators in other cornerstones. The inspector concluded that the scope of corrective actions planned and implemented by the licensee appeared to address the identified causes. However, the inspector observed an additional discrepancy in the recently completed performance indicator evaluation related to drill and exercise participation. In addition, the licensee identified an error in its evaluation of one of the other emergency preparedness performance indicators that was not detected during its evaluation. These observations demonstrated weaknesses in the licensee's corrective actions and extent of condition review. The errors in the licensee's reporting of the DEP performance indicator was significant, in that the error resulted in a change of color, (i.e., Green-to-White). Consequently, a violation of 10 CFR 50.9 of more than minor safety significance was identified. Since the inaccurate reporting occurred during the period that the NRC's Enforcement Policy afforded discretion for the non-willful submittal of inaccurate performance indicator information, the NRC is exercising enforcement discretion and not citing the violation. In the case of the White DEP performance indicator, the inspector concluded that the licensee adequately assessed the deficiencies that led to the performance issues. Based on its review, the licensee attributed the White performance indicator to the high failure rate of control room communicator drills (i.e., job performance measures). The licensee identified two apparent causes for the high failure rate: (1) weaknesses in formal training; and (2) failure to meet emergency preparedness management expectations concerning the identification and correction of drill deficiencies. The inspector reviewed the licensee's corrective actions and determined that they addressed the causes identified. As a result of the licensee's immediate corrective actions, the licensee's performance returned the performance indicator to the Green band. The inspector and the licensee concluded that the high failure rate of the control room communicators resulted, in part, from inadequate corrective actions for self-identified deficiencies. Specifically, the licensee control room communicator drills were a portion of an overall annual evaluation of non-licensed operators, which included non-emergency preparedness functions. Generally, the failure of the communications segment of the evaluation did not result in a total failure of the annual evaluation. Therefore, the licensee's remedial actions were limited and were not effective in correcting the deficiencies and preventing similar failures from occurring, as required by 10 CFR 50.47(b)(14). By letter dated 08/22/01, the NRC concluded that a violation of 10 CFR 50.47(b)(14) had occurred and using the NRC's significance determination process, determined that the finding was white.
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| Inspection Report# : 2001009(pdf)
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| 3Q/2001 Inspection Findings - Clinton Page 5 of 8 Significance: Feb 23, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to follow emergency plan for on-shift staffing For an approximate 2-month time period, the licensee failed to meet one of the minimum on-shift emergency response organization (ERO) staffing requirements contained in Table 2-1 of the licensee's emergency plan.
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| Inspection Report# : 2001003(pdf)
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| Significance: N/A Apr 28, 2000 Identified By: NRC Item Type: FIN Finding Emergency Preparedness Performance Indicator Verification Alert and Notification System, Drill & Exercise Participation, and Drill & Exercise performance indicators: The inspectors verified that the licensee had acceptably gathered information and reported these three performance indicators, which were in the green band, with the following minor exception. The inspectors identified a discrepancy with the licensee's initial assessment of the Drill and Exercise Performance (DEP) indicator related to the number of performance opportunities associated with a General Emergency declaration during a drill or an exercise. The licensee initially assumed that only three performance opportunities would exist rather than four as provided in NEI 99-02, but later recognized that they had misinterpreted the guidance. This did not affect the DEP performance indicator which was in the green band.
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| Inspection Report# : 2000009(pdf)
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| Occupational Radiation Safety Significance: Aug 21, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to maintain administrative control of high radiation area keys as required by Technical Specification 5.7.2 Technical Specification 5.7.2 requires, in part, that doors to high radiation areas in which an individual could receive a deep dose equivalent greater than or equal to 1000 millirem in one hour (at 30 centimeters) shall be provided with locked or continuously guarded doors to prevent unauthorized entry and that the keys to such doors shall be administratively controlled. During October 29 - 31, 2001, the licensee failed to maintain administrative control of a key that controlled five access points to high radiation areas specified above (i.e., lost the key and failed to perform required key inventories to identify its loss), as described in CR No. 2-00-11-016. Since the inspector concluded that sufficient barriers remained to prevent an unauthorized individual from entering the affected areas and receiving an overexposure, the inspector concluded that the incident was of very low safety significance. The licensee also reported the incident to the NRC as an occurrence for the Occupational Exposure Control Effectiveness performance indicator. This is being treated as a Non-Cited Violation.
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| Inspection Report# : 2001010(pdf)
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| Significance: SL-IV Jul 27, 2001 Identified By: NRC Item Type: NCV NonCited Violation Misuse of Radioactive Material to Alarm a PCM Radiation protection technician used contaminated material to alarm a portal contamination monitor (PCM), while an individual was performing a contamination survey. Based on the licensee's investigation, the contamination was not placed on the individual, and the individual successfully monitored through an additional PCM. This incident will be reviewed by the NRC for potential enforcement actions. Update: On July 27, 2001, the NRC identified and forwarded to the licensee (by letter) a Non-Cited Violation of the Clinton Station Facility Operating License associated with the deliberate misuse of radioactive material by a junior contract radiation protection technician. On October 20, 2000, the technician misused radioactive material to cause an erroneous alarm on a PCM, as another individual was performing a contamination survey. The licensee identified the incident, entered the incident into its corrective action program, and implemented immediate corrective actions. Since the violation was determined to be willful, the NRC did not assign a significance to the violation using the NRC's Significance Determination Process. In accordance with the NRC Enforcement Policy, the NRC determined that the incident constituted a Severity Level IV violation of the Clinton Power Station Facility Operating License. Further, the NRC determined that the violation met the criteria necessary to disposition the violation as a Non-Cited Violation (Section VI.A.1.d of the NRC Enforcement Policy).
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| Inspection Report# : 2000018(pdf)
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| Inspection Report# : 2001010(pdf)
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| Significance: Oct 25, 2000 Identified By: Licensee
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| 3Q/2001 Inspection Findings - Clinton Page 6 of 8 Item Type: NCV NonCited Violation Three individuals entered a HRA in violation of Technical Specification 5.7.1 On October 25, 2000, three individuals entered the B residual heat removal heat exchanger room (a posted high radiation area); however, the individuals were not working under a radiation work permit that allowed entry into the high radiation area and did not satisfy either of the three entry conditions of Technical Specification 5.7.1.
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| Inspection Report# : 2000018(pdf)
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| Significance: Oct 08, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an Adequate Survey to identify and to post a High Radiation Area A finding and associated Non-Cited Violation was identified concerning the failure to perform an adequate radiological survey, as required by 10 CFR 20.1501. Although the licensee identified this issue, the licensee did not thoroughly evaluate the cause(s) of the unanticipated radiological conditions and associated problems in the monitoring of radioactive waste activities, which have resulted in previous, similar incidents. The finding was of very low safety significance because the area radiation levels and the licensee's additional administrative barriers would have limited the potential for an individual inadvertently entering the area and receiving a radiation exposure in excess of regulatory limits.
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| Inspection Report# : 2001015(pdf)
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| Public Radiation Safety Significance: Dec 08, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Inadvertent Release of Radioactive Material to Unrestricted Area During September 2000, the licensee conducted a survey of tools, equipment, etc. outside of the restricted area (protected area and owner controlled area) and identified low-level contaminated materials that were not under constant surveillance or control. The failure to maintain contstant surveillance and control of the material was a violation of 10 CFR 20.1802 and was characterized as a Non-Cited Violation. Based on the licensee's conservative annual dose assessment (about 1.56 millirem) and the inability to define the origin of each of the items, the inspector concluded that the issue constituted one occurrence/event per the NRC Significance Determination Process (Green).
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| Inspection Report# : 2000021(pdf)
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| Physical Protection Miscellaneous Significance: SL-IV Aug 18, 2001 Identified By: NRC Item Type: VIO Violation Falsification of Test Records by Licensee Employee SL IV - On July 2, 2001, by separate letter, NRC issued a Severity Level IV violation of 10 CFR 50.9 for a deliberate falsification by a plant test engineer. Following investigation by the Office of Investigations, NRC determined that, on October 20, 2000, a test engineer forged another employee's signature on two test package cover sheets on by forging another employee's signature without his prior concurrence, in violation of Clinton established plant protocol and procedure.
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| Inspection Report# : 2001010(pdf)
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| Significance: SL-IV Apr 06, 2001 Identified By: NRC Item Type: VIO Violation Violation of 10 CFR 50.7 "Employee Protection" On April 6, 2001, the NRC issued the licensee a Severity Level IV Violation of 10 CFR 50.7. The NRC concluded that the licensee took adverse employment actions against an employee in the licensee's Nuclear Training Department (i.e., unfavorable 1999 performance review), in part, as a result of the employee's engagement in protected activities. In addition, the NRC learned that several training personnel may be reluctant to
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| 3Q/2001 Inspection Findings - Clinton Page 7 of 8 discuss department issues within the nuclear training department.
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| Inspection Report# : 2001007(pdf)
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| Inspection Report# : 2001010(pdf)
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| Significance: N/A Jan 26, 2001 Identified By: NRC Item Type: FIN Finding Assessment of Problem Identification and Resolution Performance The team identified that the licensee appropriately entered significant plant issues into the corrective action process by initiating condition reports.
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| Some less significant conditions adverse to quality were evaluated and corrected outside the established process. The trending program was not fully effective as a problem identification tool. Quality Assurance audits and self-assessments reviewed varied in quality. Identified issues were generally evaluated properly, although in several cases the corrective action process did not work effectively to either evaluate or prioritize issues.
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| Current station performance issues including human performance, corrective action program, surveillance testing, and labeling indicate that long term corrective actions previously taken in these areas as restart and post-restart initiatives have not been fully effective to support sustained improvement. Corrective actions were not always fully effective or timely for some individual equipment issues and the effectiveness review process (CARE) did not always identify ineffective corrective actions. The licensee had recently recognized similar deficiencies in corrective action program implementation but had not yet fully developed or completed the corrective actions to improve these areas. The inspectors did not find any reluctance by the station employees to raise safety issues.
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| Inspection Report# : 2001002(pdf)
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| Significance: N/A Dec 31, 2000 Identified By: NRC Item Type: FIN Finding Recent human performance issues have occurred which are associated with operator performance and knowledge based deficiencies that have affected plant operations and responses to transient conditions.
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| Recent human performance issues have occurred which are associated with operator performance and knowledge based deficiencies that have affected plant operations and responses to transient conditions. While the risk of the individual events was very low, the failure of operators to adequately control level parameters indicated a declining trend in this area. These issues could not be easily evaluated by present risk analysis methods because failures to follow procedures and maintaining management expectations were not modeled in the Clinton Individual Plant Evaluation. Therefore, the finding is characterized as having no color.
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| Inspection Report# : 2000020(pdf)
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| Significance: N/A Nov 14, 2000 Identified By: NRC Item Type: FIN Finding Three procedures were not written in compliance with the applicable ASME Code.
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| The inspectors reviewed three special process procedures, and identified areas where all three procedures were not written in compliance with the applicable ASME Code. The procedure deficiencies had the potential to affect the ASME Code compliance of weld fabrication and nondestructive examination used on safety-related components and piping. The inspectors noted that each of the ASME Code problems identified contained elements of human performance deficiencies. The human performance aspects, while not always being the root cause of the problem, were significant contributors leading to procedure deficiencies. While the risk of the individual examples was very low, the number of deficiencies indicated a problem with incorporation of applicable ASME Code requirements into special process procedures.
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| Inspection Report# : 2000019(pdf)
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| Significance: Feb 17, 2002 Identified By: NRC Item Type: FIN Finding A temporary modification on the "A" RR FCV control cirucuitry.
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| On December 14, 2001, the licensee installed a temporary modification on the "A" RR FCV control circuitry. The T-mod was installed to assist the operators in manually controlling the "A" RR FCV because the reliability of the normal control circuitry was in question. During the implementation portion of the T-mod installations, the "A" RR FCV unexpectedly moved from 94 percent open to 102 percent open at which point the protective position circuitry locked the valve at the 102 percent position. Recator power was observed to go from 94 percent to 98 percent during this unexpected valve movement. Following this unexpected FCV movement, opeerations personnel ordered the T-mod to be removed and operators then proceeded to manually shut down the reactor without any further movements of the "A" RR FCV.
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| Inspection Report# : 2001016(pdf)
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| Significance: Feb 17, 2002 Identified By: NRC Item Type: NCV NonCited Violation Non-cited violation of T.S. 5.4.1 for an inadequate surveillance procedure.
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| 3Q/2001 Inspection Findings - Clinton Page 8 of 8 Inspection Report# : 2001016(pdf)
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| Significance: N/A Sep 30, 2000 Identified By: NRC Item Type: FIN Finding Recent events affecting plant operations contained elements of human performance deficiencies.
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| NO COLOR. The inspectors noted that several recent events which have affected plant operations and the operability of safety-related components or other components important to safety contained elements of human performance deficiencies. The human performance aspects, while not always being the root cause of the problem, were significant contributors leading to the events. While the risk of the individual events was very low, the number of maintenance-related incidents indicated a problem exists with the control, review, and performance of maintenance activities.
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| Inspection Report# : 2000015(pdf)
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| Significance: N/A May 20, 2000 Identified By: NRC Item Type: FIN Finding Inaccurate historical data for the Safety System Functional Failure Indicator No Color. The licensee identified a failure to submit accurate information to the NRC. The inaccurate information involved the historical data submittal for the Safety System Functional Failure Performance Indicator. The error resulted in a response band color change from Green to White for the first quarter 1999 Performance Indicator. The NRC exercised Enforcement Discretion pursuant to Section VII.B.6 of the Enforcement Policy and did not cite the violation.
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| Inspection Report# : 2000008(pdf)
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| Last modified : March 26, 2002
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| 4Q/2001 Inspection Findings - Clinton Page 1 of 7 Clinton Initiating Events Significance: Aug 21, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow procedures associated with feed water level control system surveillance testing.
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| Human performance and corrective action deficiencies contributed to a Non-Cited Violation of Technical Specification 5.4.1 for failing to follow procedures. This led to the unplanned automatic reactor shutdown on July 24, 2001. The finding was of very low safety significance because no complications occurred during the unplanned automatic reactor shut down and the finding did not increase the likelihood of mitigation equipment being unavailable.
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| Inspection Report# : 2001010(pdf)
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| Significance: Dec 31, 2000 Identified By: NRC Item Type: FIN Finding Operators did not adequately control reactor vessel inventory after a reactor scram which resulted in the motor driven reactor feedwater pump tripping on high reactor vessel water level.
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| During operator response to a reactor scram on December 18, 2000, operators did not adequately control reactor vessel inventory prior to the motor driven reactor feedwater pump tripping on high reactor vessel water level. The inspectors reviewed this issue using the significance determination process for a transient with a loss of feedwater and determined this was of very low safety significance because all other reactor vessel level control systems were operable and functioned as designed.
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| Inspection Report# : 2000020(pdf)
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| Significance: Dec 31, 2000 Identified By: NRC Item Type: NCV NonCited Violation Operators failed to adequately evaluate an alarming moisture separator drain tank level annunciator that resulted in a turbine trip.
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| During plant restart following refueling outage 7, operators did not adequately evaluate an alarming moisture separator drain tank level annunciator.
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| As a result, high water level in the moisture separator drain tank caused a turbine trip with the reactor at approximately 25% power. The inspectors reviewed this issue using the significance determination process for a transient. Since only the initiating event cornerstone is affected and associated assumptions have no other impact than slightly increasing the likelihood of an uncomplicated reactor trip, the finding is considered to be of very low safety significance.
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| Inspection Report# : 2000020(pdf)
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| Significance: Dec 31, 2000 Identified By: NRC Item Type: NCV NonCited Violation Operators failed to adequately control reactor vessel water level and pressure following the automatic reactor scram which resulted in a second automatic scram.
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| Operators failed to adequately control reactor vessel water level and pressure, while attempting to open the main steam isolation valves following the automatic reactor scram on December 18, 2000. This resulted in an automatic scram signal due to low reactor vessel water level. The inspectors reviewed this issue using the significance determination process for a transient with a loss of feedwater and determined this was of very low safety significance because the event occurred while the reactor was shut down and all control rods were already fully inserted.
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| Inspection Report# : 2000020(pdf)
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| Significance: Nov 16, 2000 Identified By: NRC Item Type: NCV NonCited Violation An alternate rod insertion system initiation and a manual reactor scram occurred with the reactor shutdown as a result of an inadequate maintenance procedure.
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| During replacement of power supplies for the alternate rod insertion (ARI) system, maintenance personnel failed to fully evaluate the impacts that re-energizing the power supplies had on the ARI initiation logic. While re-energizing the power supplies, the initiation logic sensed an ARI signal (low reactor water level). This caused the vent and drain valves to close and the scram discharge volume to fill with water. Plant operators inserted a manual scram signal before the automatic high scram discharge volume set point was reached. One Non-Cited Violation was identified for having an inadequate maintenance procedure to control this activity. This finding was evaluated using the shutdown significance determination process contained in Appendix G of IMC 0609 and was determined to have very low risk significance because it did not impact any of the five shutdown safety functions identified by NUMARC 91-06.
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| 4Q/2001 Inspection Findings - Clinton Page 2 of 7 Inspection Report# : 2000017(pdf)
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| Significance: Sep 30, 2000 Identified By: NRC Item Type: FIN Finding Human performance errors and an inadequate procedure resulted in exceeding the allowed outage time for the emergency reserve auxiliary transformer static VAR compensator.
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| Human performance errors and the failure to develop an adequate procedure for the emergency reserve auxiliary transformer static VAR (Volt Ampere Reactive) compensator (ERAT-SVC) surveillance test resulted in several delays during the test. These delays caused the work to not be completed within the allowed outage time. Therefore, a request for Enforcement Discretion was presented to the NRC which was formally granted on September 20, 2000 (NOED 00-6-011). The safety significance of this finding was very low because all other emergency core cooling system trains (automatic depressurization system, low pressure core spray, and low pressure core injection), emergency diesel generators, and the reactor core isolation cooling system were operable.
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| Inspection Report# : 2000015(pdf)
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| Significance: May 17, 2000 Identified By: Self Disclosing Item Type: FIN Finding Manual reactor shutdown A labeling discrepancy contributed to the improper isolation of a protective relay for the 4.16kV Bus 1B Reserve Feed Breaker. As a result, during functional testing, the relay actuated and caused the bus to be de-energized which ultimately resulted in a manual reactor shut down. This issue was determined to be of very low risk significance due to remaining mitigation capability and recovery potential.
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| Inspection Report# : 2000008(pdf)
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| Mitigating Systems Significance: N/A Dec 29, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW 10 CFR 55.59(c)(5) REQUIREMENTS FOR RETAINING LICENSED OPERATOR REQUALIFICATION PROGRAM RECORDS The inspectors identified a Non-Cited Violation wherein the facility licensee had failed to follow the Code of Federal Regulations (CFR) Title 10, Part 55.59(c)(5), Records, requirements by failing to systematically retain all of the original or authenticated copies of the original evaluation documents during the year 2000 annual NRC examination. The finding was of very low safety significance because although the records were not the original or authenticated copies of the original, records did exist in computerized clerically transcribed documents. The computer records had not been signed, and there was no indication that they had been verified correct by the original authors. The unauthenticated documents did provide information that licensed operators, for the most part, had participated and were evaluated during the year 2000 NRC annual requalification examination. However, the inspectors determined that the finding was more than minor. Specifically, the inspectors identified at least one instance in which the transcribed information appeared to be incorrect or missing. The records failure had credible impact on safety, in that, it negatively impacted on the intent of the licensed operator requalification examination process which, in part, is to maintain a high level of confidence that licensed operators continue to possess the requisite knowledge and abilities needed to safely perform licensed duties. In addition, inadequate records keeping adversely affects the NRC's ability to regulate.
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| Inspection Report# : 2001015(pdf)
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| Significance: Jan 26, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow Condition Report process for Shutdown Service Water (SX) pipe wall thinning Corrective actions were not implemented to replace a portion of the shutdown service water (SX) system piping after pipe wall thinning was identified. The failure to take the specified corrective actions by the committed due date or to properly reevaluate the degraded condition was determined to be a Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, "Procedures." This finding was determined to have very low safety significance because the SX system remained operable and capable of performing its' safety function.
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| Inspection Report# : 2001002(pdf)
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| Significance: Jan 26, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct longstanding Reactor Core Isolation Cooling (RCIC) System valve degradation Corrective actions for a longstanding deficiency with the Reactor Core Isolation Cooling (RCIC) system steam bypass valve were not effective in stopping the leakage past the valve. This finding was determined to be a Non-Cited Violation of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action." This finding was determined to have very low risk significance because the degraded condition of the valve did not affect the operability of
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| 4Q/2001 Inspection Findings - Clinton Page 3 of 7 the RCIC system.
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| Inspection Report# : 2001002(pdf)
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| Significance: Sep 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation Procedural requirements were not followed when unexpected equipment response was encountered.
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| Maintenance personnel failed to appropriately follow procedure instructions during testing of the Division III emergency diesel generator room fire detection system. These actions led to the emergency diesel generator being rendered inoperable. The procedure violation was treated as a Non-Cited Violation. This issue was of very low safety significance since the other divisional emergency diesel generators and all emergency core cooling systems were operable at the time of discovery.
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| Inspection Report# : 2000015(pdf)
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| Significance: Jun 16, 2000 Identified By: NRC Item Type: NCV NonCited Violation The licensee failed to ensure that appropriate post-modification testing was specified and accomplished for the Division I and Division III EDG output breaker circuitry modifications The licensee failed to ensure that the appropriate post-modification testing (PMT) was specified in the Division I and Division III emergency diesel generator (EDG) output breaker circuitry modification packages and that the post-modification tests were correctly accomplished. This was required to demonstrate through component and functional testing that the modified (rewired) portions of the Division I and Division III EDG output breaker circuitry were adequately installed to accomplish the intent of the plant design changes.
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| Inspection Report# : 2000012(pdf)
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| Barrier Integrity Significance: Nov 16, 2000 Identified By: NRC Item Type: NCV NonCited Violation Secondary containment was inoperable for 6 minutes during fuel movements when interlock doors were opened.
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| Secondary containment was inoperable for 6 minutes during fuel movements when secondary containment interlock doors were inadvertently opened to move scaffolding. The inoperability was discovered when operators in the control room received an alarm indicating a loss of secondary containment vacuum. One Non-Cited Violation was identified for violating Technical Specification 3.6.4.1 which requires secondary containment operability during fuel moves. This finding was evaluated using the shutdown significance determination process contained in Appendix G of IMC 0609 and was determined to have very low risk significance because it did not meet the criteria for findings requiring a phase 2 significance evaluation.
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| Inspection Report# : 2000017(pdf)
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| Significance: Nov 14, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform radiographic examinations of Class 2 welds.
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| The inspectors identified a Non-Cited Violation for the failure to perform radiographic examinations of Class 2 welds in compliance with applicable American Society of Mechanical Engineers (ASME) Code requirements. During installation testing of the 1999 Feedwater Keep Fill FW-39 modification, five radiographic examinations had recorded geometric unsharpness values which exceeded Section III and Section V ASME Code limits. Radiographic geometric unsharpness values are used to ensure that the film is of adequate quality to see defects. In addition, inspectors identified that three examinations did not meet Section V Code requirements for documentation of radiographic technique variables which can affect the image quality of the film. The safety significance of this issue was considered very low at this time, based on the absence of adverse consequences, the presence of other image quality indicators, and because the issue did not involve the system isolation valves. The failure to comply with ASME Code radiographic examination requirements could result in the failure to detect flaws within reactor coolant boundary piping, and was considered a Non-Cited Violation of 10 CFR Part 50.55a, "Codes and Standards".
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| Inspection Report# : 2000019(pdf)
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| Emergency Preparedness Significance: Aug 21, 2001
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| 4Q/2001 Inspection Findings - Clinton Page 4 of 7 Identified By: NRC Item Type: NCV NonCited Violation Violation of 10 CFR 50.54(q) re. SCBA qualifications A Non-Cited Violation of 10 CFR 50.54(q) was identified by the NRC associated with the failure to maintain personnel qualifications for self contained breathing apparatus in accordance with the Clinton Power Station Emergency Plan. The finding was of very low safety significance because the licensee maintained an adequate number of qualified personnel to maintain minimum coverage of the required positions identified in the Emergency Plan.
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| Inspection Report# : 2001010(pdf)
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| Significance: Jun 08, 2001 Identified By: NRC Item Type: VIO Violation Supplemental Inspection -- Failure to correct self-identified defficiencies disclosed through control room communications drills This supplemental inspection was performed by the NRC to assess the licensee's evaluation associated with inaccuracies in the reporting of the Drill and Exercise Performance (DEP) performance indicator and with the performance deficiencies that resulted in a White DEP performance indicator (fourth quarter 1999 through the fourth quarter 2000). During the inspection, performed in accordance with NRC Inspection Procedure 95001, the inspector concluded that the licensee performed an adequate evaluation to determine the causes of both issues. In the case of the performance indicator errors, the licensee performed a root cause evaluation which identified a personnel error that was compounded by the lack of self-checking and verification. In addition, the licensee identified contributing causes that included the failure to provide adequate training to the emergency preparedness staff and the failure to provide adequate procedural guidance to the performance indicator data stewards and verifiers, which also applied to performance indicators in other cornerstones. The inspector concluded that the scope of corrective actions planned and implemented by the licensee appeared to address the identified causes. However, the inspector observed an additional discrepancy in the recently completed performance indicator evaluation related to drill and exercise participation. In addition, the licensee identified an error in its evaluation of one of the other emergency preparedness performance indicators that was not detected during its evaluation. These observations demonstrated weaknesses in the licensee's corrective actions and extent of condition review. The errors in the licensee's reporting of the DEP performance indicator was significant, in that the error resulted in a change of color, (i.e., Green-to-White). Consequently, a violation of 10 CFR 50.9 of more than minor safety significance was identified. Since the inaccurate reporting occurred during the period that the NRC's Enforcement Policy afforded discretion for the non-willful submittal of inaccurate performance indicator information, the NRC is exercising enforcement discretion and not citing the violation. In the case of the White DEP performance indicator, the inspector concluded that the licensee adequately assessed the deficiencies that led to the performance issues. Based on its review, the licensee attributed the White performance indicator to the high failure rate of control room communicator drills (i.e., job performance measures). The licensee identified two apparent causes for the high failure rate: (1) weaknesses in formal training; and (2) failure to meet emergency preparedness management expectations concerning the identification and correction of drill deficiencies. The inspector reviewed the licensee's corrective actions and determined that they addressed the causes identified. As a result of the licensee's immediate corrective actions, the licensee's performance returned the performance indicator to the Green band. The inspector and the licensee concluded that the high failure rate of the control room communicators resulted, in part, from inadequate corrective actions for self-identified deficiencies. Specifically, the licensee control room communicator drills were a portion of an overall annual evaluation of non-licensed operators, which included non-emergency preparedness functions. Generally, the failure of the communications segment of the evaluation did not result in a total failure of the annual evaluation. Therefore, the licensee's remedial actions were limited and were not effective in correcting the deficiencies and preventing similar failures from occurring, as required by 10 CFR 50.47(b)(14). By letter dated 08/22/01, the NRC concluded that a violation of 10 CFR 50.47(b)(14) had occurred and using the NRC's significance determination process, determined that the finding was white.
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| Inspection Report# : 2001009(pdf)
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| Significance: Feb 23, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to follow emergency plan for on-shift staffing For an approximate 2-month time period, the licensee failed to meet one of the minimum on-shift emergency response organization (ERO) staffing requirements contained in Table 2-1 of the licensee's emergency plan.
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| Inspection Report# : 2001003(pdf)
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| Significance: N/A Apr 28, 2000 Identified By: NRC Item Type: FIN Finding Emergency Preparedness Performance Indicator Verification Alert and Notification System, Drill & Exercise Participation, and Drill & Exercise performance indicators: The inspectors verified that the licensee had acceptably gathered information and reported these three performance indicators, which were in the green band, with the following minor exception. The inspectors identified a discrepancy with the licensee's initial assessment of the Drill and Exercise Performance (DEP) indicator related to the number of performance opportunities associated with a General Emergency declaration during a drill or an exercise. The licensee initially assumed that only three performance opportunities would exist rather than four as provided in NEI 99-02, but later recognized that they had misinterpreted the guidance. This did not affect the DEP performance indicator which was in the green band.
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| Inspection Report# : 2000009(pdf)
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| Occupational Radiation Safety
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| 4Q/2001 Inspection Findings - Clinton Page 5 of 7 Significance: Oct 08, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an Adequate Survey to identify and to post a High Radiation Area A finding and associated Non-Cited Violation was identified concerning the failure to perform an adequate radiological survey, as required by 10 CFR 20.1501. Although the licensee identified this issue, the licensee did not thoroughly evaluate the cause(s) of the unanticipated radiological conditions and associated problems in the monitoring of radioactive waste activities, which have resulted in previous, similar incidents. The finding was of very low safety significance because the area radiation levels and the licensee's additional administrative barriers would have limited the potential for an individual inadvertently entering the area and receiving a radiation exposure in excess of regulatory limits.
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| Inspection Report# : 2001015(pdf)
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| Significance: Aug 21, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to maintain administrative control of high radiation area keys as required by Technical Specification 5.7.2 Technical Specification 5.7.2 requires, in part, that doors to high radiation areas in which an individual could receive a deep dose equivalent greater than or equal to 1000 millirem in one hour (at 30 centimeters) shall be provided with locked or continuously guarded doors to prevent unauthorized entry and that the keys to such doors shall be administratively controlled. During October 29 - 31, 2001, the licensee failed to maintain administrative control of a key that controlled five access points to high radiation areas specified above (i.e., lost the key and failed to perform required key inventories to identify its loss), as described in CR No. 2-00-11-016. Since the inspector concluded that sufficient barriers remained to prevent an unauthorized individual from entering the affected areas and receiving an overexposure, the inspector concluded that the incident was of very low safety significance. The licensee also reported the incident to the NRC as an occurrence for the Occupational Exposure Control Effectiveness performance indicator. This is being treated as a Non-Cited Violation.
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| Inspection Report# : 2001010(pdf)
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| Significance: SL-IV Jul 27, 2001 Identified By: NRC Item Type: NCV NonCited Violation Misuse of Radioactive Material to Alarm a PCM Radiation protection technician used contaminated material to alarm a portal contamination monitor (PCM), while an individual was performing a contamination survey. Based on the licensee's investigation, the contamination was not placed on the individual, and the individual successfully monitored through an additional PCM. This incident will be reviewed by the NRC for potential enforcement actions. Update: On July 27, 2001, the NRC identified and forwarded to the licensee (by letter) a Non-Cited Violation of the Clinton Station Facility Operating License associated with the deliberate misuse of radioactive material by a junior contract radiation protection technician. On October 20, 2000, the technician misused radioactive material to cause an erroneous alarm on a PCM, as another individual was performing a contamination survey. The licensee identified the incident, entered the incident into its corrective action program, and implemented immediate corrective actions. Since the violation was determined to be willful, the NRC did not assign a significance to the violation using the NRC's Significance Determination Process. In accordance with the NRC Enforcement Policy, the NRC determined that the incident constituted a Severity Level IV violation of the Clinton Power Station Facility Operating License. Further, the NRC determined that the violation met the criteria necessary to disposition the violation as a Non-Cited Violation (Section VI.A.1.d of the NRC Enforcement Policy).
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| Inspection Report# : 2000018(pdf)
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| Inspection Report# : 2001010(pdf)
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| Significance: Oct 25, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Three individuals entered a HRA in violation of Technical Specification 5.7.1 On October 25, 2000, three individuals entered the B residual heat removal heat exchanger room (a posted high radiation area); however, the individuals were not working under a radiation work permit that allowed entry into the high radiation area and did not satisfy either of the three entry conditions of Technical Specification 5.7.1.
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| Inspection Report# : 2000018(pdf)
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| Public Radiation Safety Significance: Dec 08, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Inadvertent Release of Radioactive Material to Unrestricted Area During September 2000, the licensee conducted a survey of tools, equipment, etc. outside of the restricted area (protected area and owner controlled area) and identified low-level contaminated materials that were not under constant surveillance or control. The failure to maintain contstant surveillance and control of the material was a violation of 10 CFR 20.1802 and was characterized as a Non-Cited Violation. Based on the licensee's conservative annual dose assessment (about 1.56 millirem) and the inability to define the origin of each of the items, the inspector
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| 4Q/2001 Inspection Findings - Clinton Page 6 of 7 concluded that the issue constituted one occurrence/event per the NRC Significance Determination Process (Green).
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| Inspection Report# : 2000021(pdf)
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| Physical Protection Miscellaneous Significance: SL-IV Aug 18, 2001 Identified By: NRC Item Type: VIO Violation Falsification of Test Records by Licensee Employee SL IV - On July 2, 2001, by separate letter, NRC issued a Severity Level IV violation of 10 CFR 50.9 for a deliberate falsification by a plant test engineer. Following investigation by the Office of Investigations, NRC determined that, on October 20, 2000, a test engineer forged another employee's signature on two test package cover sheets on by forging another employee's signature without his prior concurrence, in violation of Clinton established plant protocol and procedure.
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| Inspection Report# : 2001010(pdf)
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| Significance: SL-IV Apr 06, 2001 Identified By: NRC Item Type: VIO Violation Violation of 10 CFR 50.7 "Employee Protection" On April 6, 2001, the NRC issued the licensee a Severity Level IV Violation of 10 CFR 50.7. The NRC concluded that the licensee took adverse employment actions against an employee in the licensee's Nuclear Training Department (i.e., unfavorable 1999 performance review), in part, as a result of the employee's engagement in protected activities. In addition, the NRC learned that several training personnel may be reluctant to discuss department issues within the nuclear training department.
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| Inspection Report# : 2001007(pdf)
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| Inspection Report# : 2001010(pdf)
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| Significance: N/A Jan 26, 2001 Identified By: NRC Item Type: FIN Finding Assessment of Problem Identification and Resolution Performance The team identified that the licensee appropriately entered significant plant issues into the corrective action process by initiating condition reports.
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| Some less significant conditions adverse to quality were evaluated and corrected outside the established process. The trending program was not fully effective as a problem identification tool. Quality Assurance audits and self-assessments reviewed varied in quality. Identified issues were generally evaluated properly, although in several cases the corrective action process did not work effectively to either evaluate or prioritize issues.
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| Current station performance issues including human performance, corrective action program, surveillance testing, and labeling indicate that long term corrective actions previously taken in these areas as restart and post-restart initiatives have not been fully effective to support sustained improvement. Corrective actions were not always fully effective or timely for some individual equipment issues and the effectiveness review process (CARE) did not always identify ineffective corrective actions. The licensee had recently recognized similar deficiencies in corrective action program implementation but had not yet fully developed or completed the corrective actions to improve these areas. The inspectors did not find any reluctance by the station employees to raise safety issues.
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| Inspection Report# : 2001002(pdf)
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| Significance: N/A Dec 31, 2000 Identified By: NRC Item Type: FIN Finding Recent human performance issues have occurred which are associated with operator performance and knowledge based deficiencies that have affected plant operations and responses to transient conditions.
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| Recent human performance issues have occurred which are associated with operator performance and knowledge based deficiencies that have affected plant operations and responses to transient conditions. While the risk of the individual events was very low, the failure of operators to adequately control level parameters indicated a declining trend in this area. These issues could not be easily evaluated by present risk analysis methods because failures to follow procedures and maintaining management expectations were not modeled in the Clinton Individual Plant Evaluation. Therefore, the finding is characterized as having no color.
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| Inspection Report# : 2000020(pdf)
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| Significance: N/A Nov 14, 2000 Identified By: NRC Item Type: FIN Finding Three procedures were not written in compliance with the applicable ASME Code.
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| The inspectors reviewed three special process procedures, and identified areas where all three procedures were not written in compliance with the applicable ASME Code. The procedure deficiencies had the potential to affect the ASME Code compliance of weld fabrication and nondestructive examination used on safety-related components and piping. The inspectors noted that each of the ASME Code problems identified contained elements of human performance deficiencies. The human performance aspects, while not always being the root cause of the problem, were significant contributors leading to procedure deficiencies. While the risk of the individual examples was very low, the number of deficiencies indicated a problem with incorporation of applicable ASME Code requirements into special process procedures.
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| 4Q/2001 Inspection Findings - Clinton Page 7 of 7 Inspection Report# : 2000019(pdf)
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| Significance: N/A Sep 30, 2000 Identified By: NRC Item Type: FIN Finding Recent events affecting plant operations contained elements of human performance deficiencies.
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| NO COLOR. The inspectors noted that several recent events which have affected plant operations and the operability of safety-related components or other components important to safety contained elements of human performance deficiencies. The human performance aspects, while not always being the root cause of the problem, were significant contributors leading to the events. While the risk of the individual events was very low, the number of maintenance-related incidents indicated a problem exists with the control, review, and performance of maintenance activities.
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| Inspection Report# : 2000015(pdf)
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| Significance: N/A May 20, 2000 Identified By: NRC Item Type: FIN Finding Inaccurate historical data for the Safety System Functional Failure Indicator No Color. The licensee identified a failure to submit accurate information to the NRC. The inaccurate information involved the historical data submittal for the Safety System Functional Failure Performance Indicator. The error resulted in a response band color change from Green to White for the first quarter 1999 Performance Indicator. The NRC exercised Enforcement Discretion pursuant to Section VII.B.6 of the Enforcement Policy and did not cite the violation.
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| Inspection Report# : 2000008(pdf)
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| Last modified : March 01, 2002
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| 1Q/2002 Inspection Findings - Clinton Page 1 of 8 Clinton Initiating Events Significance: Aug 21, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow procedures associated with feed water level control system surveillance testing.
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| Human performance and corrective action deficiencies contributed to a Non-Cited Violation of Technical Specification 5.4.1 for failing to follow procedures. This led to the unplanned automatic reactor shutdown on July 24, 2001. The finding was of very low safety significance because no complications occurred during the unplanned automatic reactor shut down and the finding did not increase the likelihood of mitigation equipment being unavailable.
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| Inspection Report# : 2001010(pdf)
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| Significance: Dec 31, 2000 Identified By: NRC Item Type: NCV NonCited Violation Operators failed to adequately evaluate an alarming moisture separator drain tank level annunciator that resulted in a turbine trip.
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| During plant restart following refueling outage 7, operators did not adequately evaluate an alarming moisture separator drain tank level annunciator. As a result, high water level in the moisture separator drain tank caused a turbine trip with the reactor at approximately 25% power. The inspectors reviewed this issue using the significance determination process for a transient. Since only the initiating event cornerstone is affected and associated assumptions have no other impact than slightly increasing the likelihood of an uncomplicated reactor trip, the finding is considered to be of very low safety significance.
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| Inspection Report# : 2000020(pdf)
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| Significance: Dec 31, 2000 Identified By: NRC Item Type: FIN Finding Operators did not adequately control reactor vessel inventory after a reactor scram which resulted in the motor driven reactor feedwater pump tripping on high reactor vessel water level.
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| During operator response to a reactor scram on December 18, 2000, operators did not adequately control reactor vessel inventory prior to the motor driven reactor feedwater pump tripping on high reactor vessel water level. The inspectors reviewed this issue using the significance determination process for a transient with a loss of feedwater and determined this was of very low safety significance because all other reactor vessel level control systems were operable and functioned as designed.
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| Inspection Report# : 2000020(pdf)
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| Significance: Dec 31, 2000 Identified By: NRC Item Type: NCV NonCited Violation Operators failed to adequately control reactor vessel water level and pressure following the automatic reactor scram which resulted in a second automatic scram.
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| Operators failed to adequately control reactor vessel water level and pressure, while attempting to open the main steam isolation valves following the automatic reactor scram on December 18, 2000. This resulted in an automatic scram signal due to low reactor vessel water level. The inspectors reviewed this issue using the significance determination process for a transient with a loss of feedwater and determined this was of very low safety significance because the event occurred while the reactor was shut down and all control rods were already fully inserted.
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| Inspection Report# : 2000020(pdf)
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| Significance: Nov 16, 2000 Identified By: NRC Item Type: NCV NonCited Violation An alternate rod insertion system initiation and a manual reactor scram occurred with the reactor shutdown as a result of
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| 1Q/2002 Inspection Findings - Clinton Page 2 of 8 an inadequate maintenance procedure.
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| During replacement of power supplies for the alternate rod insertion (ARI) system, maintenance personnel failed to fully evaluate the impacts that re-energizing the power supplies had on the ARI initiation logic. While re-energizing the power supplies, the initiation logic sensed an ARI signal (low reactor water level). This caused the vent and drain valves to close and the scram discharge volume to fill with water. Plant operators inserted a manual scram signal before the automatic high scram discharge volume set point was reached. One Non-Cited Violation was identified for having an inadequate maintenance procedure to control this activity. This finding was evaluated using the shutdown significance determination process contained in Appendix G of IMC 0609 and was determined to have very low risk significance because it did not impact any of the five shutdown safety functions identified by NUMARC 91-06.
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| Inspection Report# : 2000017(pdf)
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| Significance: Sep 30, 2000 Identified By: NRC Item Type: FIN Finding Human performance errors and an inadequate procedure resulted in exceeding the allowed outage time for the emergency reserve auxiliary transformer static VAR compensator.
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| Human performance errors and the failure to develop an adequate procedure for the emergency reserve auxiliary transformer static VAR (Volt Ampere Reactive) compensator (ERAT-SVC) surveillance test resulted in several delays during the test. These delays caused the work to not be completed within the allowed outage time. Therefore, a request for Enforcement Discretion was presented to the NRC which was formally granted on September 20, 2000 (NOED 00-6-011). The safety significance of this finding was very low because all other emergency core cooling system trains (automatic depressurization system, low pressure core spray, and low pressure core injection), emergency diesel generators, and the reactor core isolation cooling system were operable.
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| Inspection Report# : 2000015(pdf)
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| Significance: May 17, 2000 Identified By: Self Disclosing Item Type: FIN Finding Manual reactor shutdown A labeling discrepancy contributed to the improper isolation of a protective relay for the 4.16kV Bus 1B Reserve Feed Breaker. As a result, during functional testing, the relay actuated and caused the bus to be de-energized which ultimately resulted in a manual reactor shut down. This issue was determined to be of very low risk significance due to remaining mitigation capability and recovery potential.
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| Inspection Report# : 2000008(pdf)
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| Mitigating Systems Significance: N/A Dec 29, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW 10 CFR 55.59(c)(5) REQUIREMENTS FOR RETAINING LICENSED OPERATOR REQUALIFICATION PROGRAM RECORDS The inspectors identified a Non-Cited Violation wherein the facility licensee had failed to follow the Code of Federal Regulations (CFR) Title 10, Part 55.59(c)(5), Records, requirements by failing to systematically retain all of the original or authenticated copies of the original evaluation documents during the year 2000 annual NRC examination. The finding was of very low safety significance because although the records were not the original or authenticated copies of the original, records did exist in computerized clerically transcribed documents. The computer records had not been signed, and there was no indication that they had been verified correct by the original authors. The unauthenticated documents did provide information that licensed operators, for the most part, had participated and were evaluated during the year 2000 NRC annual requalification examination. However, the inspectors determined that the finding was more than minor. Specifically, the inspectors identified at least one instance in which the transcribed information appeared to be incorrect or missing. The records failure had credible impact on safety, in that, it negatively impacted on the intent of the licensed operator requalification examination process which, in part, is to maintain a high level of confidence that licensed operators continue to possess the requisite knowledge and abilities needed to safely perform licensed duties. In addition, inadequate records keeping adversely affects the NRC's ability to regulate.
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| Inspection Report# : 2001015(pdf)
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| Significance: Jan 26, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct longstanding Reactor Core Isolation Cooling (RCIC) System valve degradation
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| 1Q/2002 Inspection Findings - Clinton Page 3 of 8 Corrective actions for a longstanding deficiency with the Reactor Core Isolation Cooling (RCIC) system steam bypass valve were not effective in stopping the leakage past the valve. This finding was determined to be a Non-Cited Violation of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action." This finding was determined to have very low risk significance because the degraded condition of the valve did not affect the operability of the RCIC system.
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| Inspection Report# : 2001002(pdf)
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| Significance: Jan 26, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow Condition Report process for Shutdown Service Water (SX) pipe wall thinning Corrective actions were not implemented to replace a portion of the shutdown service water (SX) system piping after pipe wall thinning was identified. The failure to take the specified corrective actions by the committed due date or to properly reevaluate the degraded condition was determined to be a Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, "Procedures." This finding was determined to have very low safety significance because the SX system remained operable and capable of performing its' safety function.
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| Inspection Report# : 2001002(pdf)
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| Significance: Sep 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation Procedural requirements were not followed when unexpected equipment response was encountered.
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| Maintenance personnel failed to appropriately follow procedure instructions during testing of the Division III emergency diesel generator room fire detection system. These actions led to the emergency diesel generator being rendered inoperable. The procedure violation was treated as a Non-Cited Violation. This issue was of very low safety significance since the other divisional emergency diesel generators and all emergency core cooling systems were operable at the time of discovery.
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| Inspection Report# : 2000015(pdf)
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| Significance: Jun 16, 2000 Identified By: NRC Item Type: NCV NonCited Violation The licensee failed to ensure that appropriate post-modification testing was specified and accomplished for the Division I and Division III EDG output breaker circuitry modifications The licensee failed to ensure that the appropriate post-modification testing (PMT) was specified in the Division I and Division III emergency diesel generator (EDG) output breaker circuitry modification packages and that the post-modification tests were correctly accomplished. This was required to demonstrate through component and functional testing that the modified (rewired) portions of the Division I and Division III EDG output breaker circuitry were adequately installed to accomplish the intent of the plant design changes.
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| Inspection Report# : 2000012(pdf)
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| Barrier Integrity Significance: Nov 16, 2000 Identified By: NRC Item Type: NCV NonCited Violation Secondary containment was inoperable for 6 minutes during fuel movements when interlock doors were opened.
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| Secondary containment was inoperable for 6 minutes during fuel movements when secondary containment interlock doors were inadvertently opened to move scaffolding. The inoperability was discovered when operators in the control room received an alarm indicating a loss of secondary containment vacuum. One Non-Cited Violation was identified for violating Technical Specification 3.6.4.1 which requires secondary containment operability during fuel moves. This finding was evaluated using the shutdown significance determination process contained in Appendix G of IMC 0609 and was determined to have very low risk significance because it did not meet the criteria for findings requiring a phase 2 significance evaluation.
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| Inspection Report# : 2000017(pdf)
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| Significance: Nov 14, 2000
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| 1Q/2002 Inspection Findings - Clinton Page 4 of 8 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform radiographic examinations of Class 2 welds.
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| The inspectors identified a Non-Cited Violation for the failure to perform radiographic examinations of Class 2 welds in compliance with applicable American Society of Mechanical Engineers (ASME) Code requirements. During installation testing of the 1999 Feedwater Keep Fill FW-39 modification, five radiographic examinations had recorded geometric unsharpness values which exceeded Section III and Section V ASME Code limits. Radiographic geometric unsharpness values are used to ensure that the film is of adequate quality to see defects. In addition, inspectors identified that three examinations did not meet Section V Code requirements for documentation of radiographic technique variables which can affect the image quality of the film. The safety significance of this issue was considered very low at this time, based on the absence of adverse consequences, the presence of other image quality indicators, and because the issue did not involve the system isolation valves. The failure to comply with ASME Code radiographic examination requirements could result in the failure to detect flaws within reactor coolant boundary piping, and was considered a Non-Cited Violation of 10 CFR Part 50.55a, "Codes and Standards".
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| Inspection Report# : 2000019(pdf)
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| Emergency Preparedness Significance: Aug 21, 2001 Identified By: NRC Item Type: NCV NonCited Violation Violation of 10 CFR 50.54(q) re. SCBA qualifications A Non-Cited Violation of 10 CFR 50.54(q) was identified by the NRC associated with the failure to maintain personnel qualifications for self contained breathing apparatus in accordance with the Clinton Power Station Emergency Plan. The finding was of very low safety significance because the licensee maintained an adequate number of qualified personnel to maintain minimum coverage of the required positions identified in the Emergency Plan.
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| Inspection Report# : 2001010(pdf)
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| Significance: Jun 08, 2001 Identified By: NRC Item Type: VIO Violation Supplemental Inspection -- Failure to correct self-identified defficiencies disclosed through control room communications drills This supplemental inspection was performed by the NRC to assess the licensee's evaluation associated with inaccuracies in the reporting of the Drill and Exercise Performance (DEP) performance indicator and with the performance deficiencies that resulted in a White DEP performance indicator (fourth quarter 1999 through the fourth quarter 2000). During the inspection, performed in accordance with NRC Inspection Procedure 95001, the inspector concluded that the licensee performed an adequate evaluation to determine the causes of both issues. In the case of the performance indicator errors, the licensee performed a root cause evaluation which identified a personnel error that was compounded by the lack of self-checking and verification. In addition, the licensee identified contributing causes that included the failure to provide adequate training to the emergency preparedness staff and the failure to provide adequate procedural guidance to the performance indicator data stewards and verifiers, which also applied to performance indicators in other cornerstones. The inspector concluded that the scope of corrective actions planned and implemented by the licensee appeared to address the identified causes. However, the inspector observed an additional discrepancy in the recently completed performance indicator evaluation related to drill and exercise participation. In addition, the licensee identified an error in its evaluation of one of the other emergency preparedness performance indicators that was not detected during its evaluation. These observations demonstrated weaknesses in the licensee's corrective actions and extent of condition review. The errors in the licensee's reporting of the DEP performance indicator was significant, in that the error resulted in a change of color, (i.e., Green-to-White). Consequently, a violation of 10 CFR 50.9 of more than minor safety significance was identified. Since the inaccurate reporting occurred during the period that the NRC's Enforcement Policy afforded discretion for the non-willful submittal of inaccurate performance indicator information, the NRC is exercising enforcement discretion and not citing the violation. In the case of the White DEP performance indicator, the inspector concluded that the licensee adequately assessed the deficiencies that led to the performance issues. Based on its review, the licensee attributed the White performance indicator to the high failure rate of control room communicator drills (i.e., job performance measures). The licensee identified two apparent causes for the high failure rate: (1) weaknesses in formal training; and (2) failure to meet emergency preparedness management expectations concerning the identification and correction of drill deficiencies. The inspector reviewed the licensee's corrective actions and determined that they addressed the causes identified. As a result of the licensee's immediate corrective actions, the licensee's performance returned the performance indicator to the Green band. The inspector and the licensee concluded that the high failure rate of the control room communicators resulted, in part, from inadequate corrective actions for self-identified deficiencies. Specifically, the licensee control room communicator drills were a portion of an overall annual evaluation of non-licensed operators, which included non-emergency preparedness functions. Generally, the failure of the communications segment of the evaluation did not result in a total failure of the annual evaluation. Therefore, the licensee's remedial actions were limited and were not effective in correcting the deficiencies and
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| 1Q/2002 Inspection Findings - Clinton Page 5 of 8 preventing similar failures from occurring, as required by 10 CFR 50.47(b)(14). By letter dated 08/22/01, the NRC concluded that a violation of 10 CFR 50.47(b)(14) had occurred and using the NRC's significance determination process, determined that the finding was white.
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| Inspection Report# : 2001009(pdf)
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| Significance: Feb 23, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to follow emergency plan for on-shift staffing For an approximate 2-month time period, the licensee failed to meet one of the minimum on-shift emergency response organization (ERO) staffing requirements contained in Table 2-1 of the licensee's emergency plan.
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| Inspection Report# : 2001003(pdf)
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| Significance: N/A Apr 28, 2000 Identified By: NRC Item Type: FIN Finding Emergency Preparedness Performance Indicator Verification Alert and Notification System, Drill & Exercise Participation, and Drill & Exercise performance indicators: The inspectors verified that the licensee had acceptably gathered information and reported these three performance indicators, which were in the green band, with the following minor exception. The inspectors identified a discrepancy with the licensee's initial assessment of the Drill and Exercise Performance (DEP) indicator related to the number of performance opportunities associated with a General Emergency declaration during a drill or an exercise. The licensee initially assumed that only three performance opportunities would exist rather than four as provided in NEI 99-02, but later recognized that they had misinterpreted the guidance. This did not affect the DEP performance indicator which was in the green band.
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| Inspection Report# : 2000009(pdf)
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| Occupational Radiation Safety Significance: Oct 08, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an Adequate Survey to identify and to post a High Radiation Area A finding and associated Non-Cited Violation was identified concerning the failure to perform an adequate radiological survey, as required by 10 CFR 20.1501. Although the licensee identified this issue, the licensee did not thoroughly evaluate the cause(s) of the unanticipated radiological conditions and associated problems in the monitoring of radioactive waste activities, which have resulted in previous, similar incidents. The finding was of very low safety significance because the area radiation levels and the licensee's additional administrative barriers would have limited the potential for an individual inadvertently entering the area and receiving a radiation exposure in excess of regulatory limits.
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| Inspection Report# : 2001015(pdf)
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| Significance: Aug 21, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to maintain administrative control of high radiation area keys as required by Technical Specification 5.7.2 Technical Specification 5.7.2 requires, in part, that doors to high radiation areas in which an individual could receive a deep dose equivalent greater than or equal to 1000 millirem in one hour (at 30 centimeters) shall be provided with locked or continuously guarded doors to prevent unauthorized entry and that the keys to such doors shall be administratively controlled. During October 29
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| - 31, 2001, the licensee failed to maintain administrative control of a key that controlled five access points to high radiation areas specified above (i.e., lost the key and failed to perform required key inventories to identify its loss), as described in CR No. 2-00 016. Since the inspector concluded that sufficient barriers remained to prevent an unauthorized individual from entering the affected areas and receiving an overexposure, the inspector concluded that the incident was of very low safety significance. The licensee also reported the incident to the NRC as an occurrence for the Occupational Exposure Control Effectiveness performance indicator.
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| This is being treated as a Non-Cited Violation.
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| Inspection Report# : 2001010(pdf)
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| Significance: SL-IV Jul 27, 2001 Identified By: NRC
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| 1Q/2002 Inspection Findings - Clinton Page 6 of 8 Item Type: NCV NonCited Violation Misuse of Radioactive Material to Alarm a PCM Radiation protection technician used contaminated material to alarm a portal contamination monitor (PCM), while an individual was performing a contamination survey. Based on the licensee's investigation, the contamination was not placed on the individual, and the individual successfully monitored through an additional PCM. This incident will be reviewed by the NRC for potential enforcement actions. Update: On July 27, 2001, the NRC identified and forwarded to the licensee (by letter) a Non-Cited Violation of the Clinton Station Facility Operating License associated with the deliberate misuse of radioactive material by a junior contract radiation protection technician. On October 20, 2000, the technician misused radioactive material to cause an erroneous alarm on a PCM, as another individual was performing a contamination survey. The licensee identified the incident, entered the incident into its corrective action program, and implemented immediate corrective actions. Since the violation was determined to be willful, the NRC did not assign a significance to the violation using the NRC's Significance Determination Process. In accordance with the NRC Enforcement Policy, the NRC determined that the incident constituted a Severity Level IV violation of the Clinton Power Station Facility Operating License. Further, the NRC determined that the violation met the criteria necessary to disposition the violation as a Non-Cited Violation (Section VI.A.1.d of the NRC Enforcement Policy).
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| Inspection Report# : 2000018(pdf)
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| Inspection Report# : 2001010(pdf)
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| Significance: Oct 25, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Three individuals entered a HRA in violation of Technical Specification 5.7.1 On October 25, 2000, three individuals entered the B residual heat removal heat exchanger room (a posted high radiation area);
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| however, the individuals were not working under a radiation work permit that allowed entry into the high radiation area and did not satisfy either of the three entry conditions of Technical Specification 5.7.1.
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| Inspection Report# : 2000018(pdf)
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| Public Radiation Safety Significance: Dec 08, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Inadvertent Release of Radioactive Material to Unrestricted Area During September 2000, the licensee conducted a survey of tools, equipment, etc. outside of the restricted area (protected area and owner controlled area) and identified low-level contaminated materials that were not under constant surveillance or control. The failure to maintain contstant surveillance and control of the material was a violation of 10 CFR 20.1802 and was characterized as a Non-Cited Violation. Based on the licensee's conservative annual dose assessment (about 1.56 millirem) and the inability to define the origin of each of the items, the inspector concluded that the issue constituted one occurrence/event per the NRC Significance Determination Process (Green).
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| Inspection Report# : 2000021(pdf)
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| Physical Protection Miscellaneous Significance: Mar 31, 2002 Identified By: NRC Item Type: NCV NonCited Violation Non-Cited violation of T.S. 5.4.1. for inadequate operating procedure, resulting in ERAT-SVC breaker trip.
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| Inspection Report# : 2002005(pdf)
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| 1Q/2002 Inspection Findings - Clinton Page 7 of 8 Significance: Feb 17, 2002 Identified By: NRC Item Type: NCV NonCited Violation Non-cited violation of T.S. 5.4.1 for an inadequate surveillance procedure.
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| Inspection Report# : 2001016(pdf)
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| Significance: Feb 17, 2002 Identified By: NRC Item Type: FIN Finding A temporary modification on the "A" RR FCV control cirucuitry.
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| On December 14, 2001, the licensee installed a temporary modification on the "A" RR FCV control circuitry. The T-mod was installed to assist the operators in manually controlling the "A" RR FCV because the reliability of the normal control circuitry was in question.
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| During the implementation portion of the T-mod installations, the "A" RR FCV unexpectedly moved from 94 percent open to 102 percent open at which point the protective position circuitry locked the valve at the 102 percent position. Recator power was observed to go from 94 percent to 98 percent during this unexpected valve movement. Following this unexpected FCV movement, opeerations personnel ordered the T-mod to be removed and operators then proceeded to manually shut down the reactor without any further movements of the "A" RR FCV.
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| Inspection Report# : 2001016(pdf)
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| Significance: Feb 15, 2002 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Corrective Actions for Repetitive Failure of SLC pump motor breaker.
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| Inspection Report# : 2002003(pdf)
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| Significance: SL-IV Aug 18, 2001 Identified By: NRC Item Type: VIO Violation Falsification of Test Records by Licensee Employee SL IV - On July 2, 2001, by separate letter, NRC issued a Severity Level IV violation of 10 CFR 50.9 for a deliberate falsification by a plant test engineer. Following investigation by the Office of Investigations, NRC determined that, on October 20, 2000, a test engineer forged another employee's signature on two test package cover sheets on by forging another employee's signature without his prior concurrence, in violation of Clinton established plant protocol and procedure.
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| Inspection Report# : 2001010(pdf)
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| Significance: SL-IV Apr 06, 2001 Identified By: NRC Item Type: VIO Violation Violation of 10 CFR 50.7 "Employee Protection" On April 6, 2001, the NRC issued the licensee a Severity Level IV Violation of 10 CFR 50.7. The NRC concluded that the licensee took adverse employment actions against an employee in the licensee's Nuclear Training Department (i.e., unfavorable 1999 performance review), in part, as a result of the employee's engagement in protected activities. In addition, the NRC learned that several training personnel may be reluctant to discuss department issues within the nuclear training department.
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| Inspection Report# : 2001010(pdf)
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| Inspection Report# : 2001007(pdf)
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| Significance: N/A Jan 26, 2001 Identified By: NRC Item Type: FIN Finding Assessment of Problem Identification and Resolution Performance The team identified that the licensee appropriately entered significant plant issues into the corrective action process by initiating condition reports. Some less significant conditions adverse to quality were evaluated and corrected outside the established process.
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| The trending program was not fully effective as a problem identification tool. Quality Assurance audits and self-assessments reviewed varied in quality. Identified issues were generally evaluated properly, although in several cases the corrective action process did not work effectively to either evaluate or prioritize issues. Current station performance issues including human performance, corrective action program, surveillance testing, and labeling indicate that long term corrective actions previously taken in these areas as restart and post-restart initiatives have not been fully effective to support sustained improvement. Corrective actions were not always fully effective or timely for some individual equipment issues and the effectiveness review process (CARE) did not always identify ineffective corrective actions. The licensee had recently recognized similar deficiencies in corrective action
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| 1Q/2002 Inspection Findings - Clinton Page 8 of 8 program implementation but had not yet fully developed or completed the corrective actions to improve these areas. The inspectors did not find any reluctance by the station employees to raise safety issues.
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| Inspection Report# : 2001002(pdf)
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| Significance: N/A Dec 31, 2000 Identified By: NRC Item Type: FIN Finding Recent human performance issues have occurred which are associated with operator performance and knowledge based deficiencies that have affected plant operations and responses to transient conditions.
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| Recent human performance issues have occurred which are associated with operator performance and knowledge based deficiencies that have affected plant operations and responses to transient conditions. While the risk of the individual events was very low, the failure of operators to adequately control level parameters indicated a declining trend in this area. These issues could not be easily evaluated by present risk analysis methods because failures to follow procedures and maintaining management expectations were not modeled in the Clinton Individual Plant Evaluation. Therefore, the finding is characterized as having no color.
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| Inspection Report# : 2000020(pdf)
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| Significance: N/A Nov 14, 2000 Identified By: NRC Item Type: FIN Finding Three procedures were not written in compliance with the applicable ASME Code.
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| The inspectors reviewed three special process procedures, and identified areas where all three procedures were not written in compliance with the applicable ASME Code. The procedure deficiencies had the potential to affect the ASME Code compliance of weld fabrication and nondestructive examination used on safety-related components and piping. The inspectors noted that each of the ASME Code problems identified contained elements of human performance deficiencies. The human performance aspects, while not always being the root cause of the problem, were significant contributors leading to procedure deficiencies. While the risk of the individual examples was very low, the number of deficiencies indicated a problem with incorporation of applicable ASME Code requirements into special process procedures.
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| Inspection Report# : 2000019(pdf)
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| Significance: N/A Sep 30, 2000 Identified By: NRC Item Type: FIN Finding Recent events affecting plant operations contained elements of human performance deficiencies.
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| NO COLOR. The inspectors noted that several recent events which have affected plant operations and the operability of safety-related components or other components important to safety contained elements of human performance deficiencies. The human performance aspects, while not always being the root cause of the problem, were significant contributors leading to the events.
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| While the risk of the individual events was very low, the number of maintenance-related incidents indicated a problem exists with the control, review, and performance of maintenance activities.
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| Inspection Report# : 2000015(pdf)
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| Significance: N/A May 20, 2000 Identified By: NRC Item Type: FIN Finding Inaccurate historical data for the Safety System Functional Failure Indicator No Color. The licensee identified a failure to submit accurate information to the NRC. The inaccurate information involved the historical data submittal for the Safety System Functional Failure Performance Indicator. The error resulted in a response band color change from Green to White for the first quarter 1999 Performance Indicator. The NRC exercised Enforcement Discretion pursuant to Section VII.B.6 of the Enforcement Policy and did not cite the violation.
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| Inspection Report# : 2000008(pdf)
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| Last modified : July 22, 2002
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| 2Q/2002 Inspection Findings - Clinton Page 1 of 13 Clinton Initiating Events Significance: Aug 21, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow procedures associated with feed water level control system surveillance testing.
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| Human performance and corrective action deficiencies contributed to a Non-Cited Violation of Technical Specification 5.4.1 for failing to follow procedures. This led to the unplanned automatic reactor shutdown on July 24, 2001. The finding was of very low safety significance because no complications occurred during the unplanned automatic reactor shut down and the finding did not increase the likelihood of mitigation equipment being unavailable.
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| Inspection Report# : 2001010(pdf)
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| Significance: Dec 31, 2000 Identified By: NRC Item Type: NCV NonCited Violation Operators failed to adequately evaluate an alarming moisture separator drain tank level annunciator that resulted in a turbine trip.
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| During plant restart following refueling outage 7, operators did not adequately evaluate an alarming moisture separator drain tank level annunciator. As a result, high water level in the moisture separator drain tank caused a turbine trip with the reactor at approximately 25% power. The inspectors reviewed this issue using the significance determination process for a transient. Since only the initiating event cornerstone is affected and associated assumptions have no other impact than slightly increasing the likelihood of an uncomplicated reactor trip, the finding is considered to be of very low safety significance.
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| Inspection Report# : 2000020(pdf)
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| Significance: Dec 31, 2000 Identified By: NRC Item Type: NCV NonCited Violation Operators failed to adequately control reactor vessel water level and pressure following the automatic reactor scram which resulted in a second automatic scram.
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| Operators failed to adequately control reactor vessel water level and pressure, while attempting to open the main steam isolation valves following the automatic reactor scram on December 18, 2000. This resulted in an automatic scram signal due to low reactor vessel water level. The inspectors reviewed this issue using the significance determination process for a transient with a loss of feedwater and determined this was of very low safety significance because the event occurred while the reactor was shut down and all control rods were already fully inserted.
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| Inspection Report# : 2000020(pdf)
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| Significance: Dec 31, 2000 Identified By: NRC Item Type: FIN Finding file://C:\RROP\NRR\OVERSIGHT\ASSESS\CLIN\clin_pim.html 07/03/2003
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| 2Q/2002 Inspection Findings - Clinton Page 2 of 13 Operators did not adequately control reactor vessel inventory after a reactor scram which resulted in the motor driven reactor feedwater pump tripping on high reactor vessel water level.
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| During operator response to a reactor scram on December 18, 2000, operators did not adequately control reactor vessel inventory prior to the motor driven reactor feedwater pump tripping on high reactor vessel water level. The inspectors reviewed this issue using the significance determination process for a transient with a loss of feedwater and determined this was of very low safety significance because all other reactor vessel level control systems were operable and functioned as designed.
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| Inspection Report# : 2000020(pdf)
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| Significance: Nov 16, 2000 Identified By: NRC Item Type: NCV NonCited Violation An alternate rod insertion system initiation and a manual reactor scram occurred with the reactor shutdown as a result of an inadequate maintenance procedure.
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| During replacement of power supplies for the alternate rod insertion (ARI) system, maintenance personnel failed to fully evaluate the impacts that re-energizing the power supplies had on the ARI initiation logic. While re-energizing the power supplies, the initiation logic sensed an ARI signal (low reactor water level). This caused the vent and drain valves to close and the scram discharge volume to fill with water. Plant operators inserted a manual scram signal before the automatic high scram discharge volume set point was reached. One Non-Cited Violation was identified for having an inadequate maintenance procedure to control this activity. This finding was evaluated using the shutdown significance determination process contained in Appendix G of IMC 0609 and was determined to have very low risk significance because it did not impact any of the five shutdown safety functions identified by NUMARC 91-06.
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| Inspection Report# : 2000017(pdf)
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| Significance: Sep 30, 2000 Identified By: NRC Item Type: FIN Finding Human performance errors and an inadequate procedure resulted in exceeding the allowed outage time for the emergency reserve auxiliary transformer static VAR compensator.
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| Human performance errors and the failure to develop an adequate procedure for the emergency reserve auxiliary transformer static VAR (Volt Ampere Reactive) compensator (ERAT-SVC) surveillance test resulted in several delays during the test. These delays caused the work to not be completed within the allowed outage time. Therefore, a request for Enforcement Discretion was presented to the NRC which was formally granted on September 20, 2000 (NOED 00-6-011). The safety significance of this finding was very low because all other emergency core cooling system trains (automatic depressurization system, low pressure core spray, and low pressure core injection), emergency diesel generators, and the reactor core isolation cooling system were operable.
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| Inspection Report# : 2000015(pdf)
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| Significance: May 17, 2000 Identified By: Self Disclosing Item Type: FIN Finding Manual reactor shutdown A labeling discrepancy contributed to the improper isolation of a protective relay for the 4.16kV Bus 1B Reserve Feed Breaker. As a result, during functional testing, the relay actuated and caused the bus to be de-energized which ultimately resulted in a manual reactor shut down. This issue was determined to be of very low risk significance due to remaining mitigation capability and recovery potential.
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| Inspection Report# : 2000008(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\CLIN\clin_pim.html 07/03/2003
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| 2Q/2002 Inspection Findings - Clinton Page 3 of 13 Mitigating Systems Significance: N/A Dec 29, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW 10 CFR 55.59(c)(5) REQUIREMENTS FOR RETAINING LICENSED OPERATOR REQUALIFICATION PROGRAM RECORDS The inspectors identified a Non-Cited Violation wherein the facility licensee had failed to follow the Code of Federal Regulations (CFR) Title 10, Part 55.59(c)(5), Records, requirements by failing to systematically retain all of the original or authenticated copies of the original evaluation documents during the year 2000 annual NRC examination.
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| The finding was of very low safety significance because although the records were not the original or authenticated copies of the original, records did exist in computerized clerically transcribed documents. The computer records had not been signed, and there was no indication that they had been verified correct by the original authors. The unauthenticated documents did provide information that licensed operators, for the most part, had participated and were evaluated during the year 2000 NRC annual requalification examination. However, the inspectors determined that the finding was more than minor. Specifically, the inspectors identified at least one instance in which the transcribed information appeared to be incorrect or missing. The records failure had credible impact on safety, in that, it negatively impacted on the intent of the licensed operator requalification examination process which, in part, is to maintain a high level of confidence that licensed operators continue to possess the requisite knowledge and abilities needed to safely perform licensed duties. In addition, inadequate records keeping adversely affects the NRC's ability to regulate.
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| Inspection Report# : 2001015(pdf)
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| Significance: TBD Sep 20, 2001 Identified By: NRC Item Type: URI Unresolved item SX Spent fuel Pool Makeup Line Flow Function Not Confirmed Inspection Report# : 2002006(pdf)
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| Inspection Report# : 2001011(pdf)
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| Significance: Jan 26, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct longstanding Reactor Core Isolation Cooling (RCIC) System valve degradation Corrective actions for a longstanding deficiency with the Reactor Core Isolation Cooling (RCIC) system steam bypass valve were not effective in stopping the leakage past the valve. This finding was determined to be a Non-Cited Violation of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action." This finding was determined to have very low risk significance because the degraded condition of the valve did not affect the operability of the RCIC system.
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| Inspection Report# : 2001002(pdf)
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| Significance: Jan 26, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow Condition Report process for Shutdown Service Water (SX) pipe wall thinning Corrective actions were not implemented to replace a portion of the shutdown service water (SX) system piping after file://C:\RROP\NRR\OVERSIGHT\ASSESS\CLIN\clin_pim.html 07/03/2003
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| 2Q/2002 Inspection Findings - Clinton Page 4 of 13 pipe wall thinning was identified. The failure to take the specified corrective actions by the committed due date or to properly reevaluate the degraded condition was determined to be a Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, "Procedures." This finding was determined to have very low safety significance because the SX system remained operable and capable of performing its' safety function.
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| Inspection Report# : 2001002(pdf)
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| Significance: Sep 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation Procedural requirements were not followed when unexpected equipment response was encountered.
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| Maintenance personnel failed to appropriately follow procedure instructions during testing of the Division III emergency diesel generator room fire detection system. These actions led to the emergency diesel generator being rendered inoperable. The procedure violation was treated as a Non-Cited Violation. This issue was of very low safety significance since the other divisional emergency diesel generators and all emergency core cooling systems were operable at the time of discovery.
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| Inspection Report# : 2000015(pdf)
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| Significance: Jun 16, 2000 Identified By: NRC Item Type: NCV NonCited Violation The licensee failed to ensure that appropriate post-modification testing was specified and accomplished for the Division I and Division III EDG output breaker circuitry modifications The licensee failed to ensure that the appropriate post-modification testing (PMT) was specified in the Division I and Division III emergency diesel generator (EDG) output breaker circuitry modification packages and that the post-modification tests were correctly accomplished. This was required to demonstrate through component and functional testing that the modified (rewired) portions of the Division I and Division III EDG output breaker circuitry were adequately installed to accomplish the intent of the plant design changes.
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| Inspection Report# : 2000012(pdf)
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| Barrier Integrity Significance: Nov 16, 2000 Identified By: NRC Item Type: NCV NonCited Violation Secondary containment was inoperable for 6 minutes during fuel movements when interlock doors were opened.
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| Secondary containment was inoperable for 6 minutes during fuel movements when secondary containment interlock doors were inadvertently opened to move scaffolding. The inoperability was discovered when operators in the control room received an alarm indicating a loss of secondary containment vacuum. One Non-Cited Violation was identified for violating Technical Specification 3.6.4.1 which requires secondary containment operability during fuel moves. This finding was evaluated using the shutdown significance determination process contained in Appendix G of IMC 0609 and was determined to have very low risk significance because it did not meet the criteria for findings requiring a phase 2 significance evaluation.
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| Inspection Report# : 2000017(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\CLIN\clin_pim.html 07/03/2003
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| 2Q/2002 Inspection Findings - Clinton Page 5 of 13 Significance: Nov 14, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform radiographic examinations of Class 2 welds.
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| The inspectors identified a Non-Cited Violation for the failure to perform radiographic examinations of Class 2 welds in compliance with applicable American Society of Mechanical Engineers (ASME) Code requirements. During installation testing of the 1999 Feedwater Keep Fill FW-39 modification, five radiographic examinations had recorded geometric unsharpness values which exceeded Section III and Section V ASME Code limits. Radiographic geometric unsharpness values are used to ensure that the film is of adequate quality to see defects. In addition, inspectors identified that three examinations did not meet Section V Code requirements for documentation of radiographic technique variables which can affect the image quality of the film. The safety significance of this issue was considered very low at this time, based on the absence of adverse consequences, the presence of other image quality indicators, and because the issue did not involve the system isolation valves. The failure to comply with ASME Code radiographic examination requirements could result in the failure to detect flaws within reactor coolant boundary piping, and was considered a Non-Cited Violation of 10 CFR Part 50.55a, "Codes and Standards".
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| Inspection Report# : 2000019(pdf)
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| Emergency Preparedness Significance: Aug 21, 2001 Identified By: NRC Item Type: NCV NonCited Violation Violation of 10 CFR 50.54(q) re. SCBA qualifications A Non-Cited Violation of 10 CFR 50.54(q) was identified by the NRC associated with the failure to maintain personnel qualifications for self contained breathing apparatus in accordance with the Clinton Power Station Emergency Plan.
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| The finding was of very low safety significance because the licensee maintained an adequate number of qualified personnel to maintain minimum coverage of the required positions identified in the Emergency Plan.
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| Inspection Report# : 2001010(pdf)
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| Significance: Jun 08, 2001 Identified By: NRC Item Type: VIO Violation Supplemental Inspection -- Failure to correct self-identified defficiencies disclosed through control room communications drills This supplemental inspection was performed by the NRC to assess the licensee's evaluation associated with inaccuracies in the reporting of the Drill and Exercise Performance (DEP) performance indicator and with the performance deficiencies that resulted in a White DEP performance indicator (fourth quarter 1999 through the fourth quarter 2000). During the inspection, performed in accordance with NRC Inspection Procedure 95001, the inspector concluded that the licensee performed an adequate evaluation to determine the causes of both issues. In the case of the performance indicator errors, the licensee performed a root cause evaluation which identified a personnel error that was compounded by the lack of self-checking and verification. In addition, the licensee identified contributing causes that included the failure to provide adequate training to the emergency preparedness staff and the failure to provide adequate procedural guidance to the performance indicator data stewards and verifiers, which also applied to performance indicators in other cornerstones. The inspector concluded that the scope of corrective actions planned and file://C:\RROP\NRR\OVERSIGHT\ASSESS\CLIN\clin_pim.html 07/03/2003
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| 2Q/2002 Inspection Findings - Clinton Page 6 of 13 implemented by the licensee appeared to address the identified causes. However, the inspector observed an additional discrepancy in the recently completed performance indicator evaluation related to drill and exercise participation. In addition, the licensee identified an error in its evaluation of one of the other emergency preparedness performance indicators that was not detected during its evaluation. These observations demonstrated weaknesses in the licensee's corrective actions and extent of condition review. The errors in the licensee's reporting of the DEP performance indicator was significant, in that the error resulted in a change of color, (i.e., Green-to-White). Consequently, a violation of 10 CFR 50.9 of more than minor safety significance was identified. Since the inaccurate reporting occurred during the period that the NRC's Enforcement Policy afforded discretion for the non-willful submittal of inaccurate performance indicator information, the NRC is exercising enforcement discretion and not citing the violation. In the case of the White DEP performance indicator, the inspector concluded that the licensee adequately assessed the deficiencies that led to the performance issues. Based on its review, the licensee attributed the White performance indicator to the high failure rate of control room communicator drills (i.e., job performance measures). The licensee identified two apparent causes for the high failure rate: (1) weaknesses in formal training; and (2) failure to meet emergency preparedness management expectations concerning the identification and correction of drill deficiencies.
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| The inspector reviewed the licensee's corrective actions and determined that they addressed the causes identified. As a result of the licensee's immediate corrective actions, the licensee's performance returned the performance indicator to the Green band. The inspector and the licensee concluded that the high failure rate of the control room communicators resulted, in part, from inadequate corrective actions for self-identified deficiencies. Specifically, the licensee control room communicator drills were a portion of an overall annual evaluation of non-licensed operators, which included non-emergency preparedness functions. Generally, the failure of the communications segment of the evaluation did not result in a total failure of the annual evaluation. Therefore, the licensee's remedial actions were limited and were not effective in correcting the deficiencies and preventing similar failures from occurring, as required by 10 CFR 50.47(b)
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| (14). By letter dated 08/22/01, the NRC concluded that a violation of 10 CFR 50.47(b)(14) had occurred and using the NRC's significance determination process, determined that the finding was white.
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| Inspection Report# : 2001009(pdf)
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| Inspection Report# : 2002006(pdf)
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| Significance: Feb 23, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to follow emergency plan for on-shift staffing For an approximate 2-month time period, the licensee failed to meet one of the minimum on-shift emergency response organization (ERO) staffing requirements contained in Table 2-1 of the licensee's emergency plan.
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| Inspection Report# : 2001003(pdf)
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| Significance: N/A Apr 28, 2000 Identified By: NRC Item Type: FIN Finding Emergency Preparedness Performance Indicator Verification Alert and Notification System, Drill & Exercise Participation, and Drill & Exercise performance indicators: The inspectors verified that the licensee had acceptably gathered information and reported these three performance indicators, which were in the green band, with the following minor exception. The inspectors identified a discrepancy with the licensee's initial assessment of the Drill and Exercise Performance (DEP) indicator related to the number of performance opportunities associated with a General Emergency declaration during a drill or an exercise. The licensee initially assumed that only three performance opportunities would exist rather than four as provided in NEI 99-02, but later recognized that they had misinterpreted the guidance. This did not affect the DEP performance indicator which was in the green band.
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| Inspection Report# : 2000009(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\CLIN\clin_pim.html 07/03/2003
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| 2Q/2002 Inspection Findings - Clinton Page 7 of 13 Occupational Radiation Safety Significance: Oct 08, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an Adequate Survey to identify and to post a High Radiation Area A finding and associated Non-Cited Violation was identified concerning the failure to perform an adequate radiological survey, as required by 10 CFR 20.1501. Although the licensee identified this issue, the licensee did not thoroughly evaluate the cause(s) of the unanticipated radiological conditions and associated problems in the monitoring of radioactive waste activities, which have resulted in previous, similar incidents. The finding was of very low safety significance because the area radiation levels and the licensee's additional administrative barriers would have limited the potential for an individual inadvertently entering the area and receiving a radiation exposure in excess of regulatory limits.
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| Inspection Report# : 2001015(pdf)
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| Significance: Aug 21, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to maintain administrative control of high radiation area keys as required by Technical Specification 5.7.2 Technical Specification 5.7.2 requires, in part, that doors to high radiation areas in which an individual could receive a deep dose equivalent greater than or equal to 1000 millirem in one hour (at 30 centimeters) shall be provided with locked or continuously guarded doors to prevent unauthorized entry and that the keys to such doors shall be administratively controlled. During October 29 - 31, 2001, the licensee failed to maintain administrative control of a key that controlled five access points to high radiation areas specified above (i.e., lost the key and failed to perform required key inventories to identify its loss), as described in CR No. 2-00-11-016. Since the inspector concluded that sufficient barriers remained to prevent an unauthorized individual from entering the affected areas and receiving an overexposure, the inspector concluded that the incident was of very low safety significance. The licensee also reported the incident to the NRC as an occurrence for the Occupational Exposure Control Effectiveness performance indicator.
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| This is being treated as a Non-Cited Violation.
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| Inspection Report# : 2001010(pdf)
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| Significance: SL-IV Jul 27, 2001 Identified By: NRC Item Type: NCV NonCited Violation Misuse of Radioactive Material to Alarm a PCM Radiation protection technician used contaminated material to alarm a portal contamination monitor (PCM), while an individual was performing a contamination survey. Based on the licensee's investigation, the contamination was not placed on the individual, and the individual successfully monitored through an additional PCM. This incident will be reviewed by the NRC for potential enforcement actions. Update: On July 27, 2001, the NRC identified and forwarded to the licensee (by letter) a Non-Cited Violation of the Clinton Station Facility Operating License associated with the deliberate misuse of radioactive material by a junior contract radiation protection technician. On October 20, 2000, the technician misused radioactive material to cause an erroneous alarm on a PCM, as another individual was performing a contamination survey. The licensee identified the incident, entered the incident into its corrective action program, and implemented immediate corrective actions. Since the violation was determined to be willful, the NRC did not assign a file://C:\RROP\NRR\OVERSIGHT\ASSESS\CLIN\clin_pim.html 07/03/2003
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| 2Q/2002 Inspection Findings - Clinton Page 8 of 13 significance to the violation using the NRC's Significance Determination Process. In accordance with the NRC Enforcement Policy, the NRC determined that the incident constituted a Severity Level IV violation of the Clinton Power Station Facility Operating License. Further, the NRC determined that the violation met the criteria necessary to disposition the violation as a Non-Cited Violation (Section VI.A.1.d of the NRC Enforcement Policy).
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| Inspection Report# : 2001010(pdf)
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| Inspection Report# : 2000018(pdf)
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| Significance: Oct 25, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Three individuals entered a HRA in violation of Technical Specification 5.7.1 On October 25, 2000, three individuals entered the B residual heat removal heat exchanger room (a posted high radiation area); however, the individuals were not working under a radiation work permit that allowed entry into the high radiation area and did not satisfy either of the three entry conditions of Technical Specification 5.7.1.
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| Inspection Report# : 2000018(pdf)
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| Public Radiation Safety Significance: Dec 08, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Inadvertent Release of Radioactive Material to Unrestricted Area During September 2000, the licensee conducted a survey of tools, equipment, etc. outside of the restricted area (protected area and owner controlled area) and identified low-level contaminated materials that were not under constant surveillance or control. The failure to maintain contstant surveillance and control of the material was a violation of 10 CFR 20.1802 and was characterized as a Non-Cited Violation. Based on the licensee's conservative annual dose assessment (about 1.56 millirem) and the inability to define the origin of each of the items, the inspector concluded that the issue constituted one occurrence/event per the NRC Significance Determination Process (Green).
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| Inspection Report# : 2000021(pdf)
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| Physical Protection Miscellaneous Significance: Jun 30, 2002 Identified By: Self Disclosing Item Type: NCV NonCited Violation Technical Specification 5.4.1 was identified for an inadequate procedure used during the performance of a Division III EDG test.
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| file://C:\RROP\NRR\OVERSIGHT\ASSESS\CLIN\clin_pim.html 07/03/2003
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| 2Q/2002 Inspection Findings - Clinton Page 9 of 13 Green. A Non-Cited Violation of Technical Specification 5.4.1 was identified for an inadequate procedure used during the performance of a Division III (Div-III) emergency diesel generator (EDG) test. Errors in the procedure led to the loss of the Div-III safety-related 4160 Volt electrical bus and unplanned unavailability of the high pressure core spray (HPCS) system. The finding was greater than minor because if left uncorrected, the issue has a credible impact on safety. Further, the issue did have an impact on mitigation system operability as the loss of the Div-III electrical bus rendered the HPCS system inoperable. Using Manual Chapter 0609, "Significance Determination Process," (SDP),
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| Appendix A, phase 1 worksheet, the finding screened out as a very low safety significance issue because the event did not result in the actual loss of safety function for the HPCS system (Section 1R14).
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| Inspection Report# : 2002006(pdf)
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| Significance: Jun 30, 2002 Identified By: Self Disclosing Item Type: NCV NonCited Violation Technical Specification 5.4.1. was identified for workers failing to follow a procedure which contributed to the inadvertent lifting of a double blade guide during fuel movement operations on April 9, Green. A Non-Cited Violation of Technical Specifications (TS) 5.4.1 was identified for workers failing to follow a procedure which contributed to the inadvertent lifting of a double blade guide during fuel movement operations on April 9, 2002. This self-revealing finding was more than minor because if left uncorrected, inadvertent movement of components from the reactor core could lead to a more significant safety concern. Using the fuel barrier column on the SDP Appendix A phase 1 worksheet, the inspectors assessed the finding as a very low safety significance issue (Section 1R20).
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| Inspection Report# : 2002006(pdf)
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| Significance: N/A Jun 30, 2002 Identified By: Licensee Item Type: NCV NonCited Violation Criterion XVI of 10 CFR Part 50 Appendix B "Corrective Actions," requires, in part, that conditions adverse to quality are promptly identified and corrected.
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| Criterion XVI of 10 CFR Part 50 Appendix B "Corrective Actions," requires, in part, that conditions adverse to quality are promptly identified and corrected. Contrary to the above, on April 18, 2002, the licensee identified that ineffective corrective actions had been implemented to address a concern regarding the inadvertent actuation of reactor protection system (RPS) logic during anticipated transient without scram/alternate rod insertion testing. The April 18, 2002 event was similar in nature to an event which occurred during the previous refueling outage in October 2000. This NCV is not greater than Green because the finding involved the actuation of the reactor protection system while the reactor was shut down and the control rods were already in their safety function position (inserted). This NCV is documented in the licensee's Condition Report 113969.
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| Inspection Report# : 2002006(pdf)
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| Significance: N/A Jun 30, 2002 Identified By: Licensee Item Type: NCV NonCited Violation Technical Specification 5.7.2.b see text.
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| Technical Specification 5.7.2.b. requires that areas accessible to personnel with radiation levels greater than 1000 millirem per hour at 12 inches from the radiation source or from any surface which the radiation penetrates requires the doors be locked to prevent unauthorized entry and the keys shall be maintained under the administrative control of the Shift Manager on duty or health physics supervision.
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| Inspection Report# : 2002006(pdf)
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| Significance: N/A Jun 30, 2002 file://C:\RROP\NRR\OVERSIGHT\ASSESS\CLIN\clin_pim.html 07/03/2003
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| 2Q/2002 Inspection Findings - Clinton Page 10 of 13 Identified By: Licensee Item Type: NCV NonCited Violation Technical Specification 5.4.1 requires that written procedures be established, implemented and maintained covering the activities specified in Regulatory Guide 1.33 Appendix A.
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| Regulatory Guide 1.33, Appendix A, Item 7e. 1, requires procedures for access control to radiation areas. Exelon Nuclear procedure RP-CI-462 (Revision 0), Controls for Radiography Activities, Section 5.2.4, requires radiation protection staff to ensure that during radiography, radiation areas be identified and controlled and assure that affected areas are clear or unauthroized personnel.
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| Inspection Report# : 2002006(pdf)
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| Significance: N/A Jun 30, 2002 Identified By: Self Disclosing Item Type: FIN Finding On May 13, 2002 with the reactor at approximately 88 percent rated thermal power, the reactor automatically shutdown due to a high reactor vessel water level signal.
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| Green. A performance deficiency, associated with this automatic reactor shut down on May 13, 2002, was identified as a failure to establish preventative maintenance or inspections on the "B" turbine driven reactor feed pump (TDRFP) for similar conditions found on the "A" TDRFP (noted in December 2000) before a component failure which led to the automatic reactor shut down. This issue was more than minor because if left uncorrected (i.e. appropriate preventive maintenance not being identified and conducted), it could lead to a more significant safety concern and could cause the increased frequency of an initiating event. Consequently, the inspectors evaluated the significance of the issue using the SDP Appendix A phase 1 worksheet. Since the finding contributed only to the likelihood of a reactor trip and did not affect mitigating system availability, the inspectors determined that the finding was of very low safety significance (Section 4OA3b.2).
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| Inspection Report# : 2002006(pdf)
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| Significance: Mar 31, 2002 Identified By: NRC Item Type: NCV NonCited Violation Non-Cited violation of T.S. 5.4.1. for inadequate operating procedure, resulting in ERAT-SVC breaker trip.
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| Green: A Non-Cited Violation of Technical Specification 5.4.1. was identified for an inadequate operating procedure which contributed to an inadvertent emergency reserve auxiliary transformer static-VAR [Volts-Ampere-reactive]-
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| compensator circuit breaker trip. The result of this circuit breaker trip rendered one of the two qualified offsite power sources the transformer inoperable. The finding was of very low safety significance because it could increase the likelihood of an initiating event (reactor trip or a partial loss of offsite power) but did not increase the likelihood that any mitigation equipment would be unavailable.
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| Inspection Report# : 2002005(pdf)
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| Significance: Feb 17, 2002 Identified By: NRC Item Type: NCV NonCited Violation Non-cited violation of T.S. 5.4.1 for an inadequate surveillance procedure.
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| Green. Procedural inadequacies were determined to be a Non-Cited Violation of Technical Specification 5.4.1. These inadequacies led to the "A" residual heat removal system pump being declared operable without performing the appropriate pump supply breaker functionality checks for the conditions. The finding was of very low safety significance because the licensee subsequently tested the "A" residual heat removal system pump supply breaker with satisfactory results. Therefore, system operability was not impacted.
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| Inspection Report# : 2001016(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\CLIN\clin_pim.html 07/03/2003
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| 2Q/2002 Inspection Findings - Clinton Page 11 of 13 Significance: Feb 17, 2002 Identified By: NRC Item Type: FIN Finding A temporary modification on the "A" RR FCV control cirucuitry.
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| On December 14, 2001, the licensee installed a temporary modification on the "A" RR FCV control circuitry. The T-mod was installed to assist the operators in manually controlling the "A" RR FCV because the reliability of the normal control circuitry was in question. During the implementation portion of the T-mod installations, the "A" RR FCV unexpectedly moved from 94 percent open to 102 percent open at which point the protective position circuitry locked the valve at the 102 percent position. Recator power was observed to go from 94 percent to 98 percent during this unexpected valve movement. Following this unexpected FCV movement, opeerations personnel ordered the T-mod to be removed and operators then proceeded to manually shut down the reactor without any further movements of the "A" RR FCV.
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| Inspection Report# : 2001016(pdf)
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| Significance: Feb 15, 2002 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Corrective Actions for Repetitive Failure of SLC pump motor breaker.
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| Green. A Non-Cited Violation (NCV) of 10 CFR 50 Appendix B, Criterion XVI, "Corrective Action," for inadequate corrective action taken to prevent recurrence of a Standby Liquid Control "A" System (SLC) pump motor breaker failure was identified. This finding was determined to be of very low safety significance due to the low initiating event frequency for Anticipated Transient Without Scram, the availability of the "B" SLC pump, and the high likelihood of successful operator recovery actions.
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| Inspection Report# : 2002003(pdf)
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| Significance: SL-IV Aug 18, 2001 Identified By: NRC Item Type: VIO Violation Falsification of Test Records by Licensee Employee SL IV - On July 2, 2001, by separate letter, NRC issued a Severity Level IV violation of 10 CFR 50.9 for a deliberate falsification by a plant test engineer. Following investigation by the Office of Investigations, NRC determined that, on October 20, 2000, a test engineer forged another employee's signature on two test package cover sheets on by forging another employee's signature without his prior concurrence, in violation of Clinton established plant protocol and procedure.
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| Inspection Report# : 2001010(pdf)
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| Significance: SL-IV Apr 06, 2001 Identified By: NRC Item Type: VIO Violation Violation of 10 CFR 50.7 "Employee Protection" On April 6, 2001, the NRC issued the licensee a Severity Level IV Violation of 10 CFR 50.7. The NRC concluded that the licensee took adverse employment actions against an employee in the licensee's Nuclear Training Department (i.e.,
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| unfavorable 1999 performance review), in part, as a result of the employee's engagement in protected activities. In addition, the NRC learned that several training personnel may be reluctant to discuss department issues within the nuclear training department.
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| Inspection Report# : 2001007(pdf)
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| Inspection Report# : 2001010(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\CLIN\clin_pim.html 07/03/2003
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| 2Q/2002 Inspection Findings - Clinton Page 12 of 13 Significance: N/A Jan 26, 2001 Identified By: NRC Item Type: FIN Finding Assessment of Problem Identification and Resolution Performance The team identified that the licensee appropriately entered significant plant issues into the corrective action process by initiating condition reports. Some less significant conditions adverse to quality were evaluated and corrected outside the established process. The trending program was not fully effective as a problem identification tool. Quality Assurance audits and self-assessments reviewed varied in quality. Identified issues were generally evaluated properly, although in several cases the corrective action process did not work effectively to either evaluate or prioritize issues.
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| Current station performance issues including human performance, corrective action program, surveillance testing, and labeling indicate that long term corrective actions previously taken in these areas as restart and post-restart initiatives have not been fully effective to support sustained improvement. Corrective actions were not always fully effective or timely for some individual equipment issues and the effectiveness review process (CARE) did not always identify ineffective corrective actions. The licensee had recently recognized similar deficiencies in corrective action program implementation but had not yet fully developed or completed the corrective actions to improve these areas. The inspectors did not find any reluctance by the station employees to raise safety issues.
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| Inspection Report# : 2001002(pdf)
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| Significance: N/A Dec 31, 2000 Identified By: NRC Item Type: FIN Finding Recent human performance issues have occurred which are associated with operator performance and knowledge based deficiencies that have affected plant operations and responses to transient conditions.
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| Recent human performance issues have occurred which are associated with operator performance and knowledge based deficiencies that have affected plant operations and responses to transient conditions. While the risk of the individual events was very low, the failure of operators to adequately control level parameters indicated a declining trend in this area. These issues could not be easily evaluated by present risk analysis methods because failures to follow procedures and maintaining management expectations were not modeled in the Clinton Individual Plant Evaluation. Therefore, the finding is characterized as having no color.
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| Inspection Report# : 2000020(pdf)
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| Significance: N/A Nov 14, 2000 Identified By: NRC Item Type: FIN Finding Three procedures were not written in compliance with the applicable ASME Code.
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| The inspectors reviewed three special process procedures, and identified areas where all three procedures were not written in compliance with the applicable ASME Code. The procedure deficiencies had the potential to affect the ASME Code compliance of weld fabrication and nondestructive examination used on safety-related components and piping. The inspectors noted that each of the ASME Code problems identified contained elements of human performance deficiencies. The human performance aspects, while not always being the root cause of the problem, were significant contributors leading to procedure deficiencies. While the risk of the individual examples was very low, the number of deficiencies indicated a problem with incorporation of applicable ASME Code requirements into special process procedures.
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| Inspection Report# : 2000019(pdf)
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| Significance: N/A Sep 30, 2000 Identified By: NRC Item Type: FIN Finding Recent events affecting plant operations contained elements of human performance deficiencies.
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| NO COLOR. The inspectors noted that several recent events which have affected plant operations and the operability of safety-related components or other components important to safety contained elements of human performance file://C:\RROP\NRR\OVERSIGHT\ASSESS\CLIN\clin_pim.html 07/03/2003
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| 2Q/2002 Inspection Findings - Clinton Page 13 of 13 deficiencies. The human performance aspects, while not always being the root cause of the problem, were significant contributors leading to the events. While the risk of the individual events was very low, the number of maintenance-related incidents indicated a problem exists with the control, review, and performance of maintenance activities.
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| Inspection Report# : 2000015(pdf)
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| Significance: N/A May 20, 2000 Identified By: NRC Item Type: FIN Finding Inaccurate historical data for the Safety System Functional Failure Indicator No Color. The licensee identified a failure to submit accurate information to the NRC. The inaccurate information involved the historical data submittal for the Safety System Functional Failure Performance Indicator. The error resulted in a response band color change from Green to White for the first quarter 1999 Performance Indicator. The NRC exercised Enforcement Discretion pursuant to Section VII.B.6 of the Enforcement Policy and did not cite the violation.
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| Inspection Report# : 2000008(pdf)
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| Last modified : August 29, 2002 file://C:\RROP\NRR\OVERSIGHT\ASSESS\CLIN\clin_pim.html 07/03/2003
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| 3Q/2002 Inspection Findings - Clinton Page 1 of 11 Clinton Initiating Events Significance: May 13, 2002 Identified By: Self Disclosing Item Type: FIN Finding ON MAY 13, 2002 WITH THE REACTOR AT APPROXIMATELY 88 PERCENT RATED THERMAL POWER, THE REACTOR AUTOMATICALLY SHUTDOWN DUE TO A HIGH REACTOR VESSEL WATER LEVEL SIGNAL.
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| A performance deficiency, associated with this automatic reactor shut down on May 13, 2002, was identified as a failure to establish preventative maintenance or inspections on the "B" turbine driven reactor feed pump (TDRFP) for similar conditions found on the "A" TDRFP (noted in December 2000) before a component failure which led to the automatic reactor shut down. This issue was more than minor because if left uncorrected (i.e. appropriate preventive maintenance not being identified and conducted), it could lead to a more significant safety concern and could cause the increased frequency of an initiating event. Consequently, the inspectors evaluated the significance of the issue using the SDP Appendix A phase 1 worksheet. Since the finding contributed only to the likelihood of a reactor trip and did not affect mitigating system availability, the inspectors determined that the finding was of very low safety significance.
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| Inspection Report# : 2002006(pdf)
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| Significance: Mar 31, 2002 Identified By: NRC Item Type: NCV NonCited Violation NON-CITED VIOLATION OF T.S. 5.4.1. FOR INADEQUATE OPERATING PROCEDURE, RESULTING IN ERAT-SVC BREAKER TRIP.
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| A Non-Cited Violation of Technical Specification 5.4.1. was identified for an inadequate operating procedure which contributed to an inadvertent emergency reserve auxiliary transformer static-VAR [Volts-Ampere-reactive]-
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| compensator circuit breaker trip. The result of this circuit breaker trip rendered one of the two qualified offsite power sources the transformer inoperable. The finding was of very low safety significance because it could increase the likelihood of an initiating event (reactor trip or a partial loss of offsite power) but did not increase the likelihood that any mitigation equipment would be unavailable.
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| Inspection Report# : 2002005(pdf)
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| Significance: Feb 17, 2002 Identified By: NRC Item Type: FIN Finding A TEMPORARY MODIFICATION ON THE "A" RR FCV CONTROL CIRCUITRY.
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| On December 14, 2001, the licensee installed a temporary modification on the "A" RR FCV control circuitry. The T-mod was installed to assist the operators in manually controlling the "A" RR FCV because the reliability of the normal control circuitry was in question. During the implementation portion of the T-mod installations, the "A" RR FCV unexpectedly moved from 94 percent open to 102 percent open at which point the protective position circuitry locked the valve at the 102 percent position. Reactor power was observed to go from 94 percent to 98 percent during this unexpected valve movement. Following this unexpected FCV movement, operations personnel ordered the T-mod to be removed and operators then proceeded to manually shut down the reactor without any further movements of the "A" RR FCV.
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| Inspection Report# : 2001016(pdf)
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| 3Q/2002 Inspection Findings - Clinton Page 2 of 11 Significance: Aug 21, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow procedures associated with feed water level control system surveillance testing.
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| Human performance and corrective action deficiencies contributed to a Non-Cited Violation of Technical Specification 5.4.1 for failing to follow procedures. This led to the unplanned automatic reactor shutdown on July 24, 2001. The finding was of very low safety significance because no complications occurred during the unplanned automatic reactor shut down and the finding did not increase the likelihood of mitigation equipment being unavailable.
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| Inspection Report# : 2001010(pdf)
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| Significance: Dec 31, 2000 Identified By: NRC Item Type: FIN Finding Operators did not adequately control reactor vessel inventory after a reactor scram which resulted in the motor driven reactor feedwater pump tripping on high reactor vessel water level.
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| During operator response to a reactor scram on December 18, 2000, operators did not adequately control reactor vessel inventory prior to the motor driven reactor feedwater pump tripping on high reactor vessel water level. The inspectors reviewed this issue using the significance determination process for a transient with a loss of feedwater and determined this was of very low safety significance because all other reactor vessel level control systems were operable and functioned as designed.
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| Inspection Report# : 2000020(pdf)
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| Significance: Dec 31, 2000 Identified By: NRC Item Type: NCV NonCited Violation Operators failed to adequately evaluate an alarming moisture separator drain tank level annunciator that resulted in a turbine trip.
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| During plant restart following refueling outage 7, operators did not adequately evaluate an alarming moisture separator drain tank level annunciator. As a result, high water level in the moisture separator drain tank caused a turbine trip with the reactor at approximately 25% power. The inspectors reviewed this issue using the significance determination process for a transient. Since only the initiating event cornerstone is affected and associated assumptions have no other impact than slightly increasing the likelihood of an uncomplicated reactor trip, the finding is considered to be of very low safety significance.
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| Inspection Report# : 2000020(pdf)
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| Significance: Dec 31, 2000 Identified By: NRC Item Type: NCV NonCited Violation Operators failed to adequately control reactor vessel water level and pressure following the automatic reactor scram which resulted in a second automatic scram.
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| Operators failed to adequately control reactor vessel water level and pressure, while attempting to open the main steam isolation valves following the automatic reactor scram on December 18, 2000. This resulted in an automatic scram signal due to low reactor vessel water level. The inspectors reviewed this issue using the significance determination process for a transient with a loss of feedwater and determined this was of very low safety significance because the event occurred while the reactor was shut down and all control rods were already fully inserted.
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| Inspection Report# : 2000020(pdf)
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| Significance: Nov 16, 2000
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| 3Q/2002 Inspection Findings - Clinton Page 3 of 11 Identified By: NRC Item Type: NCV NonCited Violation An alternate rod insertion system initiation and a manual reactor scram occurred with the reactor shutdown as a result of an inadequate maintenance procedure.
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| During replacement of power supplies for the alternate rod insertion (ARI) system, maintenance personnel failed to fully evaluate the impacts that re-energizing the power supplies had on the ARI initiation logic. While re-energizing the power supplies, the initiation logic sensed an ARI signal (low reactor water level). This caused the vent and drain valves to close and the scram discharge volume to fill with water. Plant operators inserted a manual scram signal before the automatic high scram discharge volume set point was reached. One Non-Cited Violation was identified for having an inadequate maintenance procedure to control this activity. This finding was evaluated using the shutdown significance determination process contained in Appendix G of IMC 0609 and was determined to have very low risk significance because it did not impact any of the five shutdown safety functions identified by NUMARC 91-06.
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| Inspection Report# : 2000017(pdf)
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| Significance: Sep 30, 2000 Identified By: NRC Item Type: FIN Finding Human performance errors and an inadequate procedure resulted in exceeding the allowed outage time for the emergency reserve auxiliary transformer static VAR compensator.
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| Human performance errors and the failure to develop an adequate procedure for the emergency reserve auxiliary transformer static VAR (Volt Ampere Reactive) compensator (ERAT-SVC) surveillance test resulted in several delays during the test. These delays caused the work to not be completed within the allowed outage time. Therefore, a request for Enforcement Discretion was presented to the NRC which was formally granted on September 20, 2000 (NOED 00-6-011). The safety significance of this finding was very low because all other emergency core cooling system trains (automatic depressurization system, low pressure core spray, and low pressure core injection), emergency diesel generators, and the reactor core isolation cooling system were operable.
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| Inspection Report# : 2000015(pdf)
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| Significance: May 17, 2000 Identified By: Self Disclosing Item Type: FIN Finding Manual reactor shutdown A labeling discrepancy contributed to the improper isolation of a protective relay for the 4.16kV Bus 1B Reserve Feed Breaker. As a result, during functional testing, the relay actuated and caused the bus to be de-energized which ultimately resulted in a manual reactor shut down. This issue was determined to be of very low risk significance due to remaining mitigation capability and recovery potential.
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| Inspection Report# : 2000008(pdf)
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| Mitigating Systems Significance: May 29, 2002 Identified By: Self Disclosing Item Type: NCV NonCited Violation TECHNICAL SPECIFICATION 5.4.1 VIOLATION WAS IDENTIFIED FOR AN INADEQUATE PROCEDURE USED DURING THE PERFORMANCE OF A DIVISION III EDG TEST.
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| A Non-Cited Violation of Technical Specification 5.4.1 was identified for an inadequate procedure used during the performance of a Division III (Div-III) emergency diesel generator (EDG) test. Errors in the procedure led to the loss of the Div-III safety-related 4160 Volt electrical bus and unplanned unavailability of the high pressure core spray (HPCS)
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| 3Q/2002 Inspection Findings - Clinton Page 4 of 11 system. The finding was greater than minor because if left uncorrected, the issue has a credible impact on safety.
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| Further, the issue did have an impact on mitigation system operability as the loss of the Div-III electrical bus rendered the HPCS system inoperable. Using Manual Chapter 0609, "Significance Determination Process," (SDP), Appendix A, phase 1 worksheet, the finding screened out as a very low safety significance issue because the event did not result in the actual loss of safety function for the HPCS system.
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| Inspection Report# : 2002006(pdf)
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| Significance: Feb 17, 2002 Identified By: NRC Item Type: NCV NonCited Violation NON-CITED VIOLATION T.S. 5.4.1 FOR AN INADEQUATE SURVEILLANCE PROCEDURE.
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| Procedural inadequacies were determined to be a Non-Cited Violation of Technical Specification 5.4.1. These inadequacies led to the "A" residual heat removal system pump being declared operable without performing the appropriate pump supply breaker functionality checks for the conditions. The finding was of very low safety significance because the licensee subsequently tested the "A" residual heat removal system pump supply breaker with satisfactory results. Therefore, system operability was not impacted.
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| Inspection Report# : 2001016(pdf)
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| Significance: Feb 15, 2002 Identified By: NRC Item Type: NCV NonCited Violation INDAEQUATE CORRECTIVE ACTIONS FOR REPETITIVE FAILURE OF SLC PUMP MOTOR BREAKER.
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| A Non-Cited Violation (NCV) of 10 CFR 50 Appendix B, Criterion XVI, "Corrective Action," for inadequate corrective action taken to prevent recurrence of a Standby Liquid Control "A" System (SLC) pump motor breaker failure was identified. This finding was determined to be of very low safety significance due to the low initiating event frequency for Anticipated Transient Without Scram, the availability of the "B" SLC pump, and the high likelihood of successful operator recovery actions.
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| Inspection Report# : 2002003(pdf)
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| Significance: Dec 29, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW 10 CFR 55.59(c)(5) REQUIREMENTS FOR RETAINING LICENSED OPERATOR REQUALIFICATION PROGRAM RECORDS The inspectors identified a Non-Cited Violation wherein the facility licensee had failed to follow the Code of Federal Regulations (CFR) Title 10, Part 55.59(c)(5), Records, requirements by failing to systematically retain all of the original or authenticated copies of the original evaluation documents during the year 2000 annual NRC examination.
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| The finding was of very low safety significance because although the records were not the original or authenticated copies of the original, records did exist in computerized clerically transcribed documents. The computer records had not been signed, and there was no indication that they had been verified correct by the original authors. The unauthenticated documents did provide information that licensed operators, for the most part, had participated and were evaluated during the year 2000 NRC annual requalification examination. However, the inspectors determined that the finding was more than minor. Specifically, the inspectors identified at least one instance in which the transcribed information appeared to be incorrect or missing. The records failure had credible impact on safety, in that, it negatively impacted on the intent of the licensed operator requalification examination process which, in part, is to maintain a high level of confidence that licensed operators continue to possess the requisite knowledge and abilities needed to safely perform licensed duties. In addition, inadequate records keeping adversely affects the NRC's ability to regulate.
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| Inspection Report# : 2001015(pdf)
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| 3Q/2002 Inspection Findings - Clinton Page 5 of 11 Significance: Jan 26, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct longstanding Reactor Core Isolation Cooling (RCIC) System valve degradation Corrective actions for a longstanding deficiency with the Reactor Core Isolation Cooling (RCIC) system steam bypass valve were not effective in stopping the leakage past the valve. This finding was determined to be a Non-Cited Violation of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action." This finding was determined to have very low risk significance because the degraded condition of the valve did not affect the operability of the RCIC system.
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| Inspection Report# : 2001002(pdf)
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| Significance: Jan 26, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to follow Condition Report process for Shutdown Service Water (SX) pipe wall thinning Corrective actions were not implemented to replace a portion of the shutdown service water (SX) system piping after pipe wall thinning was identified. The failure to take the specified corrective actions by the committed due date or to properly reevaluate the degraded condition was determined to be a Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, "Procedures." This finding was determined to have very low safety significance because the SX system remained operable and capable of performing its' safety function.
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| Inspection Report# : 2001002(pdf)
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| Significance: Sep 30, 2000 Identified By: NRC Item Type: NCV NonCited Violation Procedural requirements were not followed when unexpected equipment response was encountered.
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| Maintenance personnel failed to appropriately follow procedure instructions during testing of the Division III emergency diesel generator room fire detection system. These actions led to the emergency diesel generator being rendered inoperable. The procedure violation was treated as a Non-Cited Violation. This issue was of very low safety significance since the other divisional emergency diesel generators and all emergency core cooling systems were operable at the time of discovery.
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| Inspection Report# : 2000015(pdf)
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| Significance: Jun 16, 2000 Identified By: NRC Item Type: NCV NonCited Violation The licensee failed to ensure that appropriate post-modification testing was specified and accomplished for the Division I and Division III EDG output breaker circuitry modifications The licensee failed to ensure that the appropriate post-modification testing (PMT) was specified in the Division I and Division III emergency diesel generator (EDG) output breaker circuitry modification packages and that the post-modification tests were correctly accomplished. This was required to demonstrate through component and functional testing that the modified (rewired) portions of the Division I and Division III EDG output breaker circuitry were adequately installed to accomplish the intent of the plant design changes.
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| Inspection Report# : 2000012(pdf)
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| Barrier Integrity
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| 3Q/2002 Inspection Findings - Clinton Page 6 of 11 Significance: Apr 09, 2002 Identified By: Self Disclosing Item Type: NCV NonCited Violation TECHNICAL SPECIFICATION 5.4.1 WAS IDENTIFIED FOR WORKERS FAILING TO FOLLOW A PROCEDURE WHICH CONTRIBUTED TO THE INADVERTENT LIFTING OF A DOUBLE BLADE GUIDE DURING FUEL MOVEMENT OPERATIONS ON APRIL 9.
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| A Non-Cited Violation of Technical Specifications (TS) 5.4.1 was identified for workers failing to follow a procedure which contributed to the inadvertent lifting of a double blade guide during fuel movement operations on April 9, 2002.
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| This self-revealing finding was more than minor because if left uncorrected, inadvertent movement of components from the reactor core could lead to a more significant safety concern. Using the fuel barrier column on the SDP Appendix A phase 1 worksheet, the inspectors assessed the finding as a very low safety significance issue.
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| Inspection Report# : 2002006(pdf)
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| Significance: Nov 16, 2000 Identified By: NRC Item Type: NCV NonCited Violation Secondary containment was inoperable for 6 minutes during fuel movements when interlock doors were opened.
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| Secondary containment was inoperable for 6 minutes during fuel movements when secondary containment interlock doors were inadvertently opened to move scaffolding. The inoperability was discovered when operators in the control room received an alarm indicating a loss of secondary containment vacuum. One Non-Cited Violation was identified for violating Technical Specification 3.6.4.1 which requires secondary containment operability during fuel moves. This finding was evaluated using the shutdown significance determination process contained in Appendix G of IMC 0609 and was determined to have very low risk significance because it did not meet the criteria for findings requiring a phase 2 significance evaluation.
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| Inspection Report# : 2000017(pdf)
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| Significance: Nov 14, 2000 Identified By: NRC Item Type: NCV NonCited Violation Failure to perform radiographic examinations of Class 2 welds.
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| The inspectors identified a Non-Cited Violation for the failure to perform radiographic examinations of Class 2 welds in compliance with applicable American Society of Mechanical Engineers (ASME) Code requirements. During installation testing of the 1999 Feedwater Keep Fill FW-39 modification, five radiographic examinations had recorded geometric unsharpness values which exceeded Section III and Section V ASME Code limits. Radiographic geometric unsharpness values are used to ensure that the film is of adequate quality to see defects. In addition, inspectors identified that three examinations did not meet Section V Code requirements for documentation of radiographic technique variables which can affect the image quality of the film. The safety significance of this issue was considered very low at this time, based on the absence of adverse consequences, the presence of other image quality indicators, and because the issue did not involve the system isolation valves. The failure to comply with ASME Code radiographic examination requirements could result in the failure to detect flaws within reactor coolant boundary piping, and was considered a Non-Cited Violation of 10 CFR Part 50.55a, "Codes and Standards".
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| Inspection Report# : 2000019(pdf)
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| Emergency Preparedness Significance: Aug 21, 2001 Identified By: NRC
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| 3Q/2002 Inspection Findings - Clinton Page 7 of 11 Item Type: NCV NonCited Violation Violation of 10 CFR 50.54(q) re. SCBA qualifications A Non-Cited Violation of 10 CFR 50.54(q) was identified by the NRC associated with the failure to maintain personnel qualifications for self contained breathing apparatus in accordance with the Clinton Power Station Emergency Plan.
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| The finding was of very low safety significance because the licensee maintained an adequate number of qualified personnel to maintain minimum coverage of the required positions identified in the Emergency Plan.
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| Inspection Report# : 2001010(pdf)
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| Significance: Jun 08, 2001 Identified By: NRC Item Type: VIO Violation Supplemental Inspection -- Failure to correct self-identified defficiencies disclosed through control room communications drills This supplemental inspection was performed by the NRC to assess the licensee's evaluation associated with inaccuracies in the reporting of the Drill and Exercise Performance (DEP) performance indicator and with the performance deficiencies that resulted in a White DEP performance indicator (fourth quarter 1999 through the fourth quarter 2000). During the inspection, performed in accordance with NRC Inspection Procedure 95001, the inspector concluded that the licensee performed an adequate evaluation to determine the causes of both issues. In the case of the performance indicator errors, the licensee performed a root cause evaluation which identified a personnel error that was compounded by the lack of self-checking and verification. In addition, the licensee identified contributing causes that included the failure to provide adequate training to the emergency preparedness staff and the failure to provide adequate procedural guidance to the performance indicator data stewards and verifiers, which also applied to performance indicators in other cornerstones. The inspector concluded that the scope of corrective actions planned and implemented by the licensee appeared to address the identified causes. However, the inspector observed an additional discrepancy in the recently completed performance indicator evaluation related to drill and exercise participation. In addition, the licensee identified an error in its evaluation of one of the other emergency preparedness performance indicators that was not detected during its evaluation. These observations demonstrated weaknesses in the licensee's corrective actions and extent of condition review. The errors in the licensee's reporting of the DEP performance indicator was significant, in that the error resulted in a change of color, (i.e., Green-to-White). Consequently, a violation of 10 CFR 50.9 of more than minor safety significance was identified. Since the inaccurate reporting occurred during the period that the NRC's Enforcement Policy afforded discretion for the non-willful submittal of inaccurate performance indicator information, the NRC is exercising enforcement discretion and not citing the violation. In the case of the White DEP performance indicator, the inspector concluded that the licensee adequately assessed the deficiencies that led to the performance issues. Based on its review, the licensee attributed the White performance indicator to the high failure rate of control room communicator drills (i.e., job performance measures). The licensee identified two apparent causes for the high failure rate: (1) weaknesses in formal training; and (2) failure to meet emergency preparedness management expectations concerning the identification and correction of drill deficiencies.
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| The inspector reviewed the licensee's corrective actions and determined that they addressed the causes identified. As a result of the licensee's immediate corrective actions, the licensee's performance returned the performance indicator to the Green band. The inspector and the licensee concluded that the high failure rate of the control room communicators resulted, in part, from inadequate corrective actions for self-identified deficiencies. Specifically, the licensee control room communicator drills were a portion of an overall annual evaluation of non-licensed operators, which included non-emergency preparedness functions. Generally, the failure of the communications segment of the evaluation did not result in a total failure of the annual evaluation. Therefore, the licensee's remedial actions were limited and were not effective in correcting the deficiencies and preventing similar failures from occurring, as required by 10 CFR 50.47(b)
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| (14). By letter dated 08/22/01, the NRC concluded that a violation of 10 CFR 50.47(b)(14) had occurred and using the NRC's significance determination process, determined that the finding was white.
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| Inspection Report# : 2002006(pdf)
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| Inspection Report# : 2001009(pdf)
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| Significance: Feb 23, 2001 Identified By: Licensee
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| 3Q/2002 Inspection Findings - Clinton Page 8 of 11 Item Type: NCV NonCited Violation Failure to follow emergency plan for on-shift staffing For an approximate 2-month time period, the licensee failed to meet one of the minimum on-shift emergency response organization (ERO) staffing requirements contained in Table 2-1 of the licensee's emergency plan.
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| Inspection Report# : 2001003(pdf)
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| Significance: N/A Apr 28, 2000 Identified By: NRC Item Type: FIN Finding Emergency Preparedness Performance Indicator Verification Alert and Notification System, Drill & Exercise Participation, and Drill & Exercise performance indicators: The inspectors verified that the licensee had acceptably gathered information and reported these three performance indicators, which were in the green band, with the following minor exception. The inspectors identified a discrepancy with the licensee's initial assessment of the Drill and Exercise Performance (DEP) indicator related to the number of performance opportunities associated with a General Emergency declaration during a drill or an exercise. The licensee initially assumed that only three performance opportunities would exist rather than four as provided in NEI 99-02, but later recognized that they had misinterpreted the guidance. This did not affect the DEP performance indicator which was in the green band.
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| Inspection Report# : 2000009(pdf)
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| Occupational Radiation Safety Significance: Oct 08, 2001 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an Adequate Survey to identify and to post a High Radiation Area A finding and associated Non-Cited Violation was identified concerning the failure to perform an adequate radiological survey, as required by 10 CFR 20.1501. Although the licensee identified this issue, the licensee did not thoroughly evaluate the cause(s) of the unanticipated radiological conditions and associated problems in the monitoring of radioactive waste activities, which have resulted in previous, similar incidents. The finding was of very low safety significance because the area radiation levels and the licensee's additional administrative barriers would have limited the potential for an individual inadvertently entering the area and receiving a radiation exposure in excess of regulatory limits.
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| Inspection Report# : 2001015(pdf)
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| Significance: Aug 21, 2001 Identified By: Licensee Item Type: NCV NonCited Violation Failure to maintain administrative control of high radiation area keys as required by Technical Specification 5.7.2 Technical Specification 5.7.2 requires, in part, that doors to high radiation areas in which an individual could receive a deep dose equivalent greater than or equal to 1000 millirem in one hour (at 30 centimeters) shall be provided with locked or continuously guarded doors to prevent unauthorized entry and that the keys to such doors shall be administratively controlled. During October 29 - 31, 2001, the licensee failed to maintain administrative control of a key that controlled five access points to high radiation areas specified above (i.e., lost the key and failed to perform required key inventories to identify its loss), as described in CR No. 2-00-11-016. Since the inspector concluded that sufficient barriers remained to prevent an unauthorized individual from entering the affected areas and receiving an overexposure, the inspector concluded that the incident was of very low safety significance. The licensee also reported the incident to the NRC as an occurrence for the Occupational Exposure Control Effectiveness performance indicator.
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| This is being treated as a Non-Cited Violation.
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| 3Q/2002 Inspection Findings - Clinton Page 9 of 11 Inspection Report# : 2001010(pdf)
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| Significance: SL-IV Jul 27, 2001 Identified By: NRC Item Type: NCV NonCited Violation Misuse of Radioactive Material to Alarm a PCM Radiation protection technician used contaminated material to alarm a portal contamination monitor (PCM), while an individual was performing a contamination survey. Based on the licensee's investigation, the contamination was not placed on the individual, and the individual successfully monitored through an additional PCM. This incident will be reviewed by the NRC for potential enforcement actions. Update: On July 27, 2001, the NRC identified and forwarded to the licensee (by letter) a Non-Cited Violation of the Clinton Station Facility Operating License associated with the deliberate misuse of radioactive material by a junior contract radiation protection technician. On October 20, 2000, the technician misused radioactive material to cause an erroneous alarm on a PCM, as another individual was performing a contamination survey. The licensee identified the incident, entered the incident into its corrective action program, and implemented immediate corrective actions. Since the violation was determined to be willful, the NRC did not assign a significance to the violation using the NRC's Significance Determination Process. In accordance with the NRC Enforcement Policy, the NRC determined that the incident constituted a Severity Level IV violation of the Clinton Power Station Facility Operating License. Further, the NRC determined that the violation met the criteria necessary to disposition the violation as a Non-Cited Violation (Section VI.A.1.d of the NRC Enforcement Policy).
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| Inspection Report# : 2001010(pdf)
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| Inspection Report# : 2000018(pdf)
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| Significance: Oct 25, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Three individuals entered a HRA in violation of Technical Specification 5.7.1 On October 25, 2000, three individuals entered the B residual heat removal heat exchanger room (a posted high radiation area); however, the individuals were not working under a radiation work permit that allowed entry into the high radiation area and did not satisfy either of the three entry conditions of Technical Specification 5.7.1.
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| Inspection Report# : 2000018(pdf)
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| Public Radiation Safety Significance: Dec 08, 2000 Identified By: Licensee Item Type: NCV NonCited Violation Inadvertent Release of Radioactive Material to Unrestricted Area During September 2000, the licensee conducted a survey of tools, equipment, etc. outside of the restricted area (protected area and owner controlled area) and identified low-level contaminated materials that were not under constant surveillance or control. The failure to maintain contstant surveillance and control of the material was a violation of 10 CFR 20.1802 and was characterized as a Non-Cited Violation. Based on the licensee's conservative annual dose assessment (about 1.56 millirem) and the inability to define the origin of each of the items, the inspector concluded that the issue constituted one occurrence/event per the NRC Significance Determination Process (Green).
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| Inspection Report# : 2000021(pdf)
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| Physical Protection
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| 3Q/2002 Inspection Findings - Clinton Page 10 of 11 Miscellaneous Significance: SL-IV Aug 18, 2001 Identified By: NRC Item Type: VIO Violation Falsification of Test Records by Licensee Employee SL IV - On July 2, 2001, by separate letter, NRC issued a Severity Level IV violation of 10 CFR 50.9 for a deliberate falsification by a plant test engineer. Following investigation by the Office of Investigations, NRC determined that, on October 20, 2000, a test engineer forged another employee's signature on two test package cover sheets on by forging another employee's signature without his prior concurrence, in violation of Clinton established plant protocol and procedure.
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| Inspection Report# : 2001010(pdf)
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| Significance: SL-IV Apr 06, 2001 Identified By: NRC Item Type: VIO Violation Violation of 10 CFR 50.7 "Employee Protection" On April 6, 2001, the NRC issued the licensee a Severity Level IV Violation of 10 CFR 50.7. The NRC concluded that the licensee took adverse employment actions against an employee in the licensee's Nuclear Training Department (i.e.,
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| unfavorable 1999 performance review), in part, as a result of the employee's engagement in protected activities. In addition, the NRC learned that several training personnel may be reluctant to discuss department issues within the nuclear training department.
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| Inspection Report# : 2001007(pdf)
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| Inspection Report# : 2001010(pdf)
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| Significance: N/A Jan 26, 2001 Identified By: NRC Item Type: FIN Finding Assessment of Problem Identification and Resolution Performance The team identified that the licensee appropriately entered significant plant issues into the corrective action process by initiating condition reports. Some less significant conditions adverse to quality were evaluated and corrected outside the established process. The trending program was not fully effective as a problem identification tool. Quality Assurance audits and self-assessments reviewed varied in quality. Identified issues were generally evaluated properly, although in several cases the corrective action process did not work effectively to either evaluate or prioritize issues.
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| Current station performance issues including human performance, corrective action program, surveillance testing, and labeling indicate that long term corrective actions previously taken in these areas as restart and post-restart initiatives have not been fully effective to support sustained improvement. Corrective actions were not always fully effective or timely for some individual equipment issues and the effectiveness review process (CARE) did not always identify ineffective corrective actions. The licensee had recently recognized similar deficiencies in corrective action program implementation but had not yet fully developed or completed the corrective actions to improve these areas. The inspectors did not find any reluctance by the station employees to raise safety issues.
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| Inspection Report# : 2001002(pdf)
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| Significance: N/A Dec 31, 2000 Identified By: NRC Item Type: FIN Finding Recent human performance issues have occurred which are associated with operator performance and knowledge based deficiencies that have affected plant operations and responses to transient conditions.
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| Recent human performance issues have occurred which are associated with operator performance and knowledge based deficiencies that have affected plant operations and responses to transient conditions. While the risk of the individual events was very low, the failure of operators to adequately control level parameters indicated a declining trend in this area. These issues could not be easily evaluated by present risk analysis methods because failures to follow procedures
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| 3Q/2002 Inspection Findings - Clinton Page 11 of 11 and maintaining management expectations were not modeled in the Clinton Individual Plant Evaluation. Therefore, the finding is characterized as having no color.
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| Inspection Report# : 2000020(pdf)
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| Significance: N/A Nov 14, 2000 Identified By: NRC Item Type: FIN Finding Three procedures were not written in compliance with the applicable ASME Code.
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| The inspectors reviewed three special process procedures, and identified areas where all three procedures were not written in compliance with the applicable ASME Code. The procedure deficiencies had the potential to affect the ASME Code compliance of weld fabrication and nondestructive examination used on safety-related components and piping. The inspectors noted that each of the ASME Code problems identified contained elements of human performance deficiencies. The human performance aspects, while not always being the root cause of the problem, were significant contributors leading to procedure deficiencies. While the risk of the individual examples was very low, the number of deficiencies indicated a problem with incorporation of applicable ASME Code requirements into special process procedures.
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| Inspection Report# : 2000019(pdf)
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| Significance: N/A Sep 30, 2000 Identified By: NRC Item Type: FIN Finding Recent events affecting plant operations contained elements of human performance deficiencies.
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| NO COLOR. The inspectors noted that several recent events which have affected plant operations and the operability of safety-related components or other components important to safety contained elements of human performance deficiencies. The human performance aspects, while not always being the root cause of the problem, were significant contributors leading to the events. While the risk of the individual events was very low, the number of maintenance-related incidents indicated a problem exists with the control, review, and performance of maintenance activities.
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| Inspection Report# : 2000015(pdf)
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| Significance: N/A May 20, 2000 Identified By: NRC Item Type: FIN Finding Inaccurate historical data for the Safety System Functional Failure Indicator No Color. The licensee identified a failure to submit accurate information to the NRC. The inaccurate information involved the historical data submittal for the Safety System Functional Failure Performance Indicator. The error resulted in a response band color change from Green to White for the first quarter 1999 Performance Indicator. The NRC exercised Enforcement Discretion pursuant to Section VII.B.6 of the Enforcement Policy and did not cite the violation.
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| Inspection Report# : 2000008(pdf)
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| Last modified : December 02, 2002
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| 4Q/2002 Inspection Findings - Clinton Page 1 of 4 Clinton Initiating Events Significance: May 13, 2002 Identified By: Self Disclosing Item Type: FIN Finding ON MAY 13, 2002 WITH THE REACTOR AT APPROXIMATELY 88 PERCENT RATED THERMAL POWER, THE REACTOR AUTOMATICALLY SHUTDOWN DUE TO A HIGH REACTOR VESSEL WATER LEVEL SIGNAL.
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| A performance deficiency, associated with this automatic reactor shut down on May 13, 2002, was identified as a failure to establish preventative maintenance or inspections on the "B" turbine driven reactor feed pump (TDRFP) for similar conditions found on the "A" TDRFP (noted in December 2000) before a component failure which led to the automatic reactor shut down. This issue was more than minor because if left uncorrected (i.e. appropriate preventive maintenance not being identified and conducted), it could lead to a more significant safety concern and could cause the increased frequency of an initiating event. Consequently, the inspectors evaluated the significance of the issue using the SDP Appendix A phase 1 worksheet. Since the finding contributed only to the likelihood of a reactor trip and did not affect mitigating system availability, the inspectors determined that the finding was of very low safety significance.
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| Inspection Report# : 2002006(pdf)
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| Significance: Mar 31, 2002 Identified By: NRC Item Type: NCV NonCited Violation NON-CITED VIOLATION OF T.S. 5.4.1. FOR INADEQUATE OPERATING PROCEDURE, RESULTING IN ERAT-SVC BREAKER TRIP.
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| A Non-Cited Violation of Technical Specification 5.4.1. was identified for an inadequate operating procedure which contributed to an inadvertent emergency reserve auxiliary transformer static-VAR [Volts-Ampere-reactive]-compensator circuit breaker trip. The result of this circuit breaker trip rendered one of the two qualified offsite power sources the transformer inoperable. The finding was of very low safety significance because it could increase the likelihood of an initiating event (reactor trip or a partial loss of offsite power) but did not increase the likelihood that any mitigation equipment would be unavailable.
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| Inspection Report# : 2002005(pdf)
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| Significance: Feb 17, 2002 Identified By: NRC Item Type: FIN Finding A TEMPORARY MODIFICATION ON THE "A" RR FCV CONTROL CIRCUITRY.
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| A finding of very low safety significance was identified associated with a temporary modification installed on the control circuitry for the "A" flow control valve (FCV) of the reactor recirculation system. Problems with the design instructions and other technician knowledge-based deficiencies resulted in an unplanned opening of the "A" flow control valve. Following this unplanned flow control valve movement, operators removed the temporary modification and manually shut down the reactor. The finding was of very low safety significance because unplanned flow control valve movement contributed only to the likelihood of a reactor trip and did not affect mitigating system availability.
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| Inspection Report# : 2001016(pdf)
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| Mitigating Systems Significance: Sep 30, 2002 Identified By: NRC Item Type: NCV NonCited Violation VIOLATION OF PROCEDURAL REQUIREMENTS CAUSED BY HUMAN PERFORMANCE IN THAT THE LICENSEE FAILED TO CONTROL AND DOCUMENT WORK ON A RISK-SIGNIFICANT, SAFETY-RELATED SYSTEM.
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| The inspectors identified a finding of very low safety significance while observing maintenance on the Division I Emergency Diesel Generator (EDG). Specifically, the inspectors identified that one of the insulated bearing bracket bolts on the generator was not properly tightened. The performance issue associated with this finding involved workers performing work steps not specified in the work procedure. Compounding the
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| 4Q/2002 Inspection Findings - Clinton Page 2 of 4 issue was that once these additional work steps were performed, they were not documented in the work procedure. The finding was more than minor because, if left uncorrected, the EDG could have become inoperable which could impact the Mitigating Systems cornerstone. The finding was of very low safety significance because the condition was found and corrected before the EDG was made operable. This finding was a violation of Technical Specification 5.4.1; however, because the licensee placed the violation into its corrective action program, this was determined to be a NCV.
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| Inspection Report# : 2002008(pdf)
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| Significance: May 29, 2002 Identified By: Self Disclosing Item Type: NCV NonCited Violation TECHNICAL SPECIFICATION 5.4.1 VIOLATION WAS IDENTIFIED FOR AN INADEQUATE PROCEDURE USED DURING THE PERFORMANCE OF A DIVISION III EDG TEST.
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| A Non-Cited Violation of Technical Specification 5.4.1 was identified for an inadequate procedure used during the performance of a Division III (Div-III) emergency diesel generator (EDG) test. Errors in the procedure led to the loss of the Div-III safety-related 4160 Volt electrical bus and unplanned unavailability of the high pressure core spray (HPCS) system. The finding was greater than minor because if left uncorrected, the issue has a credible impact on safety. Further, the issue did have an impact on mitigation system operability as the loss of the Div-III electrical bus rendered the HPCS system inoperable. Using Manual Chapter 0609, "Significance Determination Process," (SDP), Appendix A, phase 1 worksheet, the finding screened out as a very low safety significance issue because the event did not result in the actual loss of safety function for the HPCS system.
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| Inspection Report# : 2002006(pdf)
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| Significance: Feb 17, 2002 Identified By: NRC Item Type: NCV NonCited Violation NON-CITED VIOLATION T.S. 5.4.1 FOR AN INADEQUATE SURVEILLANCE PROCEDURE.
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| Procedural inadequacies were determined to be a Non-Cited Violation of Technical Specification 5.4.1. These inadequacies led to the "A" residual heat removal system pump being declared operable without performing the appropriate pump supply breaker functionality checks for the conditions. The finding was of very low safety significance because the licensee subsequently tested the "A" residual heat removal system pump supply breaker with satisfactory results. Therefore, system operability was not impacted.
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| Inspection Report# : 2001016(pdf)
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| Significance: Feb 15, 2002 Identified By: NRC Item Type: NCV NonCited Violation INDAEQUATE CORRECTIVE ACTIONS FOR REPETITIVE FAILURE OF SLC PUMP MOTOR BREAKER.
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| A Non-Cited Violation (NCV) of 10 CFR 50 Appendix B, Criterion XVI, "Corrective Action," for inadequate corrective action taken to prevent recurrence of a Standby Liquid Control "A" System (SLC) pump motor breaker failure was identified. This finding was determined to be of very low safety significance due to the low initiating event frequency for Anticipated Transient Without Scram, the availability of the "B" SLC pump, and the high likelihood of successful operator recovery actions.
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| Inspection Report# : 2002003(pdf)
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| Barrier Integrity Significance: Sep 30, 2002 Identified By: NRC Item Type: NCV NonCited Violation THE LICENSEE FAILED TO ESTABLISH WRITTEN OPERATIONAL TEST PROCEDURES TO DEMONSTRATE THE FUNCTIONAL CAPABILITY OF THE SX MAKEUP TO THE SPENT FUEL POOL.
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| The inspectors determined that the licensee failed to establish written operational test procedures to verify the functionality of the seismically qualified makeup flow path from the shutdown service water system to the spent fuel pool. The finding was more than minor because, if left uncorrected, silting in the line and pipe wall thinning could result in increased degradation and a more significant safety concern and potentially impacting the Barrier Integrity cornerstone. The finding was of very low safety significance because the as-found conditions, while degraded from original installation, met design requirements. This finding was a violation of 10 CFR 50, Appendix B, Criterion XI "Test Control;"
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| however, because the licensee placed the violation into its corrective action program, this was determined to be a NCV.
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| Inspection Report# : 2002008(pdf)
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| 4Q/2002 Inspection Findings - Clinton Page 3 of 4 Significance: Apr 09, 2002 Identified By: Self Disclosing Item Type: NCV NonCited Violation TECHNICAL SPECIFICATION 5.4.1 WAS IDENTIFIED FOR WORKERS FAILING TO FOLLOW A PROCEDURE WHICH CONTRIBUTED TO THE INADVERTENT LIFTING OF A DOUBLE BLADE GUIDE DURING FUEL MOVEMENT OPERATIONS ON APRIL 9.
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| A Non-Cited Violation of Technical Specifications (TS) 5.4.1 was identified for workers failing to follow a procedure which contributed to the inadvertent lifting of a double blade guide during fuel movement operations on April 9, 2002. This self-revealing finding was more than minor because if left uncorrected, inadvertent movement of components from the reactor core could lead to a more significant safety concern. Using the fuel barrier column on the SDP Appendix A phase 1 worksheet, the inspectors assessed the finding as a very low safety significance issue.
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| Inspection Report# : 2002006(pdf)
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| Emergency Preparedness Occupational Radiation Safety Significance: Dec 12, 2002 Identified By: Self Disclosing Item Type: NCV NonCited Violation FAILURE TO CONDUCT ADEQUATE SURVEY OF AIRBORNE RADIOACTIVE MATERIALS A finding of very low safety significance was identified through a self-revealing event, when a maintenance mechanic received an unexpected uptake of radioactive material during a valve maintenance procedure resulting in a 115 millirem committed effective dose equivalent (CEDE) dose. This self-revealing finding was caused by inadequate implementation of radiation protection procedures and improper work oversight by the radiation protection staff. The finding is more than minor because it affects the occupational radiation safety cornerstone objective for exposure/contamination control and monitoring. Although an unexpected intake occurred, the radiological conditions associated with the work activity were not of a magnitude sufficient to produce a substantial potential for an exposure in excess of regulatory limits. Therefore, the finding was of very low safety significance (i.e., not an as-low-as-reasonably-achievable finding, not an overexposure or substantial potential for an overexposure, and did not compromise the ability to assess dose). A Non-Cited Violation of 10 CFR 20.1501(a)(1)(ii) was identified for failure to conduct surveys as necessary to assess the radiological conditions and to control exposure to airborne radioactive material.
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| Inspection Report# : 2002009(pdf)
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| Public Radiation Safety Physical Protection Miscellaneous Significance: N/A Feb 15, 2002 Identified By: NRC Item Type: FIN Finding WHILE IMPROVEMENTS WERE NOTED IN THE CORRECTIVE ACTION PROGRAM, ONE EXAMPLE OF INADEQUATE CORRECTIVE ACTION WAS IDENTIFIED Improvements were noted in most areas of the corrective action program that were reviewed. While one example of inadequate corrective action was identified, the licensee generally identified, evaluated, prioritized and implemented corrective actions for identified issues in an effective manner. Improvements in these areas were primarily due to active involvement in the program by program coordinators and the management review committee. The trending program and the interface between the corrective action and maintenance work order programs
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| 4Q/2002 Inspection Findings - Clinton Page 4 of 4 were two areas that could be further improved. Continued involvement by the program coordinators and the management review committee is critical to further program improvement.
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| Inspection Report# : 2002003(pdf)
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| Last modified : March 25, 2003
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| 1Q/2003 Inspection Findings - Clinton Page 1 of 4 Clinton 1Q/2003 Plant Inspection Findings Initiating Events Significance: May 13, 2002 Identified By: Self Disclosing Item Type: FIN Finding ON MAY 13, 2002 WITH THE REACTOR AT APPROXIMATELY 88 PERCENT RATED THERMAL POWER, THE REACTOR AUTOMATICALLY SHUTDOWN DUE TO A HIGH REACTOR VESSEL WATER LEVEL SIGNAL.
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| A performance deficiency, associated with this automatic reactor shut down on May 13, 2002, was identified as a failure to establish preventative maintenance or inspections on the "B" turbine driven reactor feed pump (TDRFP) for similar conditions found on the "A" TDRFP (noted in December 2000) before a component failure which led to the automatic reactor shut down. This issue was more than minor because if left uncorrected (i.e. appropriate preventive maintenance not being identified and conducted), it could lead to a more significant safety concern and could cause the increased frequency of an initiating event. Consequently, the inspectors evaluated the significance of the issue using the SDP Appendix A phase 1 worksheet. Since the finding contributed only to the likelihood of a reactor trip and did not affect mitigating system availability, the inspectors determined that the finding was of very low safety significance.
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| Inspection Report# : 2002006(pdf)
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| Mitigating Systems Significance: Sep 30, 2002 Identified By: NRC Item Type: NCV NonCited Violation VIOLATION OF PROCEDURAL REQUIREMENTS CAUSED BY HUMAN PERFORMANCE IN THAT THE LICENSEE FAILED TO CONTROL AND DOCUMENT WORK ON A RISK-SIGNIFICANT, SAFETY-RELATED SYSTEM.
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| The inspectors identified a finding of very low safety significance while observing maintenance on the Division I Emergency Diesel Generator (EDG). Specifically, the inspectors identified that one of the insulated bearing bracket bolts on the generator was not properly tightened. The performance issue associated with this finding involved workers performing work steps not specified in the work procedure. Compounding the issue was that once these additional work steps were performed, they were not documented in the work procedure. The finding was more than minor because, if left uncorrected, the EDG could have become inoperable which could impact the Mitigating Systems cornerstone. The finding was of very low safety significance because the condition was found and corrected before the EDG was made operable. This finding was a violation of Technical Specification 5.4.1; however, because the licensee placed the violation into its corrective action program, this was determined to be a NCV.
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| Inspection Report# : 2002008(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\CLIN\clin_pim.html 07/22/2003
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| 1Q/2003 Inspection Findings - Clinton Page 2 of 4 Significance: May 29, 2002 Identified By: Self Disclosing Item Type: NCV NonCited Violation TECHNICAL SPECIFICATION 5.4.1 VIOLATION WAS IDENTIFIED FOR AN INADEQUATE PROCEDURE USED DURING THE PERFORMANCE OF A DIVISION III EDG TEST.
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| A Non-Cited Violation of Technical Specification 5.4.1 was identified for an inadequate procedure used during the performance of a Division III (Div-III) emergency diesel generator (EDG) test. Errors in the procedure led to the loss of the Div-III safety-related 4160 Volt electrical bus and unplanned unavailability of the high pressure core spray (HPCS) system. The finding was greater than minor because if left uncorrected, the issue has a credible impact on safety.
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| Further, the issue did have an impact on mitigation system operability as the loss of the Div-III electrical bus rendered the HPCS system inoperable. Using Manual Chapter 0609, "Significance Determination Process," (SDP), Appendix A, phase 1 worksheet, the finding screened out as a very low safety significance issue because the event did not result in the actual loss of safety function for the HPCS system.
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| Inspection Report# : 2002006(pdf)
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| Barrier Integrity Significance: Feb 20, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation FAILURE TO ISOLATE AN INOPERABLE PRIMARY CONTAINMENT ISOLATION VALVE WITHIN THE ALLOWED ACTION TIME A finding of very low safety significance was identified through a self-revealing event when operators failed to close a motor operated valve prior to de-energizing it when taking the valve out of service. The open valve resulted in an inoperable containment isolation pathway. The primary cause of this finding was related to the cross-cutting area of human performance. This finding is more than minor because it involved the attribute of configuration control under the Barrier Integrity Cornerstone. The finding is of very low safety significance because actual containment integrity was not breached. The failure to isolate an inoperable containment penetration was identified as a Non-cited Violation of Technical Specification 3.6.1.3.
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| Inspection Report# : 2003003(pdf)
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| Significance: Sep 30, 2002 Identified By: NRC Item Type: NCV NonCited Violation THE LICENSEE FAILED TO ESTABLISH WRITTEN OPERATIONAL TEST PROCEDURES TO DEMONSTRATE THE FUNCTIONAL CAPABILITY OF THE SX MAKEUP TO THE SPENT FUEL POOL.
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| The inspectors determined that the licensee failed to establish written operational test procedures to verify the functionality of the seismically qualified makeup flow path from the shutdown service water system to the spent fuel pool. The finding was more than minor because, if left uncorrected, silting in the line and pipe wall thinning could result in increased degradation and a more significant safety concern and potentially impacting the Barrier Integrity cornerstone. The finding was of very low safety significance because the as-found conditions, while degraded from original installation, met design requirements. This finding was a violation of 10 CFR 50, Appendix B, Criterion XI "Test Control;" however, because the licensee placed the violation into its corrective action program, this was determined to be a NCV.
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| file://C:\RROP\NRR\OVERSIGHT\ASSESS\CLIN\clin_pim.html 07/22/2003
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| 1Q/2003 Inspection Findings - Clinton Page 3 of 4 Inspection Report# : 2002008(pdf)
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| Significance: Apr 09, 2002 Identified By: Self Disclosing Item Type: NCV NonCited Violation TECHNICAL SPECIFICATION 5.4.1 WAS IDENTIFIED FOR WORKERS FAILING TO FOLLOW A PROCEDURE WHICH CONTRIBUTED TO THE INADVERTENT LIFTING OF A DOUBLE BLADE GUIDE DURING FUEL MOVEMENT OPERATIONS ON APRIL 9.
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| A Non-Cited Violation of Technical Specifications (TS) 5.4.1 was identified for workers failing to follow a procedure which contributed to the inadvertent lifting of a double blade guide during fuel movement operations on April 9, 2002.
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| This self-revealing finding was more than minor because if left uncorrected, inadvertent movement of components from the reactor core could lead to a more significant safety concern. Using the fuel barrier column on the SDP Appendix A phase 1 worksheet, the inspectors assessed the finding as a very low safety significance issue.
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| Inspection Report# : 2002006(pdf)
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| Emergency Preparedness Occupational Radiation Safety Significance: Dec 12, 2002 Identified By: Self Disclosing Item Type: NCV NonCited Violation FAILURE TO CONDUCT ADEQUATE SURVEY OF AIRBORNE RADIOACTIVE MATERIALS A finding of very low safety significance was identified through a self-revealing event, when a maintenance mechanic received an unexpected uptake of radioactive material during a valve maintenance procedure resulting in a 115 millirem committed effective dose equivalent (CEDE) dose. This self-revealing finding was caused by inadequate implementation of radiation protection procedures and improper work oversight by the radiation protection staff. The finding is more than minor because it affects the occupational radiation safety cornerstone objective for exposure/contamination control and monitoring. Although an unexpected intake occurred, the radiological conditions associated with the work activity were not of a magnitude sufficient to produce a substantial potential for an exposure in excess of regulatory limits. Therefore, the finding was of very low safety significance (i.e., not an as-low-as-reasonably-achievable finding, not an overexposure or substantial potential for an overexposure, and did not compromise the ability to assess dose). A Non-Cited Violation of 10 CFR 20.1501(a)(1)(ii) was identified for failure to conduct surveys as necessary to assess the radiological conditions and to control exposure to airborne radioactive material.
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| Inspection Report# : 2002009(pdf)
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| Public Radiation Safety file://C:\RROP\NRR\OVERSIGHT\ASSESS\CLIN\clin_pim.html 07/22/2003
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| 1Q/2003 Inspection Findings - Clinton Page 4 of 4 Physical Protection Miscellaneous Last modified : May 30, 2003 file://C:\RROP\NRR\OVERSIGHT\ASSESS\CLIN\clin_pim.html 07/22/2003
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| 2Q/2003 Inspection Findings - Clinton Page 1 of 3 Clinton 2Q/2003 Plant Inspection Findings Initiating Events Mitigating Systems Significance: May 23, 2003 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW OPERABILITY EVALUATION PROCEDURE FOR A THROUGH-WALK LEAK IN ASME CLASS III PIPING.
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| A finding of very low safety significance was identified by the inspectors for failure to follow procedures as required by technical specification. This failure to following procedure resulted in an inadequate operability evaluation being performed by the licensee. This issue also resulted in the licensee failing to declare the affected system inoperable as required by NRC regulatory guidance documents and licensee procedures. This issue was more than minor because an inadequate operability evaluation could affect the mitigating system cornerstone objective as it relates to the availability of the Division I service water system and emergency diesel generator. This issue was of very low safety significance because this qualification deficiency did not result in loss of function per GL 91-18. This issue was a non-cited violation of Technical Specification 5.4 which required the implementation of written procedures in NRC Regulatory Guide 1.33, Appendix A.
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| Inspection Report# : 2003004(pdf)
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| Significance: SL-III Jan 24, 2003 Identified By: NRC Item Type: VIO Violation FAILURE TO PROVIDE COMPLETE AND ACCURATE INFORMATION TO THE NRC WHICH IMPACTED A LICENSING DECISION.
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| Clinton Station management personnel informed NRC Region III by letter dated September 24, 2002, that two operators who had been examined for their operator licenses in August 2002 had long standing medical conditions that warranted reporting to the NRC for review. Both operators were issued a license by the NRC on August 30, 2002. The licensee originally sent NRC Form 396s for both operators to Region III on June 26, 2002, without including their medical records and did not recommend any license restrictions. One operator had a history of myocardial infarction and the other had a history of coronary heart disease. The medical conditions described above are considered potentially disqualifying in accordance with American Nuclear Standards Institute/American Nuclear Society (ANSI/ANS) 3.4, 1983, and should have been reported to the NRC with a request for issuance of a license with a "no solo" restriction. When the licensee informed the NRC on September 24, 2002, of the medical conditions of the two operators there still was no request for an amended "no solo" license for either operator. Because the issue affected the NRC's ability to perform its regulatory function, it was evaluated with the traditional enforcement process. The finding was determined to be of low safety significance because the operators had not acted in a solo capacity prior to having their license's amended. However, the regulatory significance was important because the incorrect information was provided under sworn statement to the NRC and impacted a licensing decision for the two individuals. The issue was preliminarily determined to be an apparent violation of 10 CFR 50.9.
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| file://C:\RROP\NRR\OVERSIGHT\ASSESS\CLIN\clin_pim.html 10/08/2003
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| 2Q/2003 Inspection Findings - Clinton Page 2 of 3 Inspection Report# : 2003002(pdf)
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| Significance: Sep 30, 2002 Identified By: NRC Item Type: NCV NonCited Violation VIOLATION OF PROCEDURAL REQUIREMENTS CAUSED BY HUMAN PERFORMANCE IN THAT THE LICENSEE FAILED TO CONTROL AND DOCUMENT WORK ON A RISK-SIGNIFICANT, SAFETY-RELATED SYSTEM.
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| The inspectors identified a finding of very low safety significance while observing maintenance on the Division I Emergency Diesel Generator (EDG). Specifically, the inspectors identified that one of the insulated bearing bracket bolts on the generator was not properly tightened. The performance issue associated with this finding involved workers performing work steps not specified in the work procedure. Compounding the issue was that once these additional work steps were performed, they were not documented in the work procedure. The finding was more than minor because, if left uncorrected, the EDG could have become inoperable which could impact the Mitigating Systems cornerstone. The finding was of very low safety significance because the condition was found and corrected before the EDG was made operable. This finding was a violation of Technical Specification 5.4.1; however, because the licensee placed the violation into its corrective action program, this was determined to be a NCV.
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| Inspection Report# : 2002008(pdf)
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| Barrier Integrity Significance: Feb 20, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation FAILURE TO ISOLATE AN INOPERABLE PRIMARY CONTAINMENT ISOLATION VALVE WITHIN THE ALLOWED ACTION TIME A finding of very low safety significance was identified through a self-revealing event when operators failed to close a motor operated valve prior to de-energizing it when taking the valve out of service. The open valve resulted in an inoperable containment isolation pathway. The primary cause of this finding was related to the cross-cutting area of human performance. This finding is more than minor because it involved the attribute of configuration control under the Barrier Integrity Cornerstone. The finding is of very low safety significance because actual containment integrity was not breached. The failure to isolate an inoperable containment penetration was identified as a Non-cited Violation of Technical Specification 3.6.1.3.
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| Inspection Report# : 2003003(pdf)
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| Significance: Sep 30, 2002 Identified By: NRC Item Type: NCV NonCited Violation THE LICENSEE FAILED TO ESTABLISH WRITTEN OPERATIONAL TEST PROCEDURES TO DEMONSTRATE THE FUNCTIONAL CAPABILITY OF THE SX MAKEUP TO THE SPENT FUEL POOL.
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| The inspectors determined that the licensee failed to establish written operational test procedures to verify the functionality of the seismically qualified makeup flow path from the shutdown service water system to the spent fuel pool. The finding was more than minor because, if left uncorrected, silting in the line and pipe wall thinning could result in increased degradation and a more significant safety concern and potentially impacting the Barrier Integrity file://C:\RROP\NRR\OVERSIGHT\ASSESS\CLIN\clin_pim.html 10/08/2003
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| 2Q/2003 Inspection Findings - Clinton Page 3 of 3 cornerstone. The finding was of very low safety significance because the as-found conditions, while degraded from original installation, met design requirements. This finding was a violation of 10 CFR 50, Appendix B, Criterion XI "Test Control;" however, because the licensee placed the violation into its corrective action program, this was determined to be a NCV.
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| Inspection Report# : 2002008(pdf)
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| Emergency Preparedness Occupational Radiation Safety Significance: Dec 12, 2002 Identified By: Self Disclosing Item Type: NCV NonCited Violation FAILURE TO CONDUCT ADEQUATE SURVEY OF AIRBORNE RADIOACTIVE MATERIALS A finding of very low safety significance was identified through a self-revealing event, when a maintenance mechanic received an unexpected uptake of radioactive material during a valve maintenance procedure resulting in a 115 millirem committed effective dose equivalent (CEDE) dose. This self-revealing finding was caused by inadequate implementation of radiation protection procedures and improper work oversight by the radiation protection staff. The finding is more than minor because it affects the occupational radiation safety cornerstone objective for exposure/contamination control and monitoring. Although an unexpected intake occurred, the radiological conditions associated with the work activity were not of a magnitude sufficient to produce a substantial potential for an exposure in excess of regulatory limits. Therefore, the finding was of very low safety significance (i.e., not an as-low-as-reasonably-achievable finding, not an overexposure or substantial potential for an overexposure, and did not compromise the ability to assess dose). A Non-Cited Violation of 10 CFR 20.1501(a)(1)(ii) was identified for failure to conduct surveys as necessary to assess the radiological conditions and to control exposure to airborne radioactive material.
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| Inspection Report# : 2002009(pdf)
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| Public Radiation Safety Physical Protection Miscellaneous Last modified : September 04, 2003 file://C:\RROP\NRR\OVERSIGHT\ASSESS\CLIN\clin_pim.html 10/08/2003
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| 3Q/2003 Inspection Findings - Clinton Page 1 of 3 Clinton 3Q/2003 Plant Inspection Findings Initiating Events Mitigating Systems Significance: May 23, 2003 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW OPERABILITY EVALUATION PROCEDURE FOR A THROUGH-WALL LEAK IN ASME CLASS III PIPING.
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| A finding of very low safety significance was identified by the inspectors for failure to follow procedures as required by technical specification. This failure to following procedure resulted in an inadequate operability evaluation being performed by the licensee. This issue also resulted in the licensee failing to declare the affected system inoperable as required by NRC regulatory guidance documents and licensee procedures.
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| This issue was more than minor because an inadequate operability evaluation could affect the mitigating system cornerstone objective as it relates to the availability of the Division I service water system and emergency diesel generator. This issue was of very low safety significance because this qualification deficiency did not result in loss of function per GL 91-18. This issue was a non-cited violation of Technical Specification 5.4 which required the implementation of written procedures in NRC Regulatory Guide 1.33, Appendix A.
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| Inspection Report# : 2003004(pdf)
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| Significance: SL-III Jan 24, 2003 Identified By: NRC Item Type: VIO Violation FAILURE TO PROVIDE COMPLETE AND ACCURATE INFORMATION TO THE NRC WHICH IMPACTED A LICENSING DECISION.
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| Clinton Station management personnel informed NRC Region III by letter dated September 24, 2002, that two operators who had been examined for their operator licenses in August 2002 had long standing medical conditions that warranted reporting to the NRC for review. Both operators were issued a license by the NRC on August 30, 2002. The licensee originally sent NRC Form 396s for both operators to Region III on June 26, 2002, without including their medical records and did not recommend any license restrictions. One operator had a history of myocardial infarction and the other had a history of coronary heart disease. The medical conditions described above are considered potentially disqualifying in accordance with American Nuclear Standards Institute/American Nuclear Society (ANSI/ANS) 3.4, 1983, and should have been reported to the NRC with a request for issuance of a license with a "no solo" restriction. When the licensee informed the NRC on September 24, 2002, of the medical conditions of the two operators there still was no request for an amended "no solo" license for either operator.
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| Because the issue affected the NRC's ability to perform its regulatory function, it was evaluated with the traditional enforcement process. The finding was determined to be of low safety significance because the operators had not acted in a solo capacity prior to having their license's amended. However, the regulatory significance was important because file://C:\RROP\NRR\OVERSIGHT\ASSESS\CLIN\clin_pim.html 01/12/2004
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| 3Q/2003 Inspection Findings - Clinton Page 2 of 3 the incorrect information was provided under sworn statement to the NRC and impacted a licensing decision for the two individuals. The issue was preliminarily determined to be an apparent violation of 10 CFR 50.9.
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| Inspection Report# : 2003002(pdf)
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| Barrier Integrity Significance: Feb 20, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation FAILURE TO ISOLATE AN INOPERABLE PRIMARY CONTAINMENT ISOLATION VALVE WITHIN THE ALLOWED ACTION TIME A finding of very low safety significance was identified through a self-revealing event when operators failed to close a motor operated valve prior to de-energizing it when taking the valve out of service. The open valve resulted in an inoperable containment isolation pathway. The primary cause of this finding was related to the cross-cutting area of human performance. This finding is more than minor because it involved the attribute of configuration control under the Barrier Integrity Cornerstone. The finding is of very low safety significance because actual containment integrity was not breached. The failure to isolate an inoperable containment penetration was identified as a Non-cited Violation of Technical Specification 3.6.1.3.
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| Inspection Report# : 2003003(pdf)
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| Emergency Preparedness Occupational Radiation Safety Significance: Dec 12, 2002 Identified By: Self Disclosing Item Type: NCV NonCited Violation FAILURE TO CONDUCT ADEQUATE SURVEY OF AIRBORNE RADIOACTIVE MATERIALS A finding of very low safety significance was identified through a self-revealing event, when a maintenance mechanic received an unexpected uptake of radioactive material during a valve maintenance procedure resulting in a 115 millirem committed effective dose equivalent (CEDE) dose. This self-revealing finding was caused by inadequate implementation of radiation protection procedures and improper work oversight by the radiation protection staff.
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| The finding is more than minor because it affects the occupational radiation safety cornerstone objective for exposure/contamination control and monitoring. Although an unexpected intake occurred, the radiological conditions associated with the work activity were not of a magnitude sufficient to produce a substantial potential for an exposure in excess of regulatory limits. Therefore, the finding was of very low safety significance (i.e., not an as-low-as-reasonably-achievable finding, not an overexposure or substantial potential for an overexposure, and did not compromise the ability to assess dose). A Non-Cited Violation of 10 CFR 20.1501(a)(1)(ii) was identified for failure to conduct surveys as necessary to assess the radiological conditions and to control exposure to airborne radioactive file://C:\RROP\NRR\OVERSIGHT\ASSESS\CLIN\clin_pim.html 01/12/2004
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| 3Q/2003 Inspection Findings - Clinton Page 3 of 3 material.
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| Inspection Report# : 2002009(pdf)
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| Public Radiation Safety Physical Protection Miscellaneous Last modified : December 01, 2003 file://C:\RROP\NRR\OVERSIGHT\ASSESS\CLIN\clin_pim.html 01/12/2004
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| 4Q/2003 Inspection Findings - Clinton Page 1 of 3 Clinton 4Q/2003 Plant Inspection Findings Initiating Events Significance: Dec 31, 2003 Identified By: NRC Item Type: FIN Finding AUTOMATIC SHUTDOWN SIGNAL GENERATED DUE TO PERSONNEL ERROR The inspectors identified a finding of very low safety significance concerning poor operator performance following a reactor scram on December 2, 2003. The primary cause of this finding was related to the cross-cutting area of Human Performance, in that, poor performance by operations personnel resulted in a momentary loss of reactor pressure vessel level control. This loss of level resulted in a second reactor scram signal being generated. No violations of NRC requirements occurred.
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| This finding was more than minor because the finding affected the Reactor Safety/Initiating Event objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The finding did not contribute to the likelihood of a Primary or Secondary system loss of coolant accident initiator, did not contribute to both the likelihood of a reactor trip AND the likelihood that mitigation equipment or functions will not be available, and did not increase the likelihood of a fire or internal/external flood.
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| Therefore, the finding was determined to be of very low safety significance.
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| Inspection Report# : 2003009(pdf)
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| Mitigating Systems Significance: Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PROMPTLY IMPLEMENT CORRECTIVE ACTIONS.
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| The inspectors identified a non-cited violation of 10 CFR 50 Appendix B Criterion XVI involving the licensee's failure to promptly enter an identified condition adverse to quality into their corrective action program. This finding related to the cross-cutting area of Human Performance, in that, engineering personnel were aware of a discrepant condition on the 4160 volt Bus 1C1 Reserve Feed potential transformer cubicle door but did not correct the condition for several days.
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| The inspectors determined that this issue was more than minor because the finding could be reasonably viewed as a precursor to a significant event if left uncorrected because the station personnel could fail to evaluate non-conforming conditions which could render safety related equipment inoperable. This issue was design/seismic qualification deficiency that was determined to not cause a loss of function by the licensee's evaluation. Based on this conclusion, this finding was determined to be of very low safety significance using the Phase 1 worksheets.
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| Inspection Report# : 2003009(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\CLIN\clin_pim.html 04/22/2004
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| 4Q/2003 Inspection Findings - Clinton Page 2 of 3 Significance: May 23, 2003 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW OPERABILITY EVALUATION PROCEDURE FOR A THROUGH-WALL LEAK IN ASME CLASS III PIPING.
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| A finding of very low safety significance was identified by the inspectors for failure to follow procedures as required by technical specification. This failure to following procedure resulted in an inadequate operability evaluation being performed by the licensee. This issue also resulted in the licensee failing to declare the affected system inoperable as required by NRC regulatory guidance documents and licensee procedures.
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| This issue was more than minor because an inadequate operability evaluation could affect the mitigating system cornerstone objective as it relates to the availability of the Division I service water system and emergency diesel generator. This issue was of very low safety significance because this qualification deficiency did not result in loss of function per GL 91-18. This issue was a non-cited violation of Technical Specification 5.4 which required the implementation of written procedures in NRC Regulatory Guide 1.33, Appendix A.
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| Inspection Report# : 2003004(pdf)
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| Significance: SL-III Jan 24, 2003 Identified By: NRC Item Type: VIO Violation FAILURE TO PROVIDE COMPLETE AND ACCURATE INFORMATION TO THE NRC WHICH IMPACTED A LICENSING DECISION.
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| Clinton Station management personnel informed NRC Region III by letter dated September 24, 2002, that two operators who had been examined for their operator licenses in August 2002 had long standing medical conditions that warranted reporting to the NRC for review. Both operators were issued a license by the NRC on August 30, 2002. The licensee originally sent NRC Form 396s for both operators to Region III on June 26, 2002, without including their medical records and did not recommend any license restrictions. One operator had a history of myocardial infarction and the other had a history of coronary heart disease. The medical conditions described above are considered potentially disqualifying in accordance with American Nuclear Standards Institute/American Nuclear Society (ANSI/ANS) 3.4, 1983, and should have been reported to the NRC with a request for issuance of a license with a "no solo" restriction. When the licensee informed the NRC on September 24, 2002, of the medical conditions of the two operators there still was no request for an amended "no solo" license for either operator.
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| Because the issue affected the NRC's ability to perform its regulatory function, it was evaluated with the traditional enforcement process. The finding was determined to be of low safety significance because the operators had not acted in a solo capacity prior to having their license's amended. However, the regulatory significance was important because the incorrect information was provided under sworn statement to the NRC and impacted a licensing decision for the two individuals. The issue was preliminarily determined to be an apparent violation of 10 CFR 50.9.
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| Inspection Report# : 2003002(pdf)
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| Barrier Integrity Significance: Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation file://C:\RROP\NRR\OVERSIGHT\ASSESS\CLIN\clin_pim.html 04/22/2004
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| 4Q/2003 Inspection Findings - Clinton Page 3 of 3 FAILURE TO PERFORM A TS REQUIRED SURVEILLANCE The inspectors identified a finding of very low safety significance (Green) concerning the licensee's failure to verify heatup and cooldown rates in accordance with Technical Specification (TS) following a scram on December 2, 2003.
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| This was determined to be a NCV of TS surveillance requirement 3.4.11.1.
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| This finding was more than minor because if left uncorrected, failure to perform a TS surveillance could become a more safety significant issue. This finding was not suitable for SDP evaluation but has been reviewed by NRC management and was determined to be a finding of very low safety significance. This issue may have been greater than Green if the TS temperature limitations had been exceeded and if subsequent evaluation showed a degradation of the reactor coolant system integrity.
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| Inspection Report# : 2003009(pdf)
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| Significance: Feb 20, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation FAILURE TO ISOLATE AN INOPERABLE PRIMARY CONTAINMENT ISOLATION VALVE WITHIN THE ALLOWED ACTION TIME A finding of very low safety significance was identified through a self-revealing event when operators failed to close a motor operated valve prior to de-energizing it when taking the valve out of service. The open valve resulted in an inoperable containment isolation pathway. The primary cause of this finding was related to the cross-cutting area of human performance. This finding is more than minor because it involved the attribute of configuration control under the Barrier Integrity Cornerstone. The finding is of very low safety significance because actual containment integrity was not breached. The failure to isolate an inoperable containment penetration was identified as a Non-cited Violation of Technical Specification 3.6.1.3.
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| Inspection Report# : 2003003(pdf)
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| Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Miscellaneous Last modified : March 02, 2004 file://C:\RROP\NRR\OVERSIGHT\ASSESS\CLIN\clin_pim.html 04/22/2004
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| 1Q/2004 Inspection Findings - Clinton Page 1 of 3 Clinton 1Q/2004 Plant Inspection Findings Initiating Events Significance: Dec 31, 2003 Identified By: NRC Item Type: FIN Finding AUTOMATIC SHUTDOWN SIGNAL GENERATED DUE TO PERSONNEL ERROR The inspectors identified a finding of very low safety significance concerning poor operator performance following a reactor scram on December 2, 2003. The primary cause of this finding was related to the cross-cutting area of Human Performance, in that, poor performance by operations personnel resulted in a momentary loss of reactor pressure vessel level control. This loss of level resulted in a second reactor scram signal being generated. No violations of NRC requirements occurred.
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| This finding was more than minor because the finding affected the Reactor Safety/Initiating Event objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The finding did not contribute to the likelihood of a Primary or Secondary system loss of coolant accident initiator, did not contribute to both the likelihood of a reactor trip AND the likelihood that mitigation equipment or functions will not be available, and did not increase the likelihood of a fire or internal/external flood. Therefore, the finding was determined to be of very low safety significance.
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| Inspection Report# : 2003009(pdf)
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| Mitigating Systems Significance: Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PROMPTLY IMPLEMENT CORRECTIVE ACTIONS.
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| The inspectors identified a non-cited violation of 10 CFR 50 Appendix B Criterion XVI involving the licensee's failure to promptly enter an identified condition adverse to quality into their corrective action program. This finding related to the cross-cutting area of Human Performance, in that, engineering personnel were aware of a discrepant condition on the 4160 volt Bus 1C1 Reserve Feed potential transformer cubicle door but did not correct the condition for several days.
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| The inspectors determined that this issue was more than minor because the finding could be reasonably viewed as a precursor to a significant event if left uncorrected because the station personnel could fail to evaluate non-conforming conditions which could render safety related equipment inoperable. This issue was design/seismic qualification deficiency that was determined to not cause a loss of function by the licensee's evaluation. Based on this conclusion, this finding was determined to be of very low safety significance using the Phase 1 worksheets.
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| Inspection Report# : 2003009(pdf)
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| Significance: May 23, 2003 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW OPERABILITY EVALUATION PROCEDURE FOR A THROUGH-WALL LEAK IN ASME CLASS III PIPING.
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| A finding of very low safety significance was identified by the inspectors for failure to follow procedures as required by technical specification.
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| This failure to following procedure resulted in an inadequate operability evaluation being performed by the licensee. This issue also resulted in the licensee failing to declare the affected system inoperable as required by NRC regulatory guidance documents and licensee procedures.
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| This issue was more than minor because an inadequate operability evaluation could affect the mitigating system cornerstone objective as it relates to the availability of the Division I service water system and emergency diesel generator. This issue was of very low safety significance because this qualification deficiency did not result in loss of function per GL 91-18. This issue was a non-cited violation of Technical Specification 5.4 which required the implementation of written procedures in NRC Regulatory Guide 1.33, Appendix A.
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| Inspection Report# : 2003004(pdf) 07/14/2004
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| 1Q/2004 Inspection Findings - Clinton Page 2 of 3 Significance: SL-III Jan 24, 2003 Identified By: NRC Item Type: VIO Violation FAILURE TO PROVIDE COMPLETE AND ACCURATE INFORMATION TO THE NRC WHICH IMPACTED A LICENSING DECISION.
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| Clinton Station management personnel informed NRC Region III by letter dated September 24, 2002, that two operators who had been examined for their operator licenses in August 2002 had long standing medical conditions that warranted reporting to the NRC for review. Both operators were issued a license by the NRC on August 30, 2002. The licensee originally sent NRC Form 396s for both operators to Region III on June 26, 2002, without including their medical records and did not recommend any license restrictions. One operator had a history of myocardial infarction and the other had a history of coronary heart disease. The medical conditions described above are considered potentially disqualifying in accordance with American Nuclear Standards Institute/American Nuclear Society (ANSI/ANS) 3.4, 1983, and should have been reported to the NRC with a request for issuance of a license with a "no solo" restriction. When the licensee informed the NRC on September 24, 2002, of the medical conditions of the two operators there still was no request for an amended "no solo" license for either operator.
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| Because the issue affected the NRC's ability to perform its regulatory function, it was evaluated with the traditional enforcement process. The finding was determined to be of low safety significance because the operators had not acted in a solo capacity prior to having their license's amended. However, the regulatory significance was important because the incorrect information was provided under sworn statement to the NRC and impacted a licensing decision for the two individuals. The issue was preliminarily determined to be an apparent violation of 10 CFR 50.9.
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| Inspection Report# : 2003002(pdf)
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| Barrier Integrity Significance: Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PERFORM A TS REQUIRED SURVEILLANCE The inspectors identified a finding of very low safety significance (Green) concerning the licensee's failure to verify heatup and cooldown rates in accordance with Technical Specification (TS) following a scram on December 2, 2003. This was determined to be a NCV of TS surveillance requirement 3.4.11.1.
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| This finding was more than minor because if left uncorrected, failure to perform a TS surveillance could become a more safety significant issue. This finding was not suitable for SDP evaluation but has been reviewed by NRC management and was determined to be a finding of very low safety significance. This issue may have been greater than Green if the TS temperature limitations had been exceeded and if subsequent evaluation showed a degradation of the reactor coolant system integrity.
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| Inspection Report# : 2003009(pdf)
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| Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection 07/14/2004
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| 1Q/2004 Inspection Findings - Clinton Page 3 of 3 Miscellaneous Last modified : May 05, 2004 07/14/2004
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| 2Q/2004 Inspection Findings - Clinton Page 1 of 5 Clinton 2Q/2004 Plant Inspection Findings Initiating Events Significance: Dec 31, 2003 Identified By: NRC Item Type: FIN Finding AUTOMATIC SHUTDOWN SIGNAL GENERATED DUE TO PERSONNEL ERROR The inspectors identified a finding of very low safety significance concerning poor operator performance following a reactor scram on December 2, 2003. The primary cause of this finding was related to the cross-cutting area of Human Performance, in that, poor performance by operations personnel resulted in a momentary loss of reactor pressure vessel level control. This loss of level resulted in a second reactor scram signal being generated. No violations of NRC requirements occurred.
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| This finding was more than minor because the finding affected the Reactor Safety/Initiating Event objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The finding did not contribute to the likelihood of a Primary or Secondary system loss of coolant accident initiator, did not contribute to both the likelihood of a reactor trip AND the likelihood that mitigation equipment or functions will not be available, and did not increase the likelihood of a fire or internal/external flood. Therefore, the finding was determined to be of very low safety significance.
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| Inspection Report# : 2003009(pdf)
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| Mitigating Systems Significance: Jun 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IMPLEMENT A LOCKED VALVE PROCEDURE.
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| A finding of very low safety significance was identified by the inspectors for the licensee's failure to implement a procedure to control locked valves.
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| Failing to have a locked valve procedure, combined with a shift supervisor marking the step which verified the position of the standby liquid control (SLC) tank air-sparging valve as "not applicable," based on the valve being a "locked valve" and no work having been done to the valve, allowed the air sparging valve to remain mispositioned while transitioning to Mode-2 and during Mode-1 operations. Once identified, the licensee placed the valve in the correct position. This issue was related to the Human Performance corsscutting area, in that, the failure to implement a procedure resulted in a mispositioned valve.
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| The finding was more than minor because the open air sparging valve created the potential for air-binding the pumps used to inject boron solution into the reactor, affecting the ability of the SLC system to shut the reactor down from a full power situation in the control rods failed to insert on a scram condition. The finding was of very low safety-significance because the deficiency, once evaluated, did not result in a loss of function per Generic Letter 91-18. The finding was a Non-Cited Violation of Technical Specification 5.4 which required the implementation of written procedures to control the locked valves in the plant.
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| Inspection Report# : 2004005(pdf)
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| Significance: Jun 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation SLC BORON CONCENTRATION OUTSIDE TS LIMITS FOR GREATER THAN ALLOWED OUTAGE TIME.
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| A finding of very low safety significance was identified by the inspectors for the licensee's failure to take timely corrective actions after discovering that the standby liquid control (SLC) tank air-aparging valve was in the wrong position for about 2 months. This resulted in the boron concentration in the tank being outside the Technical Specification allowed limits for greater than the Technical Specification allowed action time. Once identified, the licensee restored the concentration in the tank to within acceptable limits. This finding was related to the Problem Identification and Resolution crosscutting area, in that, the concentration in the tank remained outside limits due to the licensee's failure to identify the impact of evaporation on the solution.
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| The finding was more than minor because the boron concentration being ouside the Technical Specification allowed range affects the cross-cutting attribute of SLC system performance and also affected the SLC system's availability, reliability, and capability of responding to plant events. The finding was of very low safety significance because the as-found concentration, although above technical specification limits, did not impact the safety function of the pumps. The finding was a Non-Cited Violation of 10CFR50, Appendix B, Criterion XVI which requires condtions adverse to quality be promptly identified and corrected.
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| 2Q/2004 Inspection Findings - Clinton Page 2 of 5 Inspection Report# : 2004005(pdf)
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| Significance: Apr 07, 2004 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IDENTIFY THE EXTENT OF CONDITION FOR INCORRECT FUSES IN THE REACTOR PROTECTION SYSTEM.
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| The inspectors identified a finding of very low safety significance concerning the licesnee's failure to determine the extent of condition for improper fuses installed in the reactor protection system (RPS) electronic circuit boards. This finding was determined to be a Non-Cited Violation of 10 CFR 50 Appendix B, Criterion XVI.
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| This finding is more than minor because it affects the design and reliability of the RPS to perform its protective function of protecting the reactor core and containment. The licensee determined that although the fuses were improperly sized, the reactor protection system remained operable and could perfrom it's safety function. Therefore, this finding was determined to be of very low safety significance.
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| Inspection Report# : 2004003(pdf)
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| Significance: Apr 07, 2004 Identified By: NRC Item Type: FIN Finding FAILURE TO EVALUATE THE EXTENT OF CONDITION OF FOREIGN MATERIAL FOUND IN THE DIVISION 1 EMERGENCY DIESEL GENERATOR STARTING AIR SYSTEM.
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| The team identified a finding of very low safety significance when the licensee failed to take appropriate steps to evaluate the extent of condition of foreign material in the starting air system of an emergency diesel generator.
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| The finding is more than minor because it is associated with the Mitigating System (MS) cornerstone attribute of equipment reliability and capability of systems that respond to initiating events to prevent undesirable circumstances. This finding was of very low safety significance because once evaluated, it did not result in a loss of function per Generic Letter 91-18 (Rev 1). No vilations of NRC requirements were identified. The licensee documented this issue in condition report 213491. Additionally the licensee established action items to evaluate the source of the foregin material found in the 1A Diesel Generator air system following the March 2004 failure.
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| Inspection Report# : 2004003(pdf)
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| Significance: Mar 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation INEFFECTIVE CORRECTIVE ACTION PIPE WALL THINNING The inspectors identified a finding of very low safety-significance and an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion XVI.
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| The licensee had replaced shutdown service water (SX) system piping following cavitation induced wall thinning and weld failure leading to a through wall leak in 1999. The corrective actions included periodic non-destructive examination (NDE) monitoring of the pipe-wall for cavitation induced wall-thinning. Following an inquiry by the inspectors about heavy cavitation effects on the piping, the licensee discovered that the NDE monitoring had been performed in the wrong section of the piping. When the correct section was examined, the piping was found below manufacture's minimum allowable wall thickness. The finding affected the cross-cutting area of Human-Performance because the system manager and others had failed to identify that the corrective actions for a previous failed pipe had not been correctly implemented since 1999 and had also subsequently failed to expand the extent of condition to include verifying that all 10 predefined NDE activities established by the 1999 corrective actions were being performed in the correct location immediately downstream of SX system flow orifices.
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| The finding was more than minor because it affects the Reactor Safety/Mitigating System Cornerstone and if left uncorrected, it would become a more significant safety concern. The finding was of very low safety-significance because the SX system remained operable, both for function and for seismic considerations. The finding involved the attributes of availability and reliability of the shutdown service water system, internal flooding, and loss of heat sink as well as human performance and could have affected the mitigating systems objective of ensuring the availiability of systems that respond to initiating events to prevent undersirable consequences. The licensee entered the event into its corrective action system, performed an operability determination allowing continued use of the pipe, and replaced the piping in March 2004.
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| Inspection Report# : 2004002(pdf)
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| Significance: Mar 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation EMERGENCY CORE COOLING SYSTEM WATER HAMMER A finding of very low safety significance, with an associated Non-Cityed Violation, was self-revealed relating to a violation of the requirements of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings. The licensee failed to properly vent the high pressure core spray system before performing an integrated ECCS test resulting in a water-hammer event on the high-pressure core spray system.
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| This finding was more than minor because it affected the Mitigating Systems Cornerstone objective of maintaining mitigating systems operable. The finding was of very low safety-significance because a licensee follow-up system investigation, including a complete system walkdown by engineers, revealed that the high pressure core spray system remained operability. This issue was entered into the licensee corrective action program.
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| 2Q/2004 Inspection Findings - Clinton Page 3 of 5 Inspection Report# : 2004002(pdf)
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| Significance: Mar 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation DESIGN CONTROL OF MOTOR OPERATED VALVE MOUNTING BOLTS A finding of very low safety significance was identified by the inspectors for a violation of the requirements of 10 CFR 50, Appendix B, Criterion III, Design Control. Following the licensee's identification that the operator mounting bolts for several Limitorque SMB-2 actuators did not fit properly, the licensee installed bolts with thread engagement less than the required minimum. This was completed without performing the appropriate level design control review. The minimum thread engagement caused a residual heat removal system Limitorque SMB-2 valve actuator to wobble when operated.
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| This finding affected the cross-cutting are of problem identification and resolution because initially, the licensee did not determine cause or extent of condition of the wobbly actuator.
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| This finding was more than minor because it affected the Mitigating Systems Cornerstone objective of maintaining mitigating systems operable. The finding was of very low safety-significance because an evaluation determined that the valve would have performed its safety function when called upon during a design basis seismic event. The finding was entered into the licensee corrective action program and the licensee verified the correct installation of all SMB-2 actuator mounting bolts.
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| Inspection Report# : 2004002(pdf)
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| Significance: Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PROMPTLY IMPLEMENT CORRECTIVE ACTIONS.
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| The inspectors identified a non-cited violation of 10 CFR 50 Appendix B Criterion XVI involving the licensee's failure to promptly enter an identified condition adverse to quality into their corrective action program. This finding related to the cross-cutting area of Human Performance, in that, engineering personnel were aware of a discrepant condition on the 4160 volt Bus 1C1 Reserve Feed potential transformer cubicle door but did not correct the condition for several days.
| |
| The inspectors determined that this issue was more than minor because the finding could be reasonably viewed as a precursor to a significant event if left uncorrected because the station personnel could fail to evaluate non-conforming conditions which could render safety related equipment inoperable.
| |
| This issue was design/seismic qualification deficiency that was determined to not cause a loss of function by the licensee's evaluation. Based on this conclusion, this finding was determined to be of very low safety significance using the Phase 1 worksheets.
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| Inspection Report# : 2003009(pdf)
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| Barrier Integrity Significance: Mar 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE ULTRASONIC EXAMINATION PROCEDURES FOR WELDS SUBJECT TO THERMAL FATIGUE The inspectors identified a finding of very low safety significance associated with inadequate ultrasonic examination procedures used to examine Code welds subject to thermal fatigue.
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| This finding was more than minor because it affected the Barrier Integrity Cornertone objective of maintaining barrier integrity. In this example, the inadequate inservice inspection examination procedures could affect the reactor coolant system barrier integrity in that, if left uncorrected, it could become a more significant safety concern. The inspectors were concerned that if the required examination volumes were not achieved, that the large bore reactor coolant piping would be at an increased risk for failure due to thermal fatigue cracking. Because, there was no evidence of actual flaws, the inspectors concluded that this issue was a finding of very low safety significance.
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| Inspection Report# : 2004002(pdf)
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| Significance: Mar 31, 2004 Identified By: NRC Item Type: FIN Finding CONTAINMENT DRAW DOWN POST MAINTENANCE TESTING The inspectors identified a finding of very low safety significance associated with an improperly performed a secondary containment draw-down surveillance test. The licensee did not verify the train A standby gas treatment system was capable of drawing a vacuum after an initial test failure. No specific licensee procedure or instruction required by 10 CFR 50 Appendix B was violated; therefore, no violation of regulatory requirements occurred.
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| This finding was more than minor because it affected the Barrier Integrity Cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide release caused by accidents or events. The finding was of very low safety-significance because the system
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| 2Q/2004 Inspection Findings - Clinton Page 4 of 5 was demonstrated operable when properly tested. The licensee entered the event into its corrective action system and performed the test correctly after NRC involvement.
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| Inspection Report# : 2004002(pdf)
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| Significance: Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PERFORM A TS REQUIRED SURVEILLANCE The inspectors identified a finding of very low safety significance (Green) concerning the licensee's failure to verify heatup and cooldown rates in accordance with Technical Specification (TS) following a scram on December 2, 2003. This was determined to be a NCV of TS surveillance requirement 3.4.11.1.
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| This finding was more than minor because if left uncorrected, failure to perform a TS surveillance could become a more safety significant issue. This finding was not suitable for SDP evaluation but has been reviewed by NRC management and was determined to be a finding of very low safety significance. This issue may have been greater than Green if the TS temperature limitations had been exceeded and if subsequent evaluation showed a degradation of the reactor coolant system integrity.
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| Inspection Report# : 2003009(pdf)
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| Emergency Preparedness Occupational Radiation Safety Significance: Jun 30, 2004 Identified By: NRC Item Type: FIN Finding FAILURE TO MAINTAIN COLLECTIVE DOSES ALARA FOR RWP NO. 10002827.
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| A finding of very low safety significance was identified by the inspectors when the collective dose for RWP No. 10002827, "Drywell SRV Replacement," exceeded 5 person-rem and exceeded the licensee's dose estimate by more than 50 percent. This finding was related to the Human Performance cross-cutting area, in that, radiation protection personnel did not adequately evaluate the radiological consequences of a first-time evolution (i.e., the enhanced cool-down process). The Problem Identification and Resolution cross-cutting area was impacted, in that, the licensee did not identify the increased contact dose rates, which resulted in unplanned, unintended occupational collective dose for the work activity in a timely manner. This resulted in the total collective dose for the RWP of 11.839 person-rem versus a resonable re-estimate of 6.043 person-rem.
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| This issue was determined to be more than minor in that it was associated with the As Low As is Reasonably Achievable (ALARA) planning/dose projection attribute of the Occupational Radiation Safety Cornerstone, and affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation. The finding involved ALARA planning/work controls; however, the licensee's current 3-year rolling collective dose average was not greater than 240 person-rem per unit. Therefore, the finding was of very low safety significance. No violation of NRC requirements was identified.
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| Inspection Report# : 2004005(pdf)
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| Significance: Jun 30, 2004 Identified By: NRC Item Type: FIN Finding FAILURE TO MAINTAIN COLLECTIVE DOSES ALARA FOR RWP NO. 10002830.
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| A finding of very low safety significance was identified by the inspectors when the collective dose for RWP No. 10002830, "Drywell Main Steam and Feedwater Work," exceeded 5 person-rem and exceeded the dose estimate by more than 50 percent. This finding was related to the Human Performance cross-cutting area, in that, radiation protection personnel did not adequately evaluate the radiological consequences of a firts-time evolution (i.e., the enhanced cool-down process). The Problem Identification and Resolution cross-cutting area was impacted, in that, the licensee did not identify the increased contact dose rates, which resulted in unplanned, unintended occupational collective dose for the work activity in a timely manner. This resulted in the total collective dose for the RWP of 5.405 person-rem versus an estimate of 1.455 person-rem.
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| This issue was determined to be more than minor, in that, it was associated with the As Low As is Reasonably Achievable (ALARA) planning/dose projection attribute of the Occupational Radiation Safety Cornerstone, and affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation. The finding involved ALARA planning/work controls; however, the licensee's current 3-year rolling collective dose average was not greater than 240 person-rem per unit. Therefore, the finding was of very low safety significance. No violation of NRC requirements was identified.
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| Inspection Report# : 2004005(pdf)
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| 2Q/2004 Inspection Findings - Clinton Page 5 of 5 Significance: Mar 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ESTABLISH APPROPRIATE RADIOLOGICAL CONTROLS FOR A TS HIGH RADIATION AREA A finding of very low safety significance and an associated Non-Cited Violation were identified through a self-revealing event, when on February 6, 2004, an operator working in an area adjacent to the Inclined Fuel Transfer System (IFTS) shield wall in the Fuel Building received an unanticipated electronic dosimetry dose rate alarm. The licensee's subsequent investigation revealed that transfer of spent fuel bundles using the IFTS created a previously unidentified beam of radiation with dose rates in accessible areas in excess of 1000 millirem per hour, and thus the licensee had failed to control the area in accordance with Technical Specifications (i.e, appropriate barricades, postings, and locking mechanisms or flashing lights were not in place).
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| This issue was associated with the "Program and Process" attribute of the Occupational Radiation Safety Cornerstone and affected the cornerstone objective in ensuring adequate protection of worker health and safety from exposure to radiation from radioactive material. The issue was more than minor because it involved the occurrence of a potential for unplanned, unintended dose to individuals working in an inadequately controlled high radiation area resulting from conditions contrary to licensee technical specifications and NRC requirements. Based in part on: (1) the dose rates identified in area; (2) the typcial spent fuel bundle transit time; and (3) the length of time the operator was in the area, the inspectors determined that there was not an overexposure, nor was there a substantial potential for an overexposure. Therefore, the finding was of very low safety significance. One Non-Cited Violation for the failure to barricade, properly post, and establish a flashing light for the area surrounding the IFTS shield wall in accordance with Technical Specification 5.7.2 was identified.
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| Inspection Report# : 2004002(pdf)
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| Public Radiation Safety Physical Protection Physical Protection information not publicly available.
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| Miscellaneous Last modified : September 08, 2004
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| 3Q/2004 Inspection Findings - Clinton Page 1 of 6 Clinton 3Q/2004 Plant Inspection Findings Initiating Events Significance: Sep 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation PORTABLE FIRE EXTINGUISHER MISSING FROM ITS DESIGNATED STORAGE.
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| A finding of very low safety significance was identified by the inspectors for a violation of license-required fire protection program requirements. The licensee had removed a portable fire extinguisher from its designated storage location on the 828 foot elevation of containment and could not locate it. The fire marshal quickly replaced the missing extinguisher and conducted a walkdown of the containment to ensure no other portable fire extinguishers were missing from their required locations.
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| This finding was more than minor because left uncorrected, it would become a more significant safety concern. The licensee's ability to cope with fires of limited size in the area was impaired due to the insufficient number of extinguishers. The issue was of very low safety significance because there were two nearby hose stations which could be used for fire suppression activities. The issue was a Non-Cited Violation of the facility operating license section 2.F which required the implementation of the fire protection program.
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| Inspection Report# : 2004006(pdf)
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| Significance: Dec 31, 2003 Identified By: NRC Item Type: FIN Finding AUTOMATIC SHUTDOWN SIGNAL GENERATED DUE TO PERSONNEL ERROR The inspectors identified a finding of very low safety significance concerning poor operator performance following a reactor scram on December 2, 2003. The primary cause of this finding was related to the cross-cutting area of Human Performance, in that, poor performance by operations personnel resulted in a momentary loss of reactor pressure vessel level control. This loss of level resulted in a second reactor scram signal being generated. No violations of NRC requirements occurred.
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| This finding was more than minor because the finding affected the Reactor Safety/Initiating Event objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The finding did not contribute to the likelihood of a Primary or Secondary system loss of coolant accident initiator, did not contribute to both the likelihood of a reactor trip AND the likelihood that mitigation equipment or functions will not be available, and did not increase the likelihood of a fire or internal/external flood. Therefore, the finding was determined to be of very low safety significance.
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| Inspection Report# : 2003009(pdf)
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| Mitigating Systems Significance: Sep 30, 2004 Identified By: NRC Item Type: FIN Finding DIVISION-3 ESSENTIAL SWITCHGEAR HEAT REMOVAL (VX) SYSTEM TRIPPED DUE TO INADEQUATE IMPACT STATEMENT FOR MAINTENANCE.
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| A finding of very low safety significance was self-revealed during a maintenance activity when Division essential switchgear heat removal was lost as a result of an inadequate impact statement in the work order. The primary cause of this finding was related to the cross-cutting area of Human Performance. In addition to the maintenance planner missing the relationship between the safety and non-safety supply fan motors, several other opportunities to identify this inadequate impact statement were missed.
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| This finding was more than minor because with the division three essential switchgear heat removal system unavailable, the high pressure core spray system may be rendered inoperable. The issue was of very low safety significance because the initial temperature in the division three switchgear room was low and the loss of essential switchgear heat removal was of short duration, the high pressure core spray system was never actually inoperable. No violation of NRC requirements occurred.
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| Inspection Report# : 2004006(pdf)
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| 3Q/2004 Inspection Findings - Clinton Page 2 of 6 Significance: Jul 26, 2004 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO MAKE PLANT PERSONNEL AWARE OF A MODIFICATION WHICH MAY AFFECT THE PERFORMANCE OF THEIR DUTIES A finding of very low safety significance, with an associated Non-Cited Violation, was identified by the inspectors. Specifically, the licensee failed to analyze how a feedwater pump modification affected the operator's duties after an automatic shutdown. As a result of the modification, operators should have been directed, by procedure and training, to trip the "B" feedwater pump following an automatic shutdown.
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| One of the causes of this finding related to the cross-cutting area of problem identification and resolution, in that, the licensee did not identify the discrepant procedure or training during investigation of a previous event.
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| The issue was more than minor because if left uncorrected, it could be reasonably viewed as a precursor to a significant event. Specifically, it caused unnecessary complications to the automatic shutdown sequence, placed extra importance on the motor-driven reactor feedwater (MDRF) pump and could challenge the high-pressure emergency core cooling systems (ECCS) during a motor-driven feedwater pump outage.
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| The inspectors determined that the finding could not be evaluated in accordance with IMC 0609, "Significance Determination Process."
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| Therefore, this finding was reviewed by the Regional Branch Chief in accordance with IMC 0612, Section 05.04c, and determined to be of very low safety significance because the MDRF pump did start and the high pressure ECCS systems were operable. The finding was assigned to the mitigating system cornerstone. The issue was a Non-Cited Violation of Criterion II of 10 CFR 50 Appendix B. The licensee took immediate corrective action to revise the procedure, installed a robust barrier over the "A" feedwater pump control switch, and briefed all operators on the effects of the modification.
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| Inspection Report# : 2004007(pdf)
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| Significance: Jul 26, 2004 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO HAVE AN ADEQUATE OPERATING PROCEDURE A finding of very low safety significance, with an associated Non-Cited Violation, was self-revealed. Specifically, Clinton Power Station Procedure 3312.03, "Shutdown Cooling and Fuel Pool Cooling and Assist," was inadequate because it allowed the operators to create voids inside system piping while preparing to place the "B" residual heat removal (RHR) system in the shutdown cooling mode of operation. When sufficient differential pressure developed to open the RHR pump discharge check valve, about 2000 gallons of water unexpectedly drained from the reactor pressure vessel into the RHR system and produced a reactor automatic shutdown signal and Level 3 isolation on low reactor water level. The "B" RHR system was subsequently declared inoperable.
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| The finding was more than minor because it affected the Reactor Safety/Mitigating System Cornerstone and if left uncorrected, it would become a more significant safety concern. Specifically, voided piping could produce a system water hammer when the residual heat removal water pump is started in shutdown cooling mode and render the system inoperable. The finding was determined to be of very low safety significance because there was no design dificiency, no actual loss of safety function, no single train loss of safety function for greater than the Technical Specification allowed outage time and no risk due to external events. The licensee revised the shutdown cooling steps in the procedure, briefed all operators on the apparent cause, and entered the event into its corrective action system. The issue was a Non-Cited Violation of Criterion V of 10 CFR 50 Appendix B.
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| Inspection Report# : 2004007(pdf)
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| Significance: Jun 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation SLC BORON CONCENTRATION OUTSIDE TS LIMITS FOR GREATER THAN ALLOWED OUTAGE TIME.
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| A finding of very low safety significance was identified by the inspectors for the licensee's failure to take timely corrective actions after discovering that the standby liquid control (SLC) tank air-aparging valve was in the wrong position for about 2 months. This resulted in the boron concentration in the tank being outside the Technical Specification allowed limits for greater than the Technical Specification allowed action time. Once identified, the licensee restored the concentration in the tank to within acceptable limits. This finding was related to the Problem Identification and Resolution crosscutting area, in that, the concentration in the tank remained outside limits due to the licensee's failure to identify the impact of evaporation on the solution.
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| The finding was more than minor because the boron concentration being ouside the Technical Specification allowed range affects the cross-cutting attribute of SLC system performance and also affected the SLC system's availability, reliability, and capability of responding to plant events. The finding was of very low safety significance because the as-found concentration, although above technical specification limits, did not impact the safety function of the pumps. The finding was a Non-Cited Violation of 10CFR50, Appendix B, Criterion XVI which requires condtions adverse to quality be promptly identified and corrected.
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| Inspection Report# : 2004005(pdf)
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| Significance: Jun 30, 2004 Identified By: NRC
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| 3Q/2004 Inspection Findings - Clinton Page 3 of 6 Item Type: NCV NonCited Violation FAILURE TO IMPLEMENT A LOCKED VALVE PROCEDURE.
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| A finding of very low safety significance was identified by the inspectors for the licensee's failure to implement a procedure to control locked valves. Failing to have a locked valve procedure, combined with a shift supervisor marking the step which verified the position of the standby liquid control (SLC) tank air-sparging valve as "not applicable," based on the valve being a "locked valve" and no work having been done to the valve, allowed the air sparging valve to remain mispositioned while transitioning to Mode-2 and during Mode-1 operations. Once identified, the licensee placed the valve in the correct position. This issue was related to the Human Performance corsscutting area, in that, the failure to implement a procedure resulted in a mispositioned valve.
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| The finding was more than minor because the open air sparging valve created the potential for air-binding the pumps used to inject boron solution into the reactor, affecting the ability of the SLC system to shut the reactor down from a full power situation in the control rods failed to insert on a scram condition. The finding was of very low safety-significance because the deficiency, once evaluated, did not result in a loss of function per Generic Letter 91-18. The finding was a Non-Cited Violation of Technical Specification 5.4 which required the implementation of written procedures to control the locked valves in the plant.
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| Inspection Report# : 2004005(pdf)
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| Significance: Apr 07, 2004 Identified By: NRC Item Type: FIN Finding FAILURE TO EVALUATE THE EXTENT OF CONDITION OF FOREIGN MATERIAL FOUND IN THE DIVISION 1 EMERGENCY DIESEL GENERATOR STARTING AIR SYSTEM.
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| The team identified a finding of very low safety significance when the licensee failed to take appropriate steps to evaluate the extent of condition of foreign material in the starting air system of an emergency diesel generator.
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| The finding is more than minor because it is associated with the Mitigating System (MS) cornerstone attribute of equipment reliability and capability of systems that respond to initiating events to prevent undesirable circumstances. This finding was of very low safety significance because once evaluated, it did not result in a loss of function per Generic Letter 91-18 (Rev 1). No vilations of NRC requirements were identified. The licensee documented this issue in condition report 213491. Additionally the licensee established action items to evaluate the source of the foregin material found in the 1A Diesel Generator air system following the March 2004 failure.
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| Inspection Report# : 2004003(pdf)
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| Significance: Apr 07, 2004 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IDENTIFY THE EXTENT OF CONDITION FOR INCORRECT FUSES IN THE REACTOR PROTECTION SYSTEM.
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| The inspectors identified a finding of very low safety significance concerning the licesnee's failure to determine the extent of condition for improper fuses installed in the reactor protection system (RPS) electronic circuit boards. This finding was determined to be a Non-Cited Violation of 10 CFR 50 Appendix B, Criterion XVI.
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| This finding is more than minor because it affects the design and reliability of the RPS to perform its protective function of protecting the reactor core and containment. The licensee determined that although the fuses were improperly sized, the reactor protection system remained operable and could perfrom it's safety function. Therefore, this finding was determined to be of very low safety significance.
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| Inspection Report# : 2004003(pdf)
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| Significance: Mar 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation EMERGENCY CORE COOLING SYSTEM WATER HAMMER A finding of very low safety significance, with an associated Non-Cityed Violation, was self-revealed relating to a violation of the requirements of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings. The licensee failed to properly vent the high pressure core spray system before performing an integrated ECCS test resulting in a water-hammer event on the high-pressure core spray system.
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| This finding was more than minor because it affected the Mitigating Systems Cornerstone objective of maintaining mitigating systems operable. The finding was of very low safety-significance because a licensee follow-up system investigation, including a complete system walkdown by engineers, revealed that the high pressure core spray system remained operability. This issue was entered into the licensee corrective action program.
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| Inspection Report# : 2004002(pdf)
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| Significance: Mar 31, 2004 Identified By: NRC
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| 3Q/2004 Inspection Findings - Clinton Page 4 of 6 Item Type: NCV NonCited Violation DESIGN CONTROL OF MOTOR OPERATED VALVE MOUNTING BOLTS A finding of very low safety significance was identified by the inspectors for a violation of the requirements of 10 CFR 50, Appendix B, Criterion III, Design Control. Following the licensee's identification that the operator mounting bolts for several Limitorque SMB-2 actuators did not fit properly, the licensee installed bolts with thread engagement less than the required minimum. This was completed without performing the appropriate level design control review. The minimum thread engagement caused a residual heat removal system Limitorque SMB-2 valve actuator to wobble when operated. This finding affected the cross-cutting are of problem identification and resolution because initially, the licensee did not determine cause or extent of condition of the wobbly actuator.
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| This finding was more than minor because it affected the Mitigating Systems Cornerstone objective of maintaining mitigating systems operable. The finding was of very low safety-significance because an evaluation determined that the valve would have performed its safety function when called upon during a design basis seismic event. The finding was entered into the licensee corrective action program and the licensee verified the correct installation of all SMB-2 actuator mounting bolts.
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| Inspection Report# : 2004002(pdf)
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| Significance: Mar 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation INEFFECTIVE CORRECTIVE ACTION PIPE WALL THINNING The inspectors identified a finding of very low safety-significance and an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion XVI. The licensee had replaced shutdown service water (SX) system piping following cavitation induced wall thinning and weld failure leading to a through wall leak in 1999. The corrective actions included periodic non-destructive examination (NDE) monitoring of the pipe-wall for cavitation induced wall-thinning. Following an inquiry by the inspectors about heavy cavitation effects on the piping, the licensee discovered that the NDE monitoring had been performed in the wrong section of the piping. When the correct section was examined, the piping was found below manufacture's minimum allowable wall thickness. The finding affected the cross-cutting area of Human-Performance because the system manager and others had failed to identify that the corrective actions for a previous failed pipe had not been correctly implemented since 1999 and had also subsequently failed to expand the extent of condition to include verifying that all 10 predefined NDE activities established by the 1999 corrective actions were being performed in the correct location immediately downstream of SX system flow orifices.
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| The finding was more than minor because it affects the Reactor Safety/Mitigating System Cornerstone and if left uncorrected, it would become a more significant safety concern. The finding was of very low safety-significance because the SX system remained operable, both for function and for seismic considerations. The finding involved the attributes of availability and reliability of the shutdown service water system, internal flooding, and loss of heat sink as well as human performance and could have affected the mitigating systems objective of ensuring the availiability of systems that respond to initiating events to prevent undersirable consequences. The licensee entered the event into its corrective action system, performed an operability determination allowing continued use of the pipe, and replaced the piping in March 2004.
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| Inspection Report# : 2004002(pdf)
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| Significance: Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PROMPTLY IMPLEMENT CORRECTIVE ACTIONS.
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| The inspectors identified a non-cited violation of 10 CFR 50 Appendix B Criterion XVI involving the licensee's failure to promptly enter an identified condition adverse to quality into their corrective action program. This finding related to the cross-cutting area of Human Performance, in that, engineering personnel were aware of a discrepant condition on the 4160 volt Bus 1C1 Reserve Feed potential transformer cubicle door but did not correct the condition for several days.
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| The inspectors determined that this issue was more than minor because the finding could be reasonably viewed as a precursor to a significant event if left uncorrected because the station personnel could fail to evaluate non-conforming conditions which could render safety related equipment inoperable. This issue was design/seismic qualification deficiency that was determined to not cause a loss of function by the licensee's evaluation. Based on this conclusion, this finding was determined to be of very low safety significance using the Phase 1 worksheets.
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| Inspection Report# : 2003009(pdf)
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| Barrier Integrity Significance: Mar 31, 2004 Identified By: NRC Item Type: FIN Finding CONTAINMENT DRAW DOWN POST MAINTENANCE TESTING The inspectors identified a finding of very low safety significance associated with an improperly performed a secondary containment draw-down surveillance test. The licensee did not verify the train A standby gas treatment system was capable of drawing a vacuum after an initial
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| 3Q/2004 Inspection Findings - Clinton Page 5 of 6 test failure. No specific licensee procedure or instruction required by 10 CFR 50 Appendix B was violated; therefore, no violation of regulatory requirements occurred.
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| This finding was more than minor because it affected the Barrier Integrity Cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide release caused by accidents or events. The finding was of very low safety-significance because the system was demonstrated operable when properly tested. The licensee entered the event into its corrective action system and performed the test correctly after NRC involvement.
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| Inspection Report# : 2004002(pdf)
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| Significance: Mar 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE ULTRASONIC EXAMINATION PROCEDURES FOR WELDS SUBJECT TO THERMAL FATIGUE The inspectors identified a finding of very low safety significance associated with inadequate ultrasonic examination procedures used to examine Code welds subject to thermal fatigue.
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| This finding was more than minor because it affected the Barrier Integrity Cornertone objective of maintaining barrier integrity. In this example, the inadequate inservice inspection examination procedures could affect the reactor coolant system barrier integrity in that, if left uncorrected, it could become a more significant safety concern. The inspectors were concerned that if the required examination volumes were not achieved, that the large bore reactor coolant piping would be at an increased risk for failure due to thermal fatigue cracking. Because, there was no evidence of actual flaws, the inspectors concluded that this issue was a finding of very low safety significance.
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| Inspection Report# : 2004002(pdf)
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| Significance: Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PERFORM A TS REQUIRED SURVEILLANCE The inspectors identified a finding of very low safety significance (Green) concerning the licensee's failure to verify heatup and cooldown rates in accordance with Technical Specification (TS) following a scram on December 2, 2003. This was determined to be a NCV of TS surveillance requirement 3.4.11.1.
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| This finding was more than minor because if left uncorrected, failure to perform a TS surveillance could become a more safety significant issue. This finding was not suitable for SDP evaluation but has been reviewed by NRC management and was determined to be a finding of very low safety significance. This issue may have been greater than Green if the TS temperature limitations had been exceeded and if subsequent evaluation showed a degradation of the reactor coolant system integrity.
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| Inspection Report# : 2003009(pdf)
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| Emergency Preparedness Occupational Radiation Safety Significance: Jun 30, 2004 Identified By: NRC Item Type: FIN Finding FAILURE TO MAINTAIN COLLECTIVE DOSES ALARA FOR RWP NO. 10002827.
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| A finding of very low safety significance was identified by the inspectors when the collective dose for RWP No. 10002827, "Drywell SRV Replacement," exceeded 5 person-rem and exceeded the licensee's dose estimate by more than 50 percent. This finding was related to the Human Performance cross-cutting area, in that, radiation protection personnel did not adequately evaluate the radiological consequences of a first-time evolution (i.e., the enhanced cool-down process). The Problem Identification and Resolution cross-cutting area was impacted, in that, the licensee did not identify the increased contact dose rates, which resulted in unplanned, unintended occupational collective dose for the work activity in a timely manner. This resulted in the total collective dose for the RWP of 11.839 person-rem versus a resonable re-estimate of 6.043 person-rem.
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| This issue was determined to be more than minor in that it was associated with the As Low As is Reasonably Achievable (ALARA) planning/dose projection attribute of the Occupational Radiation Safety Cornerstone, and affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation. The finding involved ALARA planning/work controls; however,
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| 3Q/2004 Inspection Findings - Clinton Page 6 of 6 the licensee's current 3-year rolling collective dose average was not greater than 240 person-rem per unit. Therefore, the finding was of very low safety significance. No violation of NRC requirements was identified.
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| Inspection Report# : 2004005(pdf)
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| Significance: Jun 30, 2004 Identified By: NRC Item Type: FIN Finding FAILURE TO MAINTAIN COLLECTIVE DOSES ALARA FOR RWP NO. 10002830.
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| A finding of very low safety significance was identified by the inspectors when the collective dose for RWP No. 10002830, "Drywell Main Steam and Feedwater Work," exceeded 5 person-rem and exceeded the dose estimate by more than 50 percent. This finding was related to the Human Performance cross-cutting area, in that, radiation protection personnel did not adequately evaluate the radiological consequences of a firts-time evolution (i.e., the enhanced cool-down process). The Problem Identification and Resolution cross-cutting area was impacted, in that, the licensee did not identify the increased contact dose rates, which resulted in unplanned, unintended occupational collective dose for the work activity in a timely manner. This resulted in the total collective dose for the RWP of 5.405 person-rem versus an estimate of 1.455 person-rem.
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| This issue was determined to be more than minor, in that, it was associated with the As Low As is Reasonably Achievable (ALARA) planning/dose projection attribute of the Occupational Radiation Safety Cornerstone, and affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation. The finding involved ALARA planning/work controls; however, the licensee's current 3-year rolling collective dose average was not greater than 240 person-rem per unit. Therefore, the finding was of very low safety significance. No violation of NRC requirements was identified.
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| Inspection Report# : 2004005(pdf)
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| Significance: Mar 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ESTABLISH APPROPRIATE RADIOLOGICAL CONTROLS FOR A TS HIGH RADIATION AREA A finding of very low safety significance and an associated Non-Cited Violation were identified through a self-revealing event, when on February 6, 2004, an operator working in an area adjacent to the Inclined Fuel Transfer System (IFTS) shield wall in the Fuel Building received an unanticipated electronic dosimetry dose rate alarm. The licensee's subsequent investigation revealed that transfer of spent fuel bundles using the IFTS created a previously unidentified beam of radiation with dose rates in accessible areas in excess of 1000 millirem per hour, and thus the licensee had failed to control the area in accordance with Technical Specifications (i.e, appropriate barricades, postings, and locking mechanisms or flashing lights were not in place).
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| This issue was associated with the "Program and Process" attribute of the Occupational Radiation Safety Cornerstone and affected the cornerstone objective in ensuring adequate protection of worker health and safety from exposure to radiation from radioactive material. The issue was more than minor because it involved the occurrence of a potential for unplanned, unintended dose to individuals working in an inadequately controlled high radiation area resulting from conditions contrary to licensee technical specifications and NRC requirements.
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| Based in part on: (1) the dose rates identified in area; (2) the typcial spent fuel bundle transit time; and (3) the length of time the operator was in the area, the inspectors determined that there was not an overexposure, nor was there a substantial potential for an overexposure. Therefore, the finding was of very low safety significance. One Non-Cited Violation for the failure to barricade, properly post, and establish a flashing light for the area surrounding the IFTS shield wall in accordance with Technical Specification 5.7.2 was identified.
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| Inspection Report# : 2004002(pdf)
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| Public Radiation Safety Physical Protection Physical Protection information not publicly available.
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| Miscellaneous Last modified : December 29, 2004
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| 4Q/2004 Inspection Findings - Clinton Page 1 of 6 Clinton 4Q/2004 Plant Inspection Findings Initiating Events Significance: Sep 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation PORTABLE FIRE EXTINGUISHER MISSING FROM ITS DESIGNATED STORAGE.
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| A finding of very low safety significance was identified by the inspectors for a violation of license-required fire protection program requirements. The licensee had removed a portable fire extinguisher from its designated storage location on the 828 foot elevation of containment and could not locate it. The fire marshal quickly replaced the missing extinguisher and conducted a walkdown of the containment to ensure no other portable fire extinguishers were missing from their required locations.
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| This finding was more than minor because left uncorrected, it would become a more significant safety concern. The licensee's ability to cope with fires of limited size in the area was impaired due to the insufficient number of extinguishers. The issue was of very low safety significance because there were two nearby hose stations which could be used for fire suppression activities. The issue was a Non-Cited Violation of the facility operating license section 2.F which required the implementation of the fire protection program.
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| Inspection Report# : 2004006(pdf)
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| Mitigating Systems Significance: Sep 30, 2004 Identified By: NRC Item Type: FIN Finding DIVISION-3 ESSENTIAL SWITCHGEAR HEAT REMOVAL (VX) SYSTEM TRIPPED DUE TO INADEQUATE IMPACT STATEMENT FOR MAINTENANCE.
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| A finding of very low safety significance was self-revealed during a maintenance activity when Division essential switchgear heat removal was lost as a result of an inadequate impact statement in the work order. The primary cause of this finding was related to the cross-cutting area of Human Performance. In addition to the maintenance planner missing the relationship between the safety and non-safety supply fan motors, several other opportunities to identify this inadequate impact statement were missed.
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| This finding was more than minor because with the division three essential switchgear heat removal system unavailable, the high pressure core spray system may be rendered inoperable. The issue was of very low safety significance because the initial temperature in the division three switchgear room was low and the loss of essential switchgear heat removal was of short duration, the high pressure core spray system was never actually inoperable. No violation of NRC requirements occurred.
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| Inspection Report# : 2004006(pdf)
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| Significance: Jul 26, 2004 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO HAVE AN ADEQUATE OPERATING PROCEDURE A finding of very low safety significance, with an associated Non-Cited Violation, was self-revealed. Specifically, Clinton Power Station Procedure 3312.03, "Shutdown Cooling and Fuel Pool Cooling and Assist," was inadequate because it allowed the operators to create voids inside system piping while preparing to place the "B" residual heat removal (RHR) system in the shutdown cooling mode of operation. When sufficient differential pressure developed to open the RHR pump discharge check valve, about 2000 gallons of water unexpectedly drained from the reactor pressure vessel into the RHR system and produced a reactor automatic shutdown signal and Level 3 isolation on low reactor water level. The "B" RHR system was subsequently declared inoperable.
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| The finding was more than minor because it affected the Reactor Safety/Mitigating System Cornerstone and if left uncorrected, it would become a more significant safety concern. Specifically, voided piping could produce a system water hammer when the residual heat removal water pump is started in shutdown cooling mode and render the system inoperable. The finding was determined to be of very low safety significance because there was no design dificiency, no actual loss of safety function, no single train loss of safety function for greater than the Technical Specification allowed outage time and no risk due to external events. The licensee revised the shutdown cooling steps in the procedure, briefed all operators on the apparent cause, and entered the event into its corrective action system. The issue was a Non-Cited
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| 4Q/2004 Inspection Findings - Clinton Page 2 of 6 Violation of Criterion V of 10 CFR 50 Appendix B.
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| Inspection Report# : 2004007(pdf)
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| Significance: Jul 26, 2004 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO MAKE PLANT PERSONNEL AWARE OF A MODIFICATION WHICH MAY AFFECT THE PERFORMANCE OF THEIR DUTIES A finding of very low safety significance, with an associated Non-Cited Violation, was identified by the inspectors. Specifically, the licensee failed to analyze how a feedwater pump modification affected the operator's duties after an automatic shutdown. As a result of the modification, operators should have been directed, by procedure and training, to trip the "B" feedwater pump following an automatic shutdown.
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| One of the causes of this finding related to the cross-cutting area of problem identification and resolution, in that, the licensee did not identify the discrepant procedure or training during investigation of a previous event.
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| The issue was more than minor because if left uncorrected, it could be reasonably viewed as a precursor to a significant event. Specifically, it caused unnecessary complications to the automatic shutdown sequence, placed extra importance on the motor-driven reactor feedwater (MDRF) pump and could challenge the high-pressure emergency core cooling systems (ECCS) during a motor-driven feedwater pump outage.
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| The inspectors determined that the finding could not be evaluated in accordance with IMC 0609, "Significance Determination Process."
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| Therefore, this finding was reviewed by the Regional Branch Chief in accordance with IMC 0612, Section 05.04c, and determined to be of very low safety significance because the MDRF pump did start and the high pressure ECCS systems were operable. The finding was assigned to the mitigating system cornerstone. The issue was a Non-Cited Violation of Criterion II of 10 CFR 50 Appendix B. The licensee took immediate corrective action to revise the procedure, installed a robust barrier over the "A" feedwater pump control switch, and briefed all operators on the effects of the modification.
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| Inspection Report# : 2004007(pdf)
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| Significance: Jun 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IMPLEMENT A LOCKED VALVE PROCEDURE.
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| A finding of very low safety significance was identified by the inspectors for the licensee's failure to implement a procedure to control locked valves. Failing to have a locked valve procedure, combined with a shift supervisor marking the step which verified the position of the standby liquid control (SLC) tank air-sparging valve as "not applicable," based on the valve being a "locked valve" and no work having been done to the valve, allowed the air sparging valve to remain mispositioned while transitioning to Mode-2 and during Mode-1 operations. Once identified, the licensee placed the valve in the correct position. This issue was related to the Human Performance corsscutting area, in that, the failure to implement a procedure resulted in a mispositioned valve.
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| The finding was more than minor because the open air sparging valve created the potential for air-binding the pumps used to inject boron solution into the reactor, affecting the ability of the SLC system to shut the reactor down from a full power situation in the control rods failed to insert on a scram condition. The finding was of very low safety-significance because the deficiency, once evaluated, did not result in a loss of function per Generic Letter 91-18. The finding was a Non-Cited Violation of Technical Specification 5.4 which required the implementation of written procedures to control the locked valves in the plant.
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| Inspection Report# : 2004005(pdf)
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| Significance: Jun 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation SLC BORON CONCENTRATION OUTSIDE TS LIMITS FOR GREATER THAN ALLOWED OUTAGE TIME.
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| A finding of very low safety significance was identified by the inspectors for the licensee's failure to take timely corrective actions after discovering that the standby liquid control (SLC) tank air-aparging valve was in the wrong position for about 2 months. This resulted in the boron concentration in the tank being outside the Technical Specification allowed limits for greater than the Technical Specification allowed action time. Once identified, the licensee restored the concentration in the tank to within acceptable limits. This finding was related to the Problem Identification and Resolution crosscutting area, in that, the concentration in the tank remained outside limits due to the licensee's failure to identify the impact of evaporation on the solution.
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| The finding was more than minor because the boron concentration being ouside the Technical Specification allowed range affects the cross-cutting attribute of SLC system performance and also affected the SLC system's availability, reliability, and capability of responding to plant events. The finding was of very low safety significance because the as-found concentration, although above technical specification limits, did not impact the safety function of the pumps. The finding was a Non-Cited Violation of 10CFR50, Appendix B, Criterion XVI which requires condtions adverse to quality be promptly identified and corrected.
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| Inspection Report# : 2004005(pdf)
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| 4Q/2004 Inspection Findings - Clinton Page 3 of 6 Significance: Apr 07, 2004 Identified By: NRC Item Type: FIN Finding FAILURE TO EVALUATE THE EXTENT OF CONDITION OF FOREIGN MATERIAL FOUND IN THE DIVISION 1 EMERGENCY DIESEL GENERATOR STARTING AIR SYSTEM.
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| The team identified a finding of very low safety significance when the licensee failed to take appropriate steps to evaluate the extent of condition of foreign material in the starting air system of an emergency diesel generator.
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| The finding is more than minor because it is associated with the Mitigating System (MS) cornerstone attribute of equipment reliability and capability of systems that respond to initiating events to prevent undesirable circumstances. This finding was of very low safety significance because once evaluated, it did not result in a loss of function per Generic Letter 91-18 (Rev 1). No vilations of NRC requirements were identified. The licensee documented this issue in condition report 213491. Additionally the licensee established action items to evaluate the source of the foregin material found in the 1A Diesel Generator air system following the March 2004 failure.
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| Inspection Report# : 2004003(pdf)
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| Significance: Apr 07, 2004 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IDENTIFY THE EXTENT OF CONDITION FOR INCORRECT FUSES IN THE REACTOR PROTECTION SYSTEM.
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| The inspectors identified a finding of very low safety significance concerning the licesnee's failure to determine the extent of condition for improper fuses installed in the reactor protection system (RPS) electronic circuit boards. This finding was determined to be a Non-Cited Violation of 10 CFR 50 Appendix B, Criterion XVI.
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| This finding is more than minor because it affects the design and reliability of the RPS to perform its protective function of protecting the reactor core and containment. The licensee determined that although the fuses were improperly sized, the reactor protection system remained operable and could perfrom it's safety function. Therefore, this finding was determined to be of very low safety significance.
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| Inspection Report# : 2004003(pdf)
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| Significance: Mar 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation EMERGENCY CORE COOLING SYSTEM WATER HAMMER A finding of very low safety significance, with an associated Non-Cityed Violation, was self-revealed relating to a violation of the requirements of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings. The licensee failed to properly vent the high pressure core spray system before performing an integrated ECCS test resulting in a water-hammer event on the high-pressure core spray system.
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| This finding was more than minor because it affected the Mitigating Systems Cornerstone objective of maintaining mitigating systems operable. The finding was of very low safety-significance because a licensee follow-up system investigation, including a complete system walkdown by engineers, revealed that the high pressure core spray system remained operability. This issue was entered into the licensee corrective action program.
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| Inspection Report# : 2004002(pdf)
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| Significance: Mar 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation DESIGN CONTROL OF MOTOR OPERATED VALVE MOUNTING BOLTS A finding of very low safety significance was identified by the inspectors for a violation of the requirements of 10 CFR 50, Appendix B, Criterion III, Design Control. Following the licensee's identification that the operator mounting bolts for several Limitorque SMB-2 actuators did not fit properly, the licensee installed bolts with thread engagement less than the required minimum. This was completed without performing the appropriate level design control review. The minimum thread engagement caused a residual heat removal system Limitorque SMB-2 valve actuator to wobble when operated. This finding affected the cross-cutting are of problem identification and resolution because initially, the licensee did not determine cause or extent of condition of the wobbly actuator.
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| This finding was more than minor because it affected the Mitigating Systems Cornerstone objective of maintaining mitigating systems operable. The finding was of very low safety-significance because an evaluation determined that the valve would have performed its safety function when called upon during a design basis seismic event. The finding was entered into the licensee corrective action program and the licensee verified the correct installation of all SMB-2 actuator mounting bolts.
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| Inspection Report# : 2004002(pdf)
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| 4Q/2004 Inspection Findings - Clinton Page 4 of 6 Significance: Mar 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation INEFFECTIVE CORRECTIVE ACTION PIPE WALL THINNING The inspectors identified a finding of very low safety-significance and an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion XVI. The licensee had replaced shutdown service water (SX) system piping following cavitation induced wall thinning and weld failure leading to a through wall leak in 1999. The corrective actions included periodic non-destructive examination (NDE) monitoring of the pipe-wall for cavitation induced wall-thinning. Following an inquiry by the inspectors about heavy cavitation effects on the piping, the licensee discovered that the NDE monitoring had been performed in the wrong section of the piping. When the correct section was examined, the piping was found below manufacture's minimum allowable wall thickness. The finding affected the cross-cutting area of Human-Performance because the system manager and others had failed to identify that the corrective actions for a previous failed pipe had not been correctly implemented since 1999 and had also subsequently failed to expand the extent of condition to include verifying that all 10 predefined NDE activities established by the 1999 corrective actions were being performed in the correct location immediately downstream of SX system flow orifices.
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| The finding was more than minor because it affects the Reactor Safety/Mitigating System Cornerstone and if left uncorrected, it would become a more significant safety concern. The finding was of very low safety-significance because the SX system remained operable, both for function and for seismic considerations. The finding involved the attributes of availability and reliability of the shutdown service water system, internal flooding, and loss of heat sink as well as human performance and could have affected the mitigating systems objective of ensuring the availiability of systems that respond to initiating events to prevent undersirable consequences. The licensee entered the event into its corrective action system, performed an operability determination allowing continued use of the pipe, and replaced the piping in March 2004.
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| Inspection Report# : 2004002(pdf)
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| Barrier Integrity Significance: Mar 31, 2004 Identified By: NRC Item Type: FIN Finding CONTAINMENT DRAW DOWN POST MAINTENANCE TESTING The inspectors identified a finding of very low safety significance associated with an improperly performed a secondary containment draw-down surveillance test. The licensee did not verify the train A standby gas treatment system was capable of drawing a vacuum after an initial test failure. No specific licensee procedure or instruction required by 10 CFR 50 Appendix B was violated; therefore, no violation of regulatory requirements occurred.
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| This finding was more than minor because it affected the Barrier Integrity Cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide release caused by accidents or events. The finding was of very low safety-significance because the system was demonstrated operable when properly tested. The licensee entered the event into its corrective action system and performed the test correctly after NRC involvement.
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| Inspection Report# : 2004002(pdf)
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| Significance: Mar 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE ULTRASONIC EXAMINATION PROCEDURES FOR WELDS SUBJECT TO THERMAL FATIGUE The inspectors identified a finding of very low safety significance associated with inadequate ultrasonic examination procedures used to examine Code welds subject to thermal fatigue.
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| This finding was more than minor because it affected the Barrier Integrity Cornertone objective of maintaining barrier integrity. In this example, the inadequate inservice inspection examination procedures could affect the reactor coolant system barrier integrity in that, if left uncorrected, it could become a more significant safety concern. The inspectors were concerned that if the required examination volumes were not achieved, that the large bore reactor coolant piping would be at an increased risk for failure due to thermal fatigue cracking. Because, there was no evidence of actual flaws, the inspectors concluded that this issue was a finding of very low safety significance. This finding was determined to be a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion 1X.
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| Inspection Report# : 2004002(pdf)
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| Emergency Preparedness
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| 4Q/2004 Inspection Findings - Clinton Page 5 of 6 Occupational Radiation Safety Significance: Jun 30, 2004 Identified By: NRC Item Type: FIN Finding FAILURE TO MAINTAIN COLLECTIVE DOSES ALARA FOR RWP NO. 10002827.
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| A finding of very low safety significance was identified by the inspectors when the collective dose for RWP No. 10002827, "Drywell SRV Replacement," exceeded 5 person-rem and exceeded the licensee's dose estimate by more than 50 percent. This finding was related to the Human Performance cross-cutting area, in that, radiation protection personnel did not adequately evaluate the radiological consequences of a first-time evolution (i.e., the enhanced cool-down process). The Problem Identification and Resolution cross-cutting area was impacted, in that, the licensee did not identify the increased contact dose rates, which resulted in unplanned, unintended occupational collective dose for the work activity in a timely manner. This resulted in the total collective dose for the RWP of 11.839 person-rem versus a resonable re-estimate of 6.043 person-rem.
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| This issue was determined to be more than minor in that it was associated with the As Low As is Reasonably Achievable (ALARA) planning/dose projection attribute of the Occupational Radiation Safety Cornerstone, and affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation. The finding involved ALARA planning/work controls; however, the licensee's current 3-year rolling collective dose average was not greater than 240 person-rem per unit. Therefore, the finding was of very low safety significance. No violation of NRC requirements was identified.
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| Inspection Report# : 2004005(pdf)
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| Significance: Jun 30, 2004 Identified By: NRC Item Type: FIN Finding FAILURE TO MAINTAIN COLLECTIVE DOSES ALARA FOR RWP NO. 10002830.
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| A finding of very low safety significance was identified by the inspectors when the collective dose for RWP No. 10002830, "Drywell Main Steam and Feedwater Work," exceeded 5 person-rem and exceeded the dose estimate by more than 50 percent. This finding was related to the Human Performance cross-cutting area, in that, radiation protection personnel did not adequately evaluate the radiological consequences of a firts-time evolution (i.e., the enhanced cool-down process). The Problem Identification and Resolution cross-cutting area was impacted, in that, the licensee did not identify the increased contact dose rates, which resulted in unplanned, unintended occupational collective dose for the work activity in a timely manner. This resulted in the total collective dose for the RWP of 5.405 person-rem versus an estimate of 1.455 person-rem.
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| This issue was determined to be more than minor, in that, it was associated with the As Low As is Reasonably Achievable (ALARA) planning/dose projection attribute of the Occupational Radiation Safety Cornerstone, and affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation. The finding involved ALARA planning/work controls; however, the licensee's current 3-year rolling collective dose average was not greater than 240 person-rem per unit. Therefore, the finding was of very low safety significance. No violation of NRC requirements was identified.
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| Inspection Report# : 2004005(pdf)
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| Significance: Mar 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ESTABLISH APPROPRIATE RADIOLOGICAL CONTROLS FOR A TS HIGH RADIATION AREA A finding of very low safety significance and an associated Non-Cited Violation were identified through a self-revealing event, when on February 6, 2004, an operator working in an area adjacent to the Inclined Fuel Transfer System (IFTS) shield wall in the Fuel Building received an unanticipated electronic dosimetry dose rate alarm. The licensee's subsequent investigation revealed that transfer of spent fuel bundles using the IFTS created a previously unidentified beam of radiation with dose rates in accessible areas in excess of 1000 millirem per hour, and thus the licensee had failed to control the area in accordance with Technical Specifications (i.e, appropriate barricades, postings, and locking mechanisms or flashing lights were not in place).
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| This issue was associated with the "Program and Process" attribute of the Occupational Radiation Safety Cornerstone and affected the cornerstone objective in ensuring adequate protection of worker health and safety from exposure to radiation from radioactive material. The issue was more than minor because it involved the occurrence of a potential for unplanned, unintended dose to individuals working in an inadequately controlled high radiation area resulting from conditions contrary to licensee technical specifications and NRC requirements.
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| Based in part on: (1) the dose rates identified in area; (2) the typcial spent fuel bundle transit time; and (3) the length of time the operator was in the area, the inspectors determined that there was not an overexposure, nor was there a substantial potential for an overexposure. Therefore, the finding was of very low safety significance. One Non-Cited Violation for the failure to barricade, properly post, and establish a flashing light for the area surrounding the IFTS shield wall in accordance with Technical Specification 5.7.2 was identified.
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| Inspection Report# : 2004002(pdf)
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| 4Q/2004 Inspection Findings - Clinton Page 6 of 6 Public Radiation Safety Physical Protection Physical Protection information not publicly available.
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| Miscellaneous Last modified : March 09, 2005
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| 1Q/2005 Inspection Findings - Clinton Page 1 of 5 Clinton 1Q/2005 Plant Inspection Findings Initiating Events Significance: Mar 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ESTABLISH ADEQUATE COMPENSATORY ACTIONS (HOURLY FIRE WATCH) ACCORDING TO FIRE PROTECTION PROGRAM PROCEDURES.
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| A finding of very low safety significance was identified by the inspectors on March 17, 2005, for a violation of license-required fire protection program requirements. The licensee failed to establish adequate hourly fire watches for a failed ionization detector as required by the approved fire protection program procedure. Following the inspectors' identification of this issue, the licensee established an hourly fire watch that met the requirements and recommendations of the licensee's approved fire protection program procedures.
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| This finding was more than minor because if left uncorrected, it could become a more significant safety concern. The licensee's ability to quickly detect a fire in the area was impaired due to an insufficient number of smoke detectors. The issue was of very low safety significance because the fire protection element impacted by the finding was still expected to provide some defense-in-depth benefit due to a second fire detector located in the room. Additionally, there were two nearby hose stations which could be used for fire suppression activities. The issue was a Non-Cited Violation of the facility operating license section 2.F which required the implementation of the fire protection program.
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| Inspection Report# : 2005003(pdf)
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| Significance: Sep 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation PORTABLE FIRE EXTINGUISHER MISSING FROM ITS DESIGNATED STORAGE.
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| A finding of very low safety significance was identified by the inspectors for a violation of license-required fire protection program requirements. The licensee had removed a portable fire extinguisher from its designated storage location on the 828 foot elevation of containment and could not locate it. The fire marshal quickly replaced the missing extinguisher and conducted a walkdown of the containment to ensure no other portable fire extinguishers were missing from their required locations.
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| This finding was more than minor because left uncorrected, it would become a more significant safety concern. The licensee's ability to cope with fires of limited size in the area was impaired due to the insufficient number of extinguishers. The issue was of very low safety significance because there were two nearby hose stations which could be used for fire suppression activities. The issue was a Non-Cited Violation of the facility operating license section 2.F which required the implementation of the fire protection program.
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| Inspection Report# : 2004006(pdf)
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| Mitigating Systems Significance: Sep 30, 2004 Identified By: NRC Item Type: FIN Finding DIVISION-3 ESSENTIAL SWITCHGEAR HEAT REMOVAL (VX) SYSTEM TRIPPED DUE TO INADEQUATE IMPACT STATEMENT FOR MAINTENANCE.
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| A finding of very low safety significance was self-revealed during a maintenance activity when Division essential switchgear heat removal was lost as a result of an inadequate impact statement in the work order. The primary cause of this finding was related to the cross-cutting area of Human Performance. In addition to the maintenance planner missing the relationship between the safety and non-safety supply fan motors, several other opportunities to identify this inadequate impact statement were missed.
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| This finding was more than minor because with the division three essential switchgear heat removal system unavailable, the high pressure core spray system may be rendered inoperable. The issue was of very low safety significance because the initial temperature in the division three switchgear room was low and the loss of essential switchgear heat removal was of short duration, the high pressure core spray system was never actually inoperable. No violation of NRC requirements occurred.
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| Inspection Report# : 2004006(pdf)
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| 1Q/2005 Inspection Findings - Clinton Page 2 of 5 Significance: Jul 26, 2004 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO HAVE AN ADEQUATE OPERATING PROCEDURE A finding of very low safety significance, with an associated Non-Cited Violation, was self-revealed. Specifically, Clinton Power Station Procedure 3312.03, "Shutdown Cooling and Fuel Pool Cooling and Assist," was inadequate because it allowed the operators to create voids inside system piping while preparing to place the "B" residual heat removal (RHR) system in the shutdown cooling mode of operation. When sufficient differential pressure developed to open the RHR pump discharge check valve, about 2000 gallons of water unexpectedly drained from the reactor pressure vessel into the RHR system and produced a reactor automatic shutdown signal and Level 3 isolation on low reactor water level. The "B" RHR system was subsequently declared inoperable.
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| The finding was more than minor because it affected the Reactor Safety/Mitigating System Cornerstone and if left uncorrected, it would become a more significant safety concern. Specifically, voided piping could produce a system water hammer when the residual heat removal water pump is started in shutdown cooling mode and render the system inoperable. The finding was determined to be of very low safety significance because there was no design dificiency, no actual loss of safety function, no single train loss of safety function for greater than the Technical Specification allowed outage time and no risk due to external events. The licensee revised the shutdown cooling steps in the procedure, briefed all operators on the apparent cause, and entered the event into its corrective action system. The issue was a Non-Cited Violation of Criterion V of 10 CFR 50 Appendix B.
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| Inspection Report# : 2004007(pdf)
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| Significance: Jul 26, 2004 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO MAKE PLANT PERSONNEL AWARE OF A MODIFICATION WHICH MAY AFFECT THE PERFORMANCE OF THEIR DUTIES A finding of very low safety significance, with an associated Non-Cited Violation, was identified by the inspectors. Specifically, the licensee failed to analyze how a feedwater pump modification affected the operator's duties after an automatic shutdown. As a result of the modification, operators should have been directed, by procedure and training, to trip the "B" feedwater pump following an automatic shutdown.
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| One of the causes of this finding related to the cross-cutting area of problem identification and resolution, in that, the licensee did not identify the discrepant procedure or training during investigation of a previous event.
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| The issue was more than minor because if left uncorrected, it could be reasonably viewed as a precursor to a significant event. Specifically, it caused unnecessary complications to the automatic shutdown sequence, placed extra importance on the motor-driven reactor feedwater (MDRF) pump and could challenge the high-pressure emergency core cooling systems (ECCS) during a motor-driven feedwater pump outage.
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| The inspectors determined that the finding could not be evaluated in accordance with IMC 0609, "Significance Determination Process."
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| Therefore, this finding was reviewed by the Regional Branch Chief in accordance with IMC 0612, Section 05.04c, and determined to be of very low safety significance because the MDRF pump did start and the high pressure ECCS systems were operable. The finding was assigned to the mitigating system cornerstone. The issue was a Non-Cited Violation of Criterion II of 10 CFR 50 Appendix B. The licensee took immediate corrective action to revise the procedure, installed a robust barrier over the "A" feedwater pump control switch, and briefed all operators on the effects of the modification.
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| Inspection Report# : 2004007(pdf)
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| Significance: Jun 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation SLC BORON CONCENTRATION OUTSIDE TS LIMITS FOR GREATER THAN ALLOWED OUTAGE TIME.
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| A finding of very low safety significance was identified by the inspectors for the licensee's failure to take timely corrective actions after discovering that the standby liquid control (SLC) tank air-aparging valve was in the wrong position for about 2 months. This resulted in the boron concentration in the tank being outside the Technical Specification allowed limits for greater than the Technical Specification allowed action time. Once identified, the licensee restored the concentration in the tank to within acceptable limits. This finding was related to the Problem Identification and Resolution crosscutting area, in that, the concentration in the tank remained outside limits due to the licensee's failure to identify the impact of evaporation on the solution.
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| The finding was more than minor because the boron concentration being ouside the Technical Specification allowed range affects the cross-cutting attribute of SLC system performance and also affected the SLC system's availability, reliability, and capability of responding to plant events. The finding was of very low safety significance because the as-found concentration, although above technical specification limits, did not impact the safety function of the pumps. The finding was a Non-Cited Violation of 10CFR50, Appendix B, Criterion XVI which requires condtions adverse to quality be promptly identified and corrected.
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| Inspection Report# : 2004005(pdf)
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| Significance: Jun 30, 2004 Identified By: NRC
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| 1Q/2005 Inspection Findings - Clinton Page 3 of 5 Item Type: NCV NonCited Violation FAILURE TO IMPLEMENT A LOCKED VALVE PROCEDURE.
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| A finding of very low safety significance was identified by the inspectors for the licensee's failure to implement a procedure to control locked valves. Failing to have a locked valve procedure, combined with a shift supervisor marking the step which verified the position of the standby liquid control (SLC) tank air-sparging valve as "not applicable," based on the valve being a "locked valve" and no work having been done to the valve, allowed the air sparging valve to remain mispositioned while transitioning to Mode-2 and during Mode-1 operations. Once identified, the licensee placed the valve in the correct position. This issue was related to the Human Performance corsscutting area, in that, the failure to implement a procedure resulted in a mispositioned valve.
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| The finding was more than minor because the open air sparging valve created the potential for air-binding the pumps used to inject boron solution into the reactor, affecting the ability of the SLC system to shut the reactor down from a full power situation in the control rods failed to insert on a scram condition. The finding was of very low safety-significance because the deficiency, once evaluated, did not result in a loss of function per Generic Letter 91-18. The finding was a Non-Cited Violation of Technical Specification 5.4 which required the implementation of written procedures to control the locked valves in the plant.
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| Inspection Report# : 2004005(pdf)
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| Significance: Apr 07, 2004 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IDENTIFY THE EXTENT OF CONDITION FOR INCORRECT FUSES IN THE REACTOR PROTECTION SYSTEM.
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| The inspectors identified a finding of very low safety significance concerning the licesnee's failure to determine the extent of condition for improper fuses installed in the reactor protection system (RPS) electronic circuit boards. This finding was determined to be a Non-Cited Violation of 10 CFR 50 Appendix B, Criterion XVI.
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| This finding is more than minor because it affects the design and reliability of the RPS to perform its protective function of protecting the reactor core and containment. The licensee determined that although the fuses were improperly sized, the reactor protection system remained operable and could perfrom it's safety function. Therefore, this finding was determined to be of very low safety significance.
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| Inspection Report# : 2004003(pdf)
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| Significance: Apr 07, 2004 Identified By: NRC Item Type: FIN Finding FAILURE TO EVALUATE THE EXTENT OF CONDITION OF FOREIGN MATERIAL FOUND IN THE DIVISION 1 EMERGENCY DIESEL GENERATOR STARTING AIR SYSTEM.
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| The team identified a finding of very low safety significance when the licensee failed to take appropriate steps to evaluate the extent of condition of foreign material in the starting air system of an emergency diesel generator.
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| The finding is more than minor because it is associated with the Mitigating System (MS) cornerstone attribute of equipment reliability and capability of systems that respond to initiating events to prevent undesirable circumstances. This finding was of very low safety significance because once evaluated, it did not result in a loss of function per Generic Letter 91-18 (Rev 1). No vilations of NRC requirements were identified. The licensee documented this issue in condition report 213491. Additionally the licensee established action items to evaluate the source of the foregin material found in the 1A Diesel Generator air system following the March 2004 failure.
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| Inspection Report# : 2004003(pdf)
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| Barrier Integrity Significance: Mar 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW PROCEDURE AND APPROPRIATELY ANNOTATE PORTIONS AS NOT APPLICABLE DURING THE PERFORMANCE OF REQUIRED CALIBRATION PROCEDURE IN ACCORDANCE WITH TS 5.4.1.
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| Through a self-revealing event (unexpected de-energized relay found during maintenance) the inspectors identified a Non-Cited Violation (NCV) of very low safety signficiance. This finding resulted from licensee personnel incorrectly designating procedureal steps as not applicable during the performance of a calibration procedure, Clinton Power Station (CPS) 9432.60, "Channel Functional Test for Containment Building Exhaust Radiation Monitor," required by Technical Specifications. In Issue Report (IR) 289643, the licensee documented that with the realy de-energized the affected primary containment isolation valve cannot by opened without taking the corresponding Division 2 LOCA BYPASS switch to the BYPASS position (an action administratively controlled by Operations).
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| The inspectors determined that the finding was greater than minor because this issue could be reasonably viewed as a precursor to a more
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| 1Q/2005 Inspection Findings - Clinton Page 4 of 5 significant event. Additionally, this finding was associated with the Barrier Integrity Cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radioactive releases caused by accidents or events. The finding was of very low safety significance because this issue did not cause an actual open pathway in the physical integrity of reactor containment. The licensee documented the issue in IR 289643 and generated corrective actions as the result of a human performance investigation report being performed. These corrective actions included revising CPS 9432.60 to clearly identify the reason for placing the switch to BYPASS.
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| Inspection Report# : 2005003(pdf)
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| Emergency Preparedness Occupational Radiation Safety Significance: Jun 30, 2004 Identified By: NRC Item Type: FIN Finding FAILURE TO MAINTAIN COLLECTIVE DOSES ALARA FOR RWP NO. 10002827.
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| A finding of very low safety significance was identified by the inspectors when the collective dose for RWP No. 10002827, "Drywell SRV Replacement," exceeded 5 person-rem and exceeded the licensee's dose estimate by more than 50 percent. This finding was related to the Human Performance cross-cutting area, in that, radiation protection personnel did not adequately evaluate the radiological consequences of a first-time evolution (i.e., the enhanced cool-down process). The Problem Identification and Resolution cross-cutting area was impacted, in that, the licensee did not identify the increased contact dose rates, which resulted in unplanned, unintended occupational collective dose for the work activity in a timely manner. This resulted in the total collective dose for the RWP of 11.839 person-rem versus a resonable re-estimate of 6.043 person-rem.
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| This issue was determined to be more than minor in that it was associated with the As Low As is Reasonably Achievable (ALARA) planning/dose projection attribute of the Occupational Radiation Safety Cornerstone, and affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation. The finding involved ALARA planning/work controls; however, the licensee's current 3-year rolling collective dose average was not greater than 240 person-rem per unit. Therefore, the finding was of very low safety significance. No violation of NRC requirements was identified.
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| Inspection Report# : 2004005(pdf)
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| Significance: Jun 30, 2004 Identified By: NRC Item Type: FIN Finding FAILURE TO MAINTAIN COLLECTIVE DOSES ALARA FOR RWP NO. 10002830.
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| A finding of very low safety significance was identified by the inspectors when the collective dose for RWP No. 10002830, "Drywell Main Steam and Feedwater Work," exceeded 5 person-rem and exceeded the dose estimate by more than 50 percent. This finding was related to the Human Performance cross-cutting area, in that, radiation protection personnel did not adequately evaluate the radiological consequences of a firts-time evolution (i.e., the enhanced cool-down process). The Problem Identification and Resolution cross-cutting area was impacted, in that, the licensee did not identify the increased contact dose rates, which resulted in unplanned, unintended occupational collective dose for the work activity in a timely manner. This resulted in the total collective dose for the RWP of 5.405 person-rem versus an estimate of 1.455 person-rem.
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| This issue was determined to be more than minor, in that, it was associated with the As Low As is Reasonably Achievable (ALARA) planning/dose projection attribute of the Occupational Radiation Safety Cornerstone, and affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation. The finding involved ALARA planning/work controls; however, the licensee's current 3-year rolling collective dose average was not greater than 240 person-rem per unit. Therefore, the finding was of very low safety significance. No violation of NRC requirements was identified.
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| Inspection Report# : 2004005(pdf)
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| Public Radiation Safety Significance: Mar 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation
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| 1Q/2005 Inspection Findings - Clinton Page 5 of 5 FAILURE TO MAINTAIN CONTROL OF LICENSED RADIOACTIVE MATERIAL IN ACCORDANCE WITH 10 CFR 20, SUBPART 1.
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| A finding of very low safety significance and an associated Non-Cited Violation were identified through a self-revealing event on October 7, 2004, when licensee personnel discovered that three nuclear instrument detectors (containing a very small amount of radioactive material) were not adequately controlled. Licensee personnel believed that the material was contained in a small container which was sealed in 1991 as part of a disposition plan for the defective instruments. The licensee' search of other material containers and documentation failed to identify the final disposition of the radioactive material.
| |
| The issue was more than minor because it was associated with the Human Performance and Programs/Process attributes of the Public Radiation Safety Cornerstone and affected the cornerstone objective of ensuring adequate protection of public health and safety from exposure to radioactive materials potentially released into the public domain. Based on various dose calculation scenarios, the very small amount of missing radioactive material would contribute a negligible radiological dose if a member of the public were to be exposed to the material.
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| Additionally, the inspectors determined that the licensee did not have any prior radioactive material control occurrences in the previous 8 quarters. Therefore, the finding was of very low safety significance. The licensee's corrective actions for this issue included the development of procedural guidance which prohibits removing nuclear instrument detectors from the cabling as part of a disposition plan for defective units.
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| One Non-Cited Vilation for the failure to control licensed radioactive material in accordance with 10 CFR 20, Subpart 1, was identified.
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| Inspection Report# : 2005003(pdf)
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| Physical Protection Physical Protection information not publicly available.
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| Miscellaneous Last modified : June 17, 2005
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| 2Q/2005 Inspection Findings - Clinton Page 1 of 4 Clinton 2Q/2005 Plant Inspection Findings Initiating Events Significance: Mar 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ESTABLISH ADEQUATE COMPENSATORY ACTIONS (HOURLY FIRE WATCH) ACCORDING TO FIRE PROTECTION PROGRAM PROCEDURES.
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| A finding of very low safety significance was identified by the inspectors on March 17, 2005, for a violation of license-required fire protection program requirements. The licensee failed to establish adequate hourly fire watches for a failed ionization detector as required by the approved fire protection program procedure. Following the inspectors' identification of this issue, the licensee established an hourly fire watch that met the requirements and recommendations of the licensee's approved fire protection program procedures.
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| This finding was more than minor because if left uncorrected, it could become a more significant safety concern. The licensee's ability to quickly detect a fire in the area was impaired due to an insufficient number of smoke detectors. The issue was of very low safety significance because the fire protection element impacted by the finding was still expected to provide some defense-in-depth benefit due to a second fire detector located in the room. Additionally, there were two nearby hose stations which could be used for fire suppression activities. The issue was a Non-Cited Violation of the facility operating license section 2.F which required the implementation of the fire protection program.
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| Inspection Report# : 2005003(pdf)
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| Significance: Sep 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation PORTABLE FIRE EXTINGUISHER MISSING FROM ITS DESIGNATED STORAGE.
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| A finding of very low safety significance was identified by the inspectors for a violation of license-required fire protection program requirements. The licensee had removed a portable fire extinguisher from its designated storage location on the 828 foot elevation of containment and could not locate it. The fire marshal quickly replaced the missing extinguisher and conducted a walkdown of the containment to ensure no other portable fire extinguishers were missing from their required locations.
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| This finding was more than minor because left uncorrected, it would become a more significant safety concern. The licensee's ability to cope with fires of limited size in the area was impaired due to the insufficient number of extinguishers. The issue was of very low safety significance because there were two nearby hose stations which could be used for fire suppression activities. The issue was a Non-Cited Violation of the facility operating license section 2.F which required the implementation of the fire protection program.
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| Inspection Report# : 2004006(pdf)
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| Mitigating Systems Significance: Jun 30, 2005 Identified By: NRC Item Type: FIN Finding IMPROPERLY SECURED 4160V EQUIPMENT DOORS In December 2003 the inspectors identified a discrepant condition on the 4160 volt Bus 1C1 Reserve Feed potential transformer cubicle. The inspectors considered this to be an inspection finding with no violations of NRC requirements identified.
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| The inspectors determined that the issue was more than minor because the finding could be reasonably viewed as a precursor to a significant event, which if left uncorrected, could render safety related equipment inoperable. The issue was a design/seismic qualification deficiency that was determined not to cause a loss of a safety related function by the licensee's evaluation. Based on this conclusion, this finding was determined to be of very low safety significance using the Phase 1 worksheets.
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| Inspection Report# : 2005007(pdf)
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| Significance: Sep 30, 2004 Identified By: NRC
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| 2Q/2005 Inspection Findings - Clinton Page 2 of 4 Item Type: FIN Finding DIVISION-3 ESSENTIAL SWITCHGEAR HEAT REMOVAL (VX) SYSTEM TRIPPED DUE TO INADEQUATE IMPACT STATEMENT FOR MAINTENANCE.
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| A finding of very low safety significance was self-revealed during a maintenance activity when Division essential switchgear heat removal was lost as a result of an inadequate impact statement in the work order. The primary cause of this finding was related to the cross-cutting area of Human Performance. In addition to the maintenance planner missing the relationship between the safety and non-safety supply fan motors, several other opportunities to identify this inadequate impact statement were missed.
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| This finding was more than minor because with the division three essential switchgear heat removal system unavailable, the high pressure core spray system may be rendered inoperable. The issue was of very low safety significance because the initial temperature in the division three switchgear room was low and the loss of essential switchgear heat removal was of short duration, the high pressure core spray system was never actually inoperable. No violation of NRC requirements occurred.
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| Inspection Report# : 2004006(pdf)
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| Significance: Jul 26, 2004 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO MAKE PLANT PERSONNEL AWARE OF A MODIFICATION WHICH MAY AFFECT THE PERFORMANCE OF THEIR DUTIES A finding of very low safety significance, with an associated Non-Cited Violation, was identified by the inspectors. Specifically, the licensee failed to analyze how a feedwater pump modification affected the operator's duties after an automatic shutdown. As a result of the modification, operators should have been directed, by procedure and training, to trip the "B" feedwater pump following an automatic shutdown.
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| One of the causes of this finding related to the cross-cutting area of problem identification and resolution, in that, the licensee did not identify the discrepant procedure or training during investigation of a previous event.
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| The issue was more than minor because if left uncorrected, it could be reasonably viewed as a precursor to a significant event. Specifically, it caused unnecessary complications to the automatic shutdown sequence, placed extra importance on the motor-driven reactor feedwater (MDRF) pump and could challenge the high-pressure emergency core cooling systems (ECCS) during a motor-driven feedwater pump outage.
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| The inspectors determined that the finding could not be evaluated in accordance with IMC 0609, "Significance Determination Process."
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| Therefore, this finding was reviewed by the Regional Branch Chief in accordance with IMC 0612, Section 05.04c, and determined to be of very low safety significance because the MDRF pump did start and the high pressure ECCS systems were operable. The finding was assigned to the mitigating system cornerstone. The issue was a Non-Cited Violation of Criterion II of 10 CFR 50 Appendix B. The licensee took immediate corrective action to revise the procedure, installed a robust barrier over the "A" feedwater pump control switch, and briefed all operators on the effects of the modification.
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| Inspection Report# : 2004007(pdf)
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| Significance: Jul 26, 2004 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO HAVE AN ADEQUATE OPERATING PROCEDURE A finding of very low safety significance, with an associated Non-Cited Violation, was self-revealed. Specifically, Clinton Power Station Procedure 3312.03, "Shutdown Cooling and Fuel Pool Cooling and Assist," was inadequate because it allowed the operators to create voids inside system piping while preparing to place the "B" residual heat removal (RHR) system in the shutdown cooling mode of operation. When sufficient differential pressure developed to open the RHR pump discharge check valve, about 2000 gallons of water unexpectedly drained from the reactor pressure vessel into the RHR system and produced a reactor automatic shutdown signal and Level 3 isolation on low reactor water level. The "B" RHR system was subsequently declared inoperable.
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| The finding was more than minor because it affected the Reactor Safety/Mitigating System Cornerstone and if left uncorrected, it would become a more significant safety concern. Specifically, voided piping could produce a system water hammer when the residual heat removal water pump is started in shutdown cooling mode and render the system inoperable. The finding was determined to be of very low safety significance because there was no design dificiency, no actual loss of safety function, no single train loss of safety function for greater than the Technical Specification allowed outage time and no risk due to external events. The licensee revised the shutdown cooling steps in the procedure, briefed all operators on the apparent cause, and entered the event into its corrective action system. The issue was a Non-Cited Violation of Criterion V of 10 CFR 50 Appendix B.
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| Inspection Report# : 2004007(pdf)
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| Barrier Integrity Significance: Mar 31, 2005 Identified By: NRC
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| 2Q/2005 Inspection Findings - Clinton Page 3 of 4 Item Type: NCV NonCited Violation FAILURE TO FOLLOW PROCEDURE AND APPROPRIATELY ANNOTATE PORTIONS AS NOT APPLICABLE DURING THE PERFORMANCE OF REQUIRED CALIBRATION PROCEDURE IN ACCORDANCE WITH TS 5.4.1.
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| Through a self-revealing event (unexpected de-energized relay found during maintenance) the inspectors identified a Non-Cited Violation (NCV) of very low safety signficiance. This finding resulted from licensee personnel incorrectly designating procedureal steps as not applicable during the performance of a calibration procedure, Clinton Power Station (CPS) 9432.60, "Channel Functional Test for Containment Building Exhaust Radiation Monitor," required by Technical Specifications. In Issue Report (IR) 289643, the licensee documented that with the realy de-energized the affected primary containment isolation valve cannot by opened without taking the corresponding Division 2 LOCA BYPASS switch to the BYPASS position (an action administratively controlled by Operations).
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| The inspectors determined that the finding was greater than minor because this issue could be reasonably viewed as a precursor to a more significant event. Additionally, this finding was associated with the Barrier Integrity Cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radioactive releases caused by accidents or events. The finding was of very low safety significance because this issue did not cause an actual open pathway in the physical integrity of reactor containment. The licensee documented the issue in IR 289643 and generated corrective actions as the result of a human performance investigation report being performed. These corrective actions included revising CPS 9432.60 to clearly identify the reason for placing the switch to BYPASS.
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| Inspection Report# : 2005003(pdf)
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| Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Significance: Mar 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO MAINTAIN CONTROL OF LICENSED RADIOACTIVE MATERIAL IN ACCORDANCE WITH 10 CFR 20, SUBPART 1.
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| A finding of very low safety significance and an associated Non-Cited Violation were identified through a self-revealing event on October 7, 2004, when licensee personnel discovered that three nuclear instrument detectors (containing a very small amount of radioactive material) were not adequately controlled. Licensee personnel believed that the material was contained in a small container which was sealed in 1991 as part of a disposition plan for the defective instruments. The licensee' search of other material containers and documentation failed to identify the final disposition of the radioactive material.
| |
| The issue was more than minor because it was associated with the Human Performance and Programs/Process attributes of the Public Radiation Safety Cornerstone and affected the cornerstone objective of ensuring adequate protection of public health and safety from exposure to radioactive materials potentially released into the public domain. Based on various dose calculation scenarios, the very small amount of missing radioactive material would contribute a negligible radiological dose if a member of the public were to be exposed to the material.
| |
| Additionally, the inspectors determined that the licensee did not have any prior radioactive material control occurrences in the previous 8 quarters. Therefore, the finding was of very low safety significance. The licensee's corrective actions for this issue included the development of procedural guidance which prohibits removing nuclear instrument detectors from the cabling as part of a disposition plan for defective units.
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| One Non-Cited Vilation for the failure to control licensed radioactive material in accordance with 10 CFR 20, Subpart 1, was identified.
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| Inspection Report# : 2005003(pdf)
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| Physical Protection Physical Protection information not publicly available.
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| 2Q/2005 Inspection Findings - Clinton Page 4 of 4 Miscellaneous Last modified : August 24, 2005
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| 3Q/2005 Inspection Findings - Clinton Page 1 of 3 Clinton 3Q/2005 Plant Inspection Findings Initiating Events Significance: Sep 30, 2005 Identified By: NRC Item Type: FIN Finding PERFORMANCE OF WORK IN THE OFF-GAS SYSTEM THAT RESULTED IN A SUBSEQUENT LOSS IN OFF-GAS SYSTEM FLOW AND THE OPERATORS PERFORMING A RAPID POWER REDUCTION.
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| On August 29, 2005, a finding of very low safety significance was self revealed following the performance of work in the off-gas system that resulted in a subsequent loss in off-gas system flow and the operators performing a rapid power reduction. The finding involved the failure to stroke a gas dryer inlet valve to ensure the valve would operate following a packing adjustment. This issue was caused by poor work practices and communication by licensee personnel.
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| The issue was more than minor because it affected the Reactor Safety/Initiating Event cornerstone objective of limiting the likelihood of those events that upset plant stability. The finding was of very low safety significance because it would not affect the availability of mitigating systems or functions even if it had resulted in a plant trip. No violation of NRC requirements occurred. The finding also affected the cross cutting area of Human Performance.
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| Inspection Report# : 2005008(pdf)
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| Significance: Sep 30, 2005 Identified By: NRC Item Type: FIN Finding THE LICENSEE FAILED TO TAKE PROMPT ACTION TO CORRECT A PROBLEM WITHIN THE ELECTRO-HYDRAULIC CONTROL SYSTEM.
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| On July 17, 2005, a finding of very low safety significance was identified by the inspectors when the licensee failed to take prompt action to correct a problem within the electro-hydraulic control system. In April 2005, one main turbine combined intermediate valve went shut at power due to a clogged servo valve strainer, causing a plant transient. The licensee identified that other main turbine valves were susceptible to the same failure, but did not take action to correct the problem until after a second combined intermediate valve went shut three months later, causing a second plant transient.
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| The issue was more than minor becuse the licensee knew of the degraded condition and associated risks and failed to correct the problem before it resulted in a second plant transient requiring operators to respond. The finding was of very low safety significance because it would not affect the availability of mitigating systems or functions even if it had resulted in a plant trip. No violation of NRC requirements occurred.
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| Inspection Report# : 2005008(pdf)
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| Significance: Mar 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ESTABLISH ADEQUATE COMPENSATORY ACTIONS (HOURLY FIRE WATCH) ACCORDING TO FIRE PROTECTION PROGRAM PROCEDURES.
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| A finding of very low safety significance was identified by the inspectors on March 17, 2005, for a violation of license-required fire protection program requirements. The licensee failed to establish adequate hourly fire watches for a failed ionization detector as required by the approved fire protection program procedure. Following the inspectors' identification of this issue, the licensee established an hourly fire watch that met the requirements and recommendations of the licensee's approved fire protection program procedures.
| |
| This finding was more than minor because if left uncorrected, it could become a more significant safety concern. The licensee's ability to quickly detect a fire in the area was impaired due to an insufficient number of smoke detectors. The issue was of very low safety significance because the fire protection element impacted by the finding was still expected to provide some defense-in-depth benefit due to a second fire detector located in the room. Additionally, there were two nearby hose stations which could be used for fire suppression activities. The issue was a Non-Cited Violation of the facility operating license section 2.F which required the implementation of the fire protection program.
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| Inspection Report# : 2005003(pdf)
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| Mitigating Systems
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| 3Q/2005 Inspection Findings - Clinton Page 2 of 3 Significance: Jun 30, 2005 Identified By: NRC Item Type: FIN Finding IMPROPERLY SECURED 4160V EQUIPMENT DOORS In December 2003 the inspectors identified a discrepant condition on the 4160 volt Bus 1C1 Reserve Feed potential transformer cubicle. The inspectors considered this to be an inspection finding with no violations of NRC requirements identified.
| |
| The inspectors determined that the issue was more than minor because the finding could be reasonably viewed as a precursor to a significant event, which if left uncorrected, could render safety related equipment inoperable. The issue was a design/seismic qualification deficiency that was determined not to cause a loss of a safety related function by the licensee's evaluation. Based on this conclusion, this finding was determined to be of very low safety significance using the Phase 1 worksheets.
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| Inspection Report# : 2005007(pdf)
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| Significance: Sep 30, 2004 Identified By: NRC Item Type: FIN Finding DIVISION-3 ESSENTIAL SWITCHGEAR HEAT REMOVAL (VX) SYSTEM TRIPPED DUE TO INADEQUATE IMPACT STATEMENT FOR MAINTENANCE.
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| A finding of very low safety significance was self-revealed during a maintenance activity when Division essential switchgear heat removal was lost as a result of an inadequate impact statement in the work order. The primary cause of this finding was related to the cross-cutting area of Human Performance. In addition to the maintenance planner missing the relationship between the safety and non-safety supply fan motors, several other opportunities to identify this inadequate impact statement were missed.
| |
| This finding was more than minor because with the division three essential switchgear heat removal system unavailable, the high pressure core spray system may be rendered inoperable. The issue was of very low safety significance because the initial temperature in the division three switchgear room was low and the loss of essential switchgear heat removal was of short duration, the high pressure core spray system was never actually inoperable. No violation of NRC requirements occurred.
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| Inspection Report# : 2004006(pdf)
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| Barrier Integrity Significance: Mar 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW PROCEDURE AND APPROPRIATELY ANNOTATE PORTIONS AS NOT APPLICABLE DURING THE PERFORMANCE OF REQUIRED CALIBRATION PROCEDURE IN ACCORDANCE WITH TS 5.4.1.
| |
| Through a self-revealing event (unexpected de-energized relay found during maintenance) the inspectors identified a Non-Cited Violation (NCV) of very low safety signficiance. This finding resulted from licensee personnel incorrectly designating procedureal steps as not applicable during the performance of a calibration procedure, Clinton Power Station (CPS) 9432.60, "Channel Functional Test for Containment Building Exhaust Radiation Monitor," required by Technical Specifications. In Issue Report (IR) 289643, the licensee documented that with the realy de-energized the affected primary containment isolation valve cannot by opened without taking the corresponding Division 2 LOCA BYPASS switch to the BYPASS position (an action administratively controlled by Operations).
| |
| The inspectors determined that the finding was greater than minor because this issue could be reasonably viewed as a precursor to a more significant event. Additionally, this finding was associated with the Barrier Integrity Cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radioactive releases caused by accidents or events. The finding was of very low safety significance because this issue did not cause an actual open pathway in the physical integrity of reactor containment. The licensee documented the issue in IR 289643 and generated corrective actions as the result of a human performance investigation report being performed. These corrective actions included revising CPS 9432.60 to clearly identify the reason for placing the switch to BYPASS.
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| Inspection Report# : 2005003(pdf)
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| Emergency Preparedness
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| 3Q/2005 Inspection Findings - Clinton Page 3 of 3 Occupational Radiation Safety Public Radiation Safety Significance: Mar 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO MAINTAIN CONTROL OF LICENSED RADIOACTIVE MATERIAL IN ACCORDANCE WITH 10 CFR 20, SUBPART 1.
| |
| A finding of very low safety significance and an associated Non-Cited Violation were identified through a self-revealing event on October 7, 2004, when licensee personnel discovered that three nuclear instrument detectors (containing a very small amount of radioactive material) were not adequately controlled. Licensee personnel believed that the material was contained in a small container which was sealed in 1991 as part of a disposition plan for the defective instruments. The licensee' search of other material containers and documentation failed to identify the final disposition of the radioactive material.
| |
| The issue was more than minor because it was associated with the Human Performance and Programs/Process attributes of the Public Radiation Safety Cornerstone and affected the cornerstone objective of ensuring adequate protection of public health and safety from exposure to radioactive materials potentially released into the public domain. Based on various dose calculation scenarios, the very small amount of missing radioactive material would contribute a negligible radiological dose if a member of the public were to be exposed to the material.
| |
| Additionally, the inspectors determined that the licensee did not have any prior radioactive material control occurrences in the previous 8 quarters. Therefore, the finding was of very low safety significance. The licensee's corrective actions for this issue included the development of procedural guidance which prohibits removing nuclear instrument detectors from the cabling as part of a disposition plan for defective units.
| |
| One Non-Cited Vilation for the failure to control licensed radioactive material in accordance with 10 CFR 20, Subpart 1, was identified.
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| Inspection Report# : 2005003(pdf)
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| Physical Protection Physical Protection information not publicly available.
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| Miscellaneous Last modified : November 30, 2005
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| 4Q/2005 Inspection Findings - Clinton Page 1 of 4 Clinton 4Q/2005 Plant Inspection Findings Initiating Events Significance: Sep 30, 2005 Identified By: NRC Item Type: FIN Finding PERFORMANCE OF WORK IN THE OFF-GAS SYSTEM THAT RESULTED IN A SUBSEQUENT LOSS IN OFF-GAS SYSTEM FLOW AND THE OPERATORS PERFORMING A RAPID POWER REDUCTION.
| |
| On August 29, 2005, a finding of very low safety significance was self revealed following the performance of work in the off-gas system that resulted in a subsequent loss in off-gas system flow and the operators performing a rapid power reduction. The finding involved the failure to stroke a gas dryer inlet valve to ensure the valve would operate following a packing adjustment. This issue was caused by poor work practices and communication by licensee personnel.
| |
| The issue was more than minor because it affected the Reactor Safety/Initiating Event cornerstone objective of limiting the likelihood of those events that upset plant stability. The finding was of very low safety significance because it would not affect the availability of mitigating systems or functions even if it had resulted in a plant trip. No violation of NRC requirements occurred. The finding also affected the cross cutting area of Human Performance.
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| Inspection Report# : 2005008(pdf)
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| Significance: Sep 30, 2005 Identified By: NRC Item Type: FIN Finding THE LICENSEE FAILED TO TAKE PROMPT ACTION TO CORRECT A PROBLEM WITHIN THE ELECTRO-HYDRAULIC CONTROL SYSTEM.
| |
| On July 17, 2005, a finding of very low safety significance was identified by the inspectors when the licensee failed to take prompt action to correct a problem within the electro-hydraulic control system. In April 2005, one main turbine combined intermediate valve went shut at power due to a clogged servo valve strainer, causing a plant transient. The licensee identified that other main turbine valves were susceptible to the same failure, but did not take action to correct the problem until after a second combined intermediate valve went shut three months later, causing a second plant transient.
| |
| The issue was more than minor becuse the licensee knew of the degraded condition and associated risks and failed to correct the problem before it resulted in a second plant transient requiring operators to respond. The finding was of very low safety significance because it would not affect the availability of mitigating systems or functions even if it had resulted in a plant trip. No violation of NRC requirements occurred.
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| Inspection Report# : 2005008(pdf)
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| Significance: Mar 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ESTABLISH ADEQUATE COMPENSATORY ACTIONS (HOURLY FIRE WATCH) ACCORDING TO FIRE PROTECTION PROGRAM PROCEDURES.
| |
| A finding of very low safety significance was identified by the inspectors on March 17, 2005, for a violation of license-required fire protection program requirements. The licensee failed to establish adequate hourly fire watches for a failed ionization detector as required by the approved fire protection program procedure. Following the inspectors' identification of this issue, the licensee established an hourly fire watch that met the requirements and recommendations of the licensee's approved fire protection program procedures.
| |
| This finding was more than minor because if left uncorrected, it could become a more significant safety concern. The licensee's ability to quickly detect a fire in the area was impaired due to an insufficient number of smoke detectors. The issue was of very low safety significance because the fire protection element impacted by the finding was still expected to provide some defense-in-depth benefit due to a second fire detector located in the room. Additionally, there were two nearby hose stations which could be used for fire suppression activities. The issue was a Non-Cited Violation of the facility operating license section 2.F which required the implementation of the fire protection program.
| |
| Inspection Report# : 2005003(pdf)
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| Mitigating Systems
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| 4Q/2005 Inspection Findings - Clinton Page 2 of 4 Significance: Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PROVIDE ADEQUATE MAINTENANCE AND WORK INSTRUCTION IS A PERFORMANCE DEFICIENCY A self-revealing finding involving a non-cited violation (NCV) of Technical Specification 5.4.1 "Procedures," was identified. On September 30, 2005, the Division III emergency diesel generator failed to properly run following maintenance activities, due to the inadequate maintenance instructions. The inadequate maintenance instructions resulted in air being trapped in the governor oil system during the replacement of the governor's servo booster motor. The licensee determined that this issue was the result of a maintenance planner's failure to follow administrative guidelines for technical review during the development of the maintenance instructions. This issue resulted in extended outage and unavailability time for the emergency diesel generator.
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| The inspectors determined that despite the fact that the issue involved work in progress, this issue was more than minor because the finding affected the Mitigating Systems Cornerstone objective of ensuring the availability of mitigating systems to prevent undesirable consequences.
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| The issue resulted in the emergency diesel generator being unavailable for longer than expected by the plant staff. Following the initial maintenance run of the diesel generator, operators declared that the diesel generator was available for use if needed to respond to an event.
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| Corrective actions by the licensee included developing lesson-learned information to share with other maintenance planners. Additionally, the licensee planned to add technical guidance related to venting air from the diesel governor to the diesel maintenance training material. The finding also affected the cross cutting area of human performance since the licensee's maintenance personnel failed to request technical guidance from the site engineering staff as directed by the licensee's administrative procedures.
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| Inspection Report# : 2005009(pdf)
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| Significance: Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO CORRECTLY IDENTIFY AND CORRECT THE CAUSE OF THE 2005 125 VDC CIRCUIT FAILURE WAS A PERFORMANCE DEFICIENCY.
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| The inspectors identified a finding involving a non-cited violation for inadequate corrective action. The licensee's failure to properly identify and correct a degraded electrical circuit in 2004, involving a high resistance connection on a fuse holder, resulted in the Division II emergency diesel generator subsystem being vulnerable to electrical circuit failure if called upon to complete its support function. The high resistance connection was caused by degraded grease-like material and dirt. This issue also resluted in the Division II diesel generator failure during a subsequent surveillance test.
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| The inspectors determined that the finding was greater than minor because the finding affected the Mitigating Systems Cornerstone objective of ensuring the availability, reliability, and sapability of mitigating systems to prevent undesirable consequences. The Division II emergency diesel generator 125 VDC system is a backup to the AC oil system in case of a loss of offsite power. Offsite power was not lost, therefore, there was not an actual loss of safety function for the diesel. Corrective actions by the licensee included replacing the fuse and fuse holder and expediting actions to address the extent of condition relative to the as-found condition of the fuse and fuse holder. The finding also affected the cross cutting area of problem identification and resolution since the licensee failed to adequately address the degraded circuit condition in a timely manner.
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| Inspection Report# : 2005009(pdf)
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| Significance: Jun 30, 2005 Identified By: NRC Item Type: FIN Finding IMPROPERLY SECURED 4160V EQUIPMENT DOORS In December 2003 the inspectors identified a discrepant condition on the 4160 volt Bus 1C1 Reserve Feed potential transformer cubicle. The inspectors considered this to be an inspection finding with no violations of NRC requirements identified.
| |
| The inspectors determined that the issue was more than minor because the finding could be reasonably viewed as a precursor to a significant event, which if left uncorrected, could render safety related equipment inoperable. The issue was a design/seismic qualification deficiency that was determined not to cause a loss of a safety related function by the licensee's evaluation. Based on this conclusion, this finding was determined to be of very low safety significance using the Phase 1 worksheets.
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| Inspection Report# : 2005007(pdf)
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| Significance: Sep 30, 2004 Identified By: NRC Item Type: FIN Finding DIVISION-3 ESSENTIAL SWITCHGEAR HEAT REMOVAL (VX) SYSTEM TRIPPED DUE TO INADEQUATE IMPACT STATEMENT FOR MAINTENANCE.
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| A finding of very low safety significance was self-revealed during a maintenance activity when Division essential switchgear heat removal was lost as a result of an inadequate impact statement in the work order. The primary cause of this finding was related to the cross-cutting area of Human Performance. In addition to the maintenance planner missing the relationship between the safety and non-safety supply fan motors,
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| 4Q/2005 Inspection Findings - Clinton Page 3 of 4 several other opportunities to identify this inadequate impact statement were missed.
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| This finding was more than minor because with the division three essential switchgear heat removal system unavailable, the high pressure core spray system may be rendered inoperable. The issue was of very low safety significance because the initial temperature in the division three switchgear room was low and the loss of essential switchgear heat removal was of short duration, the high pressure core spray system was never actually inoperable. No violation of NRC requirements occurred.
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| Inspection Report# : 2004006(pdf)
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| Barrier Integrity Significance: Mar 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW PROCEDURE AND APPROPRIATELY ANNOTATE PORTIONS AS NOT APPLICABLE DURING THE PERFORMANCE OF REQUIRED CALIBRATION PROCEDURE IN ACCORDANCE WITH TS 5.4.1.
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| Through a self-revealing event (unexpected de-energized relay found during maintenance) the inspectors identified a Non-Cited Violation (NCV) of very low safety signficiance. This finding resulted from licensee personnel incorrectly designating procedureal steps as not applicable during the performance of a calibration procedure, Clinton Power Station (CPS) 9432.60, "Channel Functional Test for Containment Building Exhaust Radiation Monitor," required by Technical Specifications. In Issue Report (IR) 289643, the licensee documented that with the realy de-energized the affected primary containment isolation valve cannot by opened without taking the corresponding Division 2 LOCA BYPASS switch to the BYPASS position (an action administratively controlled by Operations).
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| The inspectors determined that the finding was greater than minor because this issue could be reasonably viewed as a precursor to a more significant event. Additionally, this finding was associated with the Barrier Integrity Cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radioactive releases caused by accidents or events. The finding was of very low safety significance because this issue did not cause an actual open pathway in the physical integrity of reactor containment. The licensee documented the issue in IR 289643 and generated corrective actions as the result of a human performance investigation report being performed. These corrective actions included revising CPS 9432.60 to clearly identify the reason for placing the switch to BYPASS.
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| Inspection Report# : 2005003(pdf)
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| Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Significance: Mar 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO MAINTAIN CONTROL OF LICENSED RADIOACTIVE MATERIAL IN ACCORDANCE WITH 10 CFR 20, SUBPART 1.
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| A finding of very low safety significance and an associated Non-Cited Violation were identified through a self-revealing event on October 7, 2004, when licensee personnel discovered that three nuclear instrument detectors (containing a very small amount of radioactive material) were not adequately controlled. Licensee personnel believed that the material was contained in a small container which was sealed in 1991 as part of a disposition plan for the defective instruments. The licensee' search of other material containers and documentation failed to identify the final disposition of the radioactive material.
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| The issue was more than minor because it was associated with the Human Performance and Programs/Process attributes of the Public Radiation Safety Cornerstone and affected the cornerstone objective of ensuring adequate protection of public health and safety from exposure to radioactive materials potentially released into the public domain. Based on various dose calculation scenarios, the very small amount of missing radioactive material would contribute a negligible radiological dose if a member of the public were to be exposed to the material.
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| 4Q/2005 Inspection Findings - Clinton Page 4 of 4 Additionally, the inspectors determined that the licensee did not have any prior radioactive material control occurrences in the previous 8 quarters. Therefore, the finding was of very low safety significance. The licensee's corrective actions for this issue included the development of procedural guidance which prohibits removing nuclear instrument detectors from the cabling as part of a disposition plan for defective units.
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| One Non-Cited Vilation for the failure to control licensed radioactive material in accordance with 10 CFR 20, Subpart 1, was identified.
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| Inspection Report# : 2005003(pdf)
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| Physical Protection Physical Protection information not publicly available.
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| Miscellaneous Last modified : March 03, 2006
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| 1Q/2006 Inspection Findings - Clinton Page 1 of 4 Clinton 1Q/2006 Plant Inspection Findings Initiating Events Significance: Sep 30, 2005 Identified By: NRC Item Type: FIN Finding PERFORMANCE OF WORK IN THE OFF-GAS SYSTEM THAT RESULTED IN A SUBSEQUENT LOSS IN OFF-GAS SYSTEM FLOW AND THE OPERATORS PERFORMING A RAPID POWER REDUCTION.
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| On August 29, 2005, a finding of very low safety significance was self revealed following the performance of work in the off-gas system that resulted in a subsequent loss in off-gas system flow and the operators performing a rapid power reduction. The finding involved the failure to stroke a gas dryer inlet valve to ensure the valve would operate following a packing adjustment. This issue was caused by poor work practices and communication by licensee personnel.
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| The issue was more than minor because it affected the Reactor Safety/Initiating Event cornerstone objective of limiting the likelihood of those events that upset plant stability. The finding was of very low safety significance because it would not affect the availability of mitigating systems or functions even if it had resulted in a plant trip. No violation of NRC requirements occurred. The finding also affected the cross cutting area of Human Performance.
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| Inspection Report# : 2005008(pdf)
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| Significance: Sep 30, 2005 Identified By: NRC Item Type: FIN Finding THE LICENSEE FAILED TO TAKE PROMPT ACTION TO CORRECT A PROBLEM WITHIN THE ELECTRO-HYDRAULIC CONTROL SYSTEM.
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| On July 17, 2005, a finding of very low safety significance was identified by the inspectors when the licensee failed to take prompt action to correct a problem within the electro-hydraulic control system. In April 2005, one main turbine combined intermediate valve went shut at power due to a clogged servo valve strainer, causing a plant transient. The licensee identified that other main turbine valves were susceptible to the same failure, but did not take action to correct the problem until after a second combined intermediate valve went shut three months later, causing a second plant transient.
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| The issue was more than minor becuse the licensee knew of the degraded condition and associated risks and failed to correct the problem before it resulted in a second plant transient requiring operators to respond. The finding was of very low safety significance because it would not affect the availability of mitigating systems or functions even if it had resulted in a plant trip. No violation of NRC requirements occurred.
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| Inspection Report# : 2005008(pdf)
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| Mitigating Systems Significance: Mar 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE DESIGN CONTROL DURING REVIEW OF ENGINEERING CHANGE PACKAGE 356820 "SHUTDOWN COOLING HEADER LEAK-OFF LINE".
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| In February 2006, a finding of very low safety significance involving a Non-Cited Violation of 10 CFR 50, Appendix B, Criteria III, "Design Control," was identified. During a review of Engineering Change Package 356820, "Shutdown Cooling Header Leak-off line," the inspectors identified that the design change, as installed, would adversely impact the functionality of both the Division 2 residual heat removal system's water leg (keep-fill) pump and the C residual heat removal pump. This adverse condition would be caused by the introduction of high temperature water on the suction side of both pumps. The design change was being installed to prevent pressurization of the shutdown cooling header due to leakage through the reactor coolant system pressure isolation valves.
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| This issue was more than minor because the finding affected the Mitigating Systems cornerstone objective of ensuring the availability of mitigating systems to prevent undesirable consequences (Design Control attributes). The finding was of very low safety significance because, with the expected operator actions, this condition would not result in a loss of operability. This conclusion was made based on the flow limiting characteristics of the leak-off line orifice with the suction cooling header volume at saturated conditions in conjunction with the subsequent
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| 1Q/2006 Inspection Findings - Clinton Page 2 of 4 operator alarm response requirements. Corrective actions by the licensee included procedure revisions and local monitoring of the C residual heat removal suction line temperature once the leak-off line was placed in service.
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| Inspection Report# : 2006002(pdf)
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| Significance: Mar 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE TEST CONTROL DURING THE REVIEW OF THE LICENSEE'S SURVEILLANCE TEST TO DETERMINE OPERABILITY OF THE SHUTDOWN SERVICE WATER SYSTEM.
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| On February 2, 2006, the inspectors identified a finding involving a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XI, "Test Controls." During a review of the licensee's surveillance test to determine the operability of the shutdown service water system, the inspectors identified that the system's leakage could exceed both the administrative and operability limits established by design basis documents, without the test detecting the actual leak rate. This condition was caused by an inadequate test connection.
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| This issue was more than minor because the finding affected the Mitigating Systems cornerstone objective of ensuring the availability of mitigating systems to prevent undesirable consequences. An adverse condition would have been masked by leakage that exceeded both administrative and operability limits, and would not have been identified under testing conditions mandated by the licensee's testing program.
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| the finding was of very low safety significance because the actual measured leakage was well below the capability of accurately being measured, and this issue did not result in a system operability concern. As part of the corrective actions, the licensee planned to performed an extent of condition review to ensure that no other system leakage tests were affected by this issue.
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| Inspection Report# : 2006002(pdf)
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| Significance: Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PROVIDE ADEQUATE MAINTENANCE AND WORK INSTRUCTION IS A PERFORMANCE DEFICIENCY A self-revealing finding involving a non-cited violation (NCV) of Technical Specification 5.4.1 "Procedures," was identified. On September 30, 2005, the Division III emergency diesel generator failed to properly run following maintenance activities, due to the inadequate maintenance instructions. The inadequate maintenance instructions resulted in air being trapped in the governor oil system during the replacement of the governor's servo booster motor. The licensee determined that this issue was the result of a maintenance planner's failure to follow administrative guidelines for technical review during the development of the maintenance instructions. This issue resulted in extended outage and unavailability time for the emergency diesel generator.
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| The inspectors determined that despite the fact that the issue involved work in progress, this issue was more than minor because the finding affected the Mitigating Systems Cornerstone objective of ensuring the availability of mitigating systems to prevent undesirable consequences.
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| The issue resulted in the emergency diesel generator being unavailable for longer than expected by the plant staff. Following the initial maintenance run of the diesel generator, operators declared that the diesel generator was available for use if needed to respond to an event.
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| Corrective actions by the licensee included developing lesson-learned information to share with other maintenance planners. Additionally, the licensee planned to add technical guidance related to venting air from the diesel governor to the diesel maintenance training material. The finding also affected the cross cutting area of human performance since the licensee's maintenance personnel failed to request technical guidance from the site engineering staff as directed by the licensee's administrative procedures.
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| Inspection Report# : 2005009(pdf)
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| Significance: Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO CORRECTLY IDENTIFY AND CORRECT THE CAUSE OF THE 2005 125 VDC CIRCUIT FAILURE WAS A PERFORMANCE DEFICIENCY.
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| The inspectors identified a finding involving a non-cited violation for inadequate corrective action. The licensee's failure to properly identify and correct a degraded electrical circuit in 2004, involving a high resistance connection on a fuse holder, resulted in the Division II emergency diesel generator subsystem being vulnerable to electrical circuit failure if called upon to complete its support function. The high resistance connection was caused by degraded grease-like material and dirt. This issue also resluted in the Division II diesel generator failure during a subsequent surveillance test.
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| The inspectors determined that the finding was greater than minor because the finding affected the Mitigating Systems Cornerstone objective of ensuring the availability, reliability, and sapability of mitigating systems to prevent undesirable consequences. The Division II emergency diesel generator 125 VDC system is a backup to the AC oil system in case of a loss of offsite power. Offsite power was not lost, therefore, there was not an actual loss of safety function for the diesel. Corrective actions by the licensee included replacing the fuse and fuse holder and expediting actions to address the extent of condition relative to the as-found condition of the fuse and fuse holder. The finding also affected the cross cutting area of problem identification and resolution since the licensee failed to adequately address the degraded circuit condition in a timely manner.
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| Inspection Report# : 2005009(pdf)
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| 1Q/2006 Inspection Findings - Clinton Page 3 of 4 Significance: Dec 02, 2005 Identified By: NRC Item Type: NCV NonCited Violation NON-CONSERVATIVE ACCEPTANCE CRITERIA A finding of very low safety significance was identified by the inspectors for a violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control" requirements. Specifically, the licensee failed to incorporate the most restrictive hydraulic conditions into the calculation which established the acceptance criteria for a technical specification surveillance test. This resulted in a HPCS system hydraulic calculation that was non-conservative when determining the pumps minimum acceptance criteria. Once identified, the licensee evaluated operability and entered the finding into their corrective action program to revise the affected documents.
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| The finding was more than minor because the failure to account for all modes of HPCS system operation in the surveillance tests acceptance criteria could result in unacceptable degradation and could have affected the mitigating systems cornerstone objective. The finding was of very low safety significance because the licensees analysis showed that adequate design margin existed for the HPCS system and did not represent an actual loss of a safety function.
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| Inspection Report# : 2005002(pdf)
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| Significance: Dec 02, 2005 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE HEAT EXCHANGER THERMAL PERFORMANCE TESTING A finding of very low safety significance was identified by the inspectors for a violation of 10 CFR Part 50, Appendix B, Criterion XI, "Test Control" requirements. Specifically, in 2000, 2002 and 2003, the licensee failed to recognize that the calculated value for the diesel generator (DG) jacket-water (JW) flow rate, as determined from test data obtained during thermal performance testing of the Division III DG JW cooler heat exchanger (HX), was significantly higher than the flow rate that could be attained by the engine-driven water pump. Once identified, the licensee entered the finding into their corrective action program as Condition Report (CR) 426459, NRC SSD&PC Is the Calculated Process Flow Rate Reasonable, dated November 21, 2005, and CR429726, Discrepancies Not Identified in Corrective Action Process, dated December 2, 2005, to evaluate and/or revise the affected test procedures.
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| The finding was more than minor because the failure to account for flow rates that were significantly greater than that identified by the equipments design specification produced equipment performance data that did not accurately demonstrate the HXs availability and reliability. The finding was of very low safety significance because the licensees evaluation showed that the Division III DGs JW Cooler HX would have performed its safety function and did not represent an actual loss of a safety function. A contributing cause of the finding was related to the cross-cutting element of problem identification and resolution. Specifically, a similar issue was identified during another NRC inspection in 2001; however, the licensee did not properly evaluate and take actions. As a result, testing done in 2002 and 2003 showed the same discrepant flow rates.
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| Inspection Report# : 2005002(pdf)
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| Significance: Jun 30, 2005 Identified By: NRC Item Type: FIN Finding IMPROPERLY SECURED 4160V EQUIPMENT DOORS In December 2003 the inspectors identified a discrepant condition on the 4160 volt Bus 1C1 Reserve Feed potential transformer cubicle. The inspectors considered this to be an inspection finding with no violations of NRC requirements identified.
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| The inspectors determined that the issue was more than minor because the finding could be reasonably viewed as a precursor to a significant event, which if left uncorrected, could render safety related equipment inoperable. The issue was a design/seismic qualification deficiency that was determined not to cause a loss of a safety related function by the licensee's evaluation. Based on this conclusion, this finding was determined to be of very low safety significance using the Phase 1 worksheets.
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| Inspection Report# : 2005007(pdf)
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| Significance: Sep 30, 2004 Identified By: NRC Item Type: FIN Finding DIVISION-3 ESSENTIAL SWITCHGEAR HEAT REMOVAL (VX) SYSTEM TRIPPED DUE TO INADEQUATE IMPACT STATEMENT FOR MAINTENANCE.
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| A finding of very low safety significance was self-revealed during a maintenance activity when Division essential switchgear heat removal was lost as a result of an inadequate impact statement in the work order. The primary cause of this finding was related to the cross-cutting area of Human Performance. In addition to the maintenance planner missing the relationship between the safety and non-safety supply fan motors, several other opportunities to identify this inadequate impact statement were missed.
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| This finding was more than minor because with the division three essential switchgear heat removal system unavailable, the high pressure core spray system may be rendered inoperable. The issue was of very low safety significance because the initial temperature in the division three
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| 1Q/2006 Inspection Findings - Clinton Page 4 of 4 switchgear room was low and the loss of essential switchgear heat removal was of short duration, the high pressure core spray system was never actually inoperable. No violation of NRC requirements occurred.
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| Inspection Report# : 2004006(pdf)
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| Barrier Integrity Emergency Preparedness Occupational Radiation Safety Significance: Mar 31, 2006 Identified By: NRC Item Type: FIN Finding FAILURE TO MAINTAIN COLLECTIVE RADIATION DOSE TO OCCUPATIONAL WORKERS INVOLVED IN REFUEL FLOOR WORK ALARA.
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| An inspector-identified finding of very low safety significance was identified for the failure to maintain the collective dose As-Low-As-Is-Reasonably-Achievable (ALARA) for refuel floor non-cavity work that was conducted during the February 2006 refueling outage. The additional, unintended dose was attributable to deficiencies in both work planning and work execution. The actual collective dose for this work activity was approximately 14 person-rem compared to the licensee's initial dose estimate of 4.4 person-rem. A revised dose estimate of about 7 person-rem was determined by the inspectors based on reasonably unexpected changes in radiological conditions and equipment problems.
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| Consequently, the collective dose for this work exceeded 5 rem and exceeded the revised dose projection by more than 50 percent.
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| The issue was more than minor because it was associated with the Program/Process (ALARA planning) attribute of the Occupational Radiation Safety cornerstone and affected the cornerstone objective to ensure adequate protection of worker health and safety from exposure to radiation.
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| This issue represents a finding of very low safety significance because it involved ALARA planning; however, the Clinton plant's current 3-year rolling average collective dose does not exceed 240 person-rem. The licensee entered this radiological work planning/dose performance problem into its outage lessons learned database to allow the development of measures to better plan and execute refuel floor work during future refueling outages.
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| Inspection Report# : 2006002(pdf)
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| Public Radiation Safety Physical Protection Physical Protection information not publicly available.
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| Miscellaneous Last modified : May 25, 2006
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| 2Q/2006 Inspection Findings - Clinton Page 1 of 4 Clinton 2Q/2006 Plant Inspection Findings Initiating Events Significance: Mar 20, 2006 Identified By: Self-Revealing Item Type: FIN Finding Inadequate workmanship resulted in generator trip and reactor scram.
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| The inspectors considered the failure to adequately tighten terminal screws in the main generator output current transformer circuit a performance deficiency. This issue was caused by inadequate workmanship. The inspectors determined it was more than minor because the finding affected the reactor safety/initiating events cornerstone objective of limiting the likelihood of those events that upset plant stability. The finding also affected the cross-cutting area of human performance because the contract workers failed to tighten the terminal screws of the current transformer and the licensee failed to ensure the GE workers were using the appropriate lifted and landed leads documents to aid in performance of this job. Although this failure occurred in C1R08 in April, 2002, the inspectors determined this deficiency to be reflective of recent licensee performance because, up to the March 2006 scram event, there was no procedure or process in place to ensure GE followed the licensees lifted and landed lead procedures.
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| As a result of the root cause for this event, the licensee initiated a corrective action to revise the GE quality control check-list to confirm that requirements similar to wire removal/jumper installation procedures are incorporated. Although this finding did contribute to the likelihood of a reactor trip, it did not affect the function or availability of any mitigation equipment. Therefore, the inspectors concluded that this issue was a finding of very low safety significance (Green).
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| Inspection Report# : 2006004(pdf)
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| Significance: Sep 30, 2005 Identified By: NRC Item Type: FIN Finding PERFORMANCE OF WORK IN THE OFF-GAS SYSTEM THAT RESULTED IN A SUBSEQUENT LOSS IN OFF-GAS SYSTEM FLOW AND THE OPERATORS PERFORMING A RAPID POWER REDUCTION.
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| On August 29, 2005, a finding of very low safety significance was self revealed following the performance of work in the off-gas system that resulted in a subsequent loss in off-gas system flow and the operators performing a rapid power reduction. The finding involved the failure to stroke a gas dryer inlet valve to ensure the valve would operate following a packing adjustment. This issue was caused by poor work practices and communication by licensee personnel.
| |
| The issue was more than minor because it affected the Reactor Safety/Initiating Event cornerstone objective of limiting the likelihood of those events that upset plant stability. The finding was of very low safety significance because it would not affect the availability of mitigating systems or functions even if it had resulted in a plant trip. No violation of NRC requirements occurred. The finding also affected the cross cutting area of Human Performance.
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| Inspection Report# : 2005008(pdf)
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| Significance: Sep 30, 2005 Identified By: NRC Item Type: FIN Finding THE LICENSEE FAILED TO TAKE PROMPT ACTION TO CORRECT A PROBLEM WITHIN THE ELECTRO-HYDRAULIC CONTROL SYSTEM.
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| On July 17, 2005, a finding of very low safety significance was identified by the inspectors when the licensee failed to take prompt action to correct a problem within the electro-hydraulic control system. In April 2005, one main turbine combined intermediate valve went shut at power due to a clogged servo valve strainer, causing a plant transient. The licensee identified that other main turbine valves were susceptible to the same failure, but did not take action to correct the problem until after a second combined intermediate valve went shut three months later, causing a second plant transient.
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| The issue was more than minor becuse the licensee knew of the degraded condition and associated risks and failed to correct the problem before it resulted in a second plant transient requiring operators to respond. The finding was of very low safety significance because it would not affect the availability of mitigating systems or functions even if it had resulted in a plant trip. No violation of NRC requirements occurred.
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| Inspection Report# : 2005008(pdf)
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| Mitigating Systems
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| 2Q/2006 Inspection Findings - Clinton Page 2 of 4 Significance: Mar 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE DESIGN CONTROL DURING REVIEW OF ENGINEERING CHANGE PACKAGE 356820 "SHUTDOWN COOLING HEADER LEAK-OFF LINE".
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| In February 2006, a finding of very low safety significance involving a Non-Cited Violation of 10 CFR 50, Appendix B, Criteria III, "Design Control," was identified. During a review of Engineering Change Package 356820, "Shutdown Cooling Header Leak-off line," the inspectors identified that the design change, as installed, would adversely impact the functionality of both the Division 2 residual heat removal system's water leg (keep-fill) pump and the C residual heat removal pump. This adverse condition would be caused by the introduction of high temperature water on the suction side of both pumps. The design change was being installed to prevent pressurization of the shutdown cooling header due to leakage through the reactor coolant system pressure isolation valves.
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| This issue was more than minor because the finding affected the Mitigating Systems cornerstone objective of ensuring the availability of mitigating systems to prevent undesirable consequences (Design Control attributes). The finding was of very low safety significance because, with the expected operator actions, this condition would not result in a loss of operability. This conclusion was made based on the flow limiting characteristics of the leak-off line orifice with the suction cooling header volume at saturated conditions in conjunction with the subsequent operator alarm response requirements. Corrective actions by the licensee included procedure revisions and local monitoring of the C residual heat removal suction line temperature once the leak-off line was placed in service.
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| Inspection Report# : 2006002(pdf)
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| Significance: Mar 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE TEST CONTROL DURING THE REVIEW OF THE LICENSEE'S SURVEILLANCE TEST TO DETERMINE OPERABILITY OF THE SHUTDOWN SERVICE WATER SYSTEM.
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| On February 2, 2006, the inspectors identified a finding involving a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XI, "Test Controls." During a review of the licensee's surveillance test to determine the operability of the shutdown service water system, the inspectors identified that the system's leakage could exceed both the administrative and operability limits established by design basis documents, without the test detecting the actual leak rate. This condition was caused by an inadequate test connection.
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| This issue was more than minor because the finding affected the Mitigating Systems cornerstone objective of ensuring the availability of mitigating systems to prevent undesirable consequences. An adverse condition would have been masked by leakage that exceeded both administrative and operability limits, and would not have been identified under testing conditions mandated by the licensee's testing program. the finding was of very low safety significance because the actual measured leakage was well below the capability of accurately being measured, and this issue did not result in a system operability concern. As part of the corrective actions, the licensee planned to performed an extent of condition review to ensure that no other system leakage tests were affected by this issue.
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| Inspection Report# : 2006002(pdf)
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| Significance: Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PROVIDE ADEQUATE MAINTENANCE AND WORK INSTRUCTION IS A PERFORMANCE DEFICIENCY A self-revealing finding involving a non-cited violation (NCV) of Technical Specification 5.4.1 "Procedures," was identified. On September 30, 2005, the Division III emergency diesel generator failed to properly run following maintenance activities, due to the inadequate maintenance instructions. The inadequate maintenance instructions resulted in air being trapped in the governor oil system during the replacement of the governor's servo booster motor. The licensee determined that this issue was the result of a maintenance planner's failure to follow administrative guidelines for technical review during the development of the maintenance instructions. This issue resulted in extended outage and unavailability time for the emergency diesel generator.
| |
| The inspectors determined that despite the fact that the issue involved work in progress, this issue was more than minor because the finding affected the Mitigating Systems Cornerstone objective of ensuring the availability of mitigating systems to prevent undesirable consequences. The issue resulted in the emergency diesel generator being unavailable for longer than expected by the plant staff. Following the initial maintenance run of the diesel generator, operators declared that the diesel generator was available for use if needed to respond to an event. Corrective actions by the licensee included developing lesson-learned information to share with other maintenance planners. Additionally, the licensee planned to add technical guidance related to venting air from the diesel governor to the diesel maintenance training material. The finding also affected the cross cutting area of human performance since the licensee's maintenance personnel failed to request technical guidance from the site engineering staff as directed by the licensee's administrative procedures.
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| Inspection Report# : 2005009(pdf)
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| Significance: Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO CORRECTLY IDENTIFY AND CORRECT THE CAUSE OF THE 2005 125 VDC CIRCUIT FAILURE WAS A
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| 2Q/2006 Inspection Findings - Clinton Page 3 of 4 PERFORMANCE DEFICIENCY.
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| The inspectors identified a finding involving a non-cited violation for inadequate corrective action. The licensee's failure to properly identify and correct a degraded electrical circuit in 2004, involving a high resistance connection on a fuse holder, resulted in the Division II emergency diesel generator subsystem being vulnerable to electrical circuit failure if called upon to complete its support function. The high resistance connection was caused by degraded grease-like material and dirt. This issue also resluted in the Division II diesel generator failure during a subsequent surveillance test.
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| The inspectors determined that the finding was greater than minor because the finding affected the Mitigating Systems Cornerstone objective of ensuring the availability, reliability, and sapability of mitigating systems to prevent undesirable consequences. The Division II emergency diesel generator 125 VDC system is a backup to the AC oil system in case of a loss of offsite power. Offsite power was not lost, therefore, there was not an actual loss of safety function for the diesel. Corrective actions by the licensee included replacing the fuse and fuse holder and expediting actions to address the extent of condition relative to the as-found condition of the fuse and fuse holder. The finding also affected the cross cutting area of problem identification and resolution since the licensee failed to adequately address the degraded circuit condition in a timely manner.
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| Inspection Report# : 2005009(pdf)
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| Significance: Dec 02, 2005 Identified By: NRC Item Type: NCV NonCited Violation NON-CONSERVATIVE ACCEPTANCE CRITERIA A finding of very low safety significance was identified by the inspectors for a violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control" requirements. Specifically, the licensee failed to incorporate the most restrictive hydraulic conditions into the calculation which established the acceptance criteria for a technical specification surveillance test. This resulted in a HPCS system hydraulic calculation that was non-conservative when determining the pumps minimum acceptance criteria. Once identified, the licensee evaluated operability and entered the finding into their corrective action program to revise the affected documents.
| |
| The finding was more than minor because the failure to account for all modes of HPCS system operation in the surveillance tests acceptance criteria could result in unacceptable degradation and could have affected the mitigating systems cornerstone objective. The finding was of very low safety significance because the licensees analysis showed that adequate design margin existed for the HPCS system and did not represent an actual loss of a safety function.
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| Inspection Report# : 2005002(pdf)
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| Significance: Dec 02, 2005 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE HEAT EXCHANGER THERMAL PERFORMANCE TESTING A finding of very low safety significance was identified by the inspectors for a violation of 10 CFR Part 50, Appendix B, Criterion XI, "Test Control" requirements. Specifically, in 2000, 2002 and 2003, the licensee failed to recognize that the calculated value for the diesel generator (DG) jacket-water (JW) flow rate, as determined from test data obtained during thermal performance testing of the Division III DG JW cooler heat exchanger (HX), was significantly higher than the flow rate that could be attained by the engine-driven water pump. Once identified, the licensee entered the finding into their corrective action program as Condition Report (CR) 426459, NRC SSD&PC Is the Calculated Process Flow Rate Reasonable, dated November 21, 2005, and CR429726, Discrepancies Not Identified in Corrective Action Process, dated December 2, 2005, to evaluate and/or revise the affected test procedures.
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| The finding was more than minor because the failure to account for flow rates that were significantly greater than that identified by the equipments design specification produced equipment performance data that did not accurately demonstrate the HXs availability and reliability. The finding was of very low safety significance because the licensees evaluation showed that the Division III DGs JW Cooler HX would have performed its safety function and did not represent an actual loss of a safety function. A contributing cause of the finding was related to the cross-cutting element of problem identification and resolution. Specifically, a similar issue was identified during another NRC inspection in 2001; however, the licensee did not properly evaluate and take actions. As a result, testing done in 2002 and 2003 showed the same discrepant flow rates.
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| Inspection Report# : 2005002(pdf)
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| Significance: Sep 30, 2004 Identified By: NRC Item Type: FIN Finding DIVISION-3 ESSENTIAL SWITCHGEAR HEAT REMOVAL (VX) SYSTEM TRIPPED DUE TO INADEQUATE IMPACT STATEMENT FOR MAINTENANCE.
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| A finding of very low safety significance was self-revealed during a maintenance activity when Division essential switchgear heat removal was lost as a result of an inadequate impact statement in the work order. The primary cause of this finding was related to the cross-cutting area of Human Performance. In addition to the maintenance planner missing the relationship between the safety and non-safety supply fan motors, several other opportunities to identify this inadequate impact statement were missed.
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| This finding was more than minor because with the division three essential switchgear heat removal system unavailable, the high pressure core spray system may be rendered inoperable. The issue was of very low safety significance because the initial temperature in the division three switchgear room was low and the loss of essential switchgear heat removal was of short duration, the high pressure core spray system was never
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| 2Q/2006 Inspection Findings - Clinton Page 4 of 4 actually inoperable. No violation of NRC requirements occurred.
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| Inspection Report# : 2004006(pdf)
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| Barrier Integrity Emergency Preparedness Occupational Radiation Safety Significance: Mar 31, 2006 Identified By: NRC Item Type: FIN Finding FAILURE TO MAINTAIN COLLECTIVE RADIATION DOSE TO OCCUPATIONAL WORKERS INVOLVED IN REFUEL FLOOR WORK ALARA.
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| An inspector-identified finding of very low safety significance was identified for the failure to maintain the collective dose As-Low-As-Is-Reasonably-Achievable (ALARA) for refuel floor non-cavity work that was conducted during the February 2006 refueling outage. The additional, unintended dose was attributable to deficiencies in both work planning and work execution. The actual collective dose for this work activity was approximately 14 person-rem compared to the licensee's initial dose estimate of 4.4 person-rem. A revised dose estimate of about 7 person-rem was determined by the inspectors based on reasonably unexpected changes in radiological conditions and equipment problems. Consequently, the collective dose for this work exceeded 5 rem and exceeded the revised dose projection by more than 50 percent.
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| The issue was more than minor because it was associated with the Program/Process (ALARA planning) attribute of the Occupational Radiation Safety cornerstone and affected the cornerstone objective to ensure adequate protection of worker health and safety from exposure to radiation. This issue represents a finding of very low safety significance because it involved ALARA planning; however, the Clinton plant's current 3-year rolling average collective dose does not exceed 240 person-rem. The licensee entered this radiological work planning/dose performance problem into its outage lessons learned database to allow the development of measures to better plan and execute refuel floor work during future refueling outages.
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| Inspection Report# : 2006002(pdf)
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| Public Radiation Safety Physical Protection Physical Protection information not publicly available.
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| Miscellaneous Last modified : August 25, 2006
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| 3Q/2006 Inspection Findings - Clinton Page 1 of 5 Clinton 3Q/2006 Plant Inspection Findings Initiating Events Significance: Sep 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation THE INSPECTORS DETERMINED THAT THE FAILURE TO APPROPRIATELY IDENTIFY AND CORRECT THE CAUSE OF THE DIVISION 4 NSPS INVERTER IN MARCH WAS A PERFROMANCE DEFICIENCY.
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| A finding of very low safety significance and a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, was self revealed following a reactor scram on August 27, 2006, due to the licensee's failure to identify and correct a condition adverse to quality in March 2006. The licensee determined and corrected the actual cause of the failure and revised procurement procedures to disallow purchase of parts manufactured under the same process as the failed board.
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| Additionally, the licensee commenced a common cause evaluation to assist in planning and developing additional corrective actions to address whether there are issues involving the licensee proficiency in identifying causes of operational occurrences.
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| The finding was more than minor because it resulted in a reactor scram and was associated with the equipment performance attribute of the initiating events cornerstone. The finding was of very low safety significance because it would not affect the availability of a mitigating system. The finding was also determined to affect the cross-cutting area of problem identification and resolution in that the actual cause of the March 26, 2006 failure was not properly identified, resulting in the corrective action not addressing the cause, and a more significant failure occurring in August 2006.
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| Inspection Report# : 2006007(pdf)
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| Significance: Mar 20, 2006 Identified By: Self-Revealing Item Type: FIN Finding INADEQUATE WORKMANSHIP RESULTED IN GENERATOR TRIP AND REACTOR SCRAM.
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| The inspectors considered the failure to adequately tighten terminal screws in the main generator output current transformer circuit a performance deficiency. This issue was caused by inadequate workmanship. The inspectors determined it was more than minor because the finding affected the reactor safety/initiating events cornerstone objective of limiting the likelihood of those events that upset plant stability. The finding also affected the cross-cutting area of human performance because the contract workers failed to tighten the terminal screws of the current transformer and the licensee failed to ensure the GE workers were using the appropriate lifted and landed leads documents to aid in performance of this job.
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| Although this failure occurred in C1R08 in April, 2002, the inspectors determined this deficiency to be reflective of recent licensee performance because, up to the March 2006 scram event, there was no procedure or process in place to ensure GE followed the licensees lifted and landed lead procedures. As a result of the root cause for this event, the licensee initiated a corrective action to revise the GE quality control check-list to confirm that requirements similar to wire removal/jumper installation procedures are incorporated. Although this finding did contribute to the likelihood of a reactor trip, it did not affect the function or availability of any mitigation equipment. Therefore, the inspectors concluded that this issue was a finding of very low safety significance (Green).
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| Inspection Report# : 2006004(pdf)
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| Mitigating Systems Significance: Mar 31, 2006 Identified By: NRC
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| 3Q/2006 Inspection Findings - Clinton Page 2 of 5 Item Type: NCV NonCited Violation INADEQUATE DESIGN CONTROL DURING REVIEW OF ENGINEERING CHANGE PACKAGE 356820 "SHUTDOWN COOLING HEADER LEAK-OFF LINE".
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| In February 2006, a finding of very low safety significance involving a Non-Cited Violation of 10 CFR 50, Appendix B, Criteria III, "Design Control," was identified. During a review of Engineering Change Package 356820, "Shutdown Cooling Header Leak-off line," the inspectors identified that the design change, as installed, would adversely impact the functionality of both the Division 2 residual heat removal system's water leg (keep-fill) pump and the C residual heat removal pump. This adverse condition would be caused by the introduction of high temperature water on the suction side of both pumps. The design change was being installed to prevent pressurization of the shutdown cooling header due to leakage through the reactor coolant system pressure isolation valves.
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| This issue was more than minor because the finding affected the Mitigating Systems cornerstone objective of ensuring the availability of mitigating systems to prevent undesirable consequences (Design Control attributes). The finding was of very low safety significance because, with the expected operator actions, this condition would not result in a loss of operability.
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| This conclusion was made based on the flow limiting characteristics of the leak-off line orifice with the suction cooling header volume at saturated conditions in conjunction with the subsequent operator alarm response requirements. Corrective actions by the licensee included procedure revisions and local monitoring of the C residual heat removal suction line temperature once the leak-off line was placed in service.
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| Inspection Report# : 2006002(pdf)
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| Significance: Mar 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE TEST CONTROL DURING THE REVIEW OF THE LICENSEE'S SURVEILLANCE TEST TO DETERMINE OPERABILITY OF THE SHUTDOWN SERVICE WATER SYSTEM.
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| On February 2, 2006, the inspectors identified a finding involving a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XI, "Test Controls." During a review of the licensee's surveillance test to determine the operability of the shutdown service water system, the inspectors identified that the system's leakage could exceed both the administrative and operability limits established by design basis documents, without the test detecting the actual leak rate. This condition was caused by an inadequate test connection.
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| This issue was more than minor because the finding affected the Mitigating Systems cornerstone objective of ensuring the availability of mitigating systems to prevent undesirable consequences. An adverse condition would have been masked by leakage that exceeded both administrative and operability limits, and would not have been identified under testing conditions mandated by the licensee's testing program. the finding was of very low safety significance because the actual measured leakage was well below the capability of accurately being measured, and this issue did not result in a system operability concern. As part of the corrective actions, the licensee planned to performed an extent of condition review to ensure that no other system leakage tests were affected by this issue.
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| Inspection Report# : 2006002(pdf)
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| Significance: Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PROVIDE ADEQUATE MAINTENANCE AND WORK INSTRUCTION IS A PERFORMANCE DEFICIENCY A self-revealing finding involving a non-cited violation (NCV) of Technical Specification 5.4.1 "Procedures," was identified. On September 30, 2005, the Division III emergency diesel generator failed to properly run following maintenance activities, due to the inadequate maintenance instructions. The inadequate maintenance instructions resulted in air being trapped in the governor oil system during the replacement of the governor's servo booster motor. The licensee determined that this issue was the result of a maintenance planner's failure to follow administrative guidelines for technical review during the development of the maintenance instructions. This issue resulted in extended outage and unavailability time for the emergency diesel generator.
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| The inspectors determined that despite the fact that the issue involved work in progress, this issue was more than minor because the finding affected the Mitigating Systems Cornerstone objective of ensuring the availability of mitigating
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| 3Q/2006 Inspection Findings - Clinton Page 3 of 5 systems to prevent undesirable consequences. The issue resulted in the emergency diesel generator being unavailable for longer than expected by the plant staff. Following the initial maintenance run of the diesel generator, operators declared that the diesel generator was available for use if needed to respond to an event. Corrective actions by the licensee included developing lesson-learned information to share with other maintenance planners. Additionally, the licensee planned to add technical guidance related to venting air from the diesel governor to the diesel maintenance training material. The finding also affected the cross cutting area of human performance since the licensee's maintenance personnel failed to request technical guidance from the site engineering staff as directed by the licensee's administrative procedures.
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| Inspection Report# : 2005009(pdf)
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| Significance: Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO CORRECTLY IDENTIFY AND CORRECT THE CAUSE OF THE 2005 125 VDC CIRCUIT FAILURE WAS A PERFORMANCE DEFICIENCY.
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| The inspectors identified a finding involving a non-cited violation for inadequate corrective action. The licensee's failure to properly identify and correct a degraded electrical circuit in 2004, involving a high resistance connection on a fuse holder, resulted in the Division II emergency diesel generator subsystem being vulnerable to electrical circuit failure if called upon to complete its support function. The high resistance connection was caused by degraded grease-like material and dirt. This issue also resluted in the Division II diesel generator failure during a subsequent surveillance test.
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| The inspectors determined that the finding was greater than minor because the finding affected the Mitigating Systems Cornerstone objective of ensuring the availability, reliability, and sapability of mitigating systems to prevent undesirable consequences. The Division II emergency diesel generator 125 VDC system is a backup to the AC oil system in case of a loss of offsite power. Offsite power was not lost, therefore, there was not an actual loss of safety function for the diesel.
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| Corrective actions by the licensee included replacing the fuse and fuse holder and expediting actions to address the extent of condition relative to the as-found condition of the fuse and fuse holder. The finding also affected the cross cutting area of problem identification and resolution since the licensee failed to adequately address the degraded circuit condition in a timely manner.
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| Inspection Report# : 2005009(pdf)
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| Significance: Dec 02, 2005 Identified By: NRC Item Type: NCV NonCited Violation NON-CONSERVATIVE ACCEPTANCE CRITERIA A finding of very low safety significance was identified by the inspectors for a violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control" requirements. Specifically, the licensee failed to incorporate the most restrictive hydraulic conditions into the calculation which established the acceptance criteria for a technical specification surveillance test. This resulted in a HPCS system hydraulic calculation that was non-conservative when determining the pumps minimum acceptance criteria. Once identified, the licensee evaluated operability and entered the finding into their corrective action program to revise the affected documents.
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| The finding was more than minor because the failure to account for all modes of HPCS system operation in the surveillance tests acceptance criteria could result in unacceptable degradation and could have affected the mitigating systems cornerstone objective. The finding was of very low safety significance because the licensees analysis showed that adequate design margin existed for the HPCS system and did not represent an actual loss of a safety function.
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| Inspection Report# : 2005002(pdf)
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| Significance: Dec 02, 2005 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE HEAT EXCHANGER THERMAL PERFORMANCE TESTING A finding of very low safety significance was identified by the inspectors for a violation of 10 CFR Part 50, Appendix B, Criterion XI, "Test Control" requirements. Specifically, in 2000, 2002 and 2003, the licensee failed to recognize that the calculated value for the diesel generator (DG) jacket-water (JW) flow rate, as determined from test data
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| 3Q/2006 Inspection Findings - Clinton Page 4 of 5 obtained during thermal performance testing of the Division III DG JW cooler heat exchanger (HX), was significantly higher than the flow rate that could be attained by the engine-driven water pump. Once identified, the licensee entered the finding into their corrective action program as Condition Report (CR) 426459, NRC SSD&PC Is the Calculated Process Flow Rate Reasonable, dated November 21, 2005, and CR429726, Discrepancies Not Identified in Corrective Action Process, dated December 2, 2005, to evaluate and/or revise the affected test procedures.
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| The finding was more than minor because the failure to account for flow rates that were significantly greater than that identified by the equipments design specification produced equipment performance data that did not accurately demonstrate the HXs availability and reliability. The finding was of very low safety significance because the licensees evaluation showed that the Division III DGs JW Cooler HX would have performed its safety function and did not represent an actual loss of a safety function. A contributing cause of the finding was related to the cross-cutting element of problem identification and resolution. Specifically, a similar issue was identified during another NRC inspection in 2001; however, the licensee did not properly evaluate and take actions. As a result, testing done in 2002 and 2003 showed the same discrepant flow rates.
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| Inspection Report# : 2005002(pdf)
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| Barrier Integrity Emergency Preparedness Occupational Radiation Safety Significance: Mar 31, 2006 Identified By: NRC Item Type: FIN Finding FAILURE TO MAINTAIN COLLECTIVE RADIATION DOSE TO OCCUPATIONAL WORKERS INVOLVED IN REFUEL FLOOR WORK ALARA.
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| An inspector-identified finding of very low safety significance was identified for the failure to maintain the collective dose As-Low-As-Is-Reasonably-Achievable (ALARA) for refuel floor non-cavity work that was conducted during the February 2006 refueling outage. The additional, unintended dose was attributable to deficiencies in both work planning and work execution. The actual collective dose for this work activity was approximately 14 person-rem compared to the licensee's initial dose estimate of 4.4 person-rem. A revised dose estimate of about 7 person-rem was determined by the inspectors based on reasonably unexpected changes in radiological conditions and equipment problems. Consequently, the collective dose for this work exceeded 5 rem and exceeded the revised dose projection by more than 50 percent.
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| The issue was more than minor because it was associated with the Program/Process (ALARA planning) attribute of the Occupational Radiation Safety cornerstone and affected the cornerstone objective to ensure adequate protection of worker health and safety from exposure to radiation. This issue represents a finding of very low safety significance because it involved ALARA planning; however, the Clinton plant's current 3-year rolling average collective dose does not exceed 240 person-rem. The licensee entered this radiological work planning/dose performance problem into its outage lessons learned database to allow the development of measures to better plan and execute refuel floor work during future refueling outages.
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| Inspection Report# : 2006002(pdf)
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| Public Radiation Safety
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| 3Q/2006 Inspection Findings - Clinton Page 5 of 5 Physical Protection Physical Protection information not publicly available.
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| Miscellaneous Last modified : December 21, 2006
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| 4Q/2006 Inspection Findings - Clinton Page 1 of 4 Clinton 4Q/2006 Plant Inspection Findings Initiating Events Significance: Sep 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation THE INSPECTORS DETERMINED THAT THE FAILURE TO APPROPRIATELY IDENTIFY AND CORRECT THE CAUSE OF THE DIVISION 4 NSPS INVERTER IN MARCH WAS A PERFROMANCE DEFICIENCY.
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| A finding of very low safety significance and a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, was self revealed following a reactor scram on August 27, 2006, due to the licensee's failure to identify and correct a condition adverse to quality in March 2006. The licensee determined and corrected the actual cause of the failure and revised procurement procedures to disallow purchase of parts manufactured under the same process as the failed board.
| |
| Additionally, the licensee commenced a common cause evaluation to assist in planning and developing additional corrective actions to address whether there are issues involving the licensee proficiency in identifying causes of operational occurrences.
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| The finding was more than minor because it resulted in a reactor scram and was associated with the equipment performance attribute of the initiating events cornerstone. The finding was of very low safety significance because it would not affect the availability of a mitigating system. The finding was also determined to affect the cross-cutting area of problem identification and resolution in that the actual cause of the March 26, 2006 failure was not properly identified, resulting in the corrective action not addressing the cause, and a more significant failure occurring in August 2006.
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| Inspection Report# : 2006007 (pdf)
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| Significance: Mar 20, 2006 Identified By: Self-Revealing Item Type: FIN Finding INADEQUATE WORKMANSHIP RESULTED IN GENERATOR TRIP AND REACTOR SCRAM.
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| The inspectors considered the failure to adequately tighten terminal screws in the main generator output current transformer circuit a performance deficiency. This issue was caused by inadequate workmanship. The inspectors determined it was more than minor because the finding affected the reactor safety/initiating events cornerstone objective of limiting the likelihood of those events that upset plant stability. The finding also affected the cross-cutting area of human performance because the contract workers failed to tighten the terminal screws of the current transformer and the licensee failed to ensure the GE workers were using the appropriate lifted and landed leads documents to aid in performance of this job.
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| Although this failure occurred in C1R08 in April, 2002, the inspectors determined this deficiency to be reflective of recent licensee performance because, up to the March 2006 scram event, there was no procedure or process in place to ensure GE followed the licensees lifted and landed lead procedures. As a result of the root cause for this event, the licensee initiated a corrective action to revise the GE quality control check-list to confirm that requirements similar to wire removal/jumper installation procedures are incorporated. Although this finding did contribute to the likelihood of a reactor trip, it did not affect the function or availability of any mitigation equipment. Therefore, the inspectors concluded that this issue was a finding of very low safety significance (Green).
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| Inspection Report# : 2006004 (pdf)
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| Mitigating Systems Significance: TBD Nov 17, 2006 Identified By: NRC Item Type: AV Apparent Violation
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| 4Q/2006 Inspection Findings - Clinton Page 2 of 4 HPCS OPERABILITY QUESTIONED DUE TO VORTEXING A finding of greater than very low safety significance was identified by the inspectors for an apparent violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control requirements. Specifically, the licensee failed to adequately address vortexing in the reactor core isolation cooling (RCIC) water storage tank. As a result, the setpoint for the high pressure core spray (HPCS) pump suction source to swap from the RCIC tank to the suppression pool may be too low and result in significant air entrainment such that the HPCS pump would not be capable of completing its safety function. As a corrective action, on December 1, 2005, the licensee shifted the HPCS and RCIC inventory source to the suppression pool as a conservative measure. Vortexing from the suppression pool should not occur due to the depth of the HPCS and RCIC suction lines and the use of the suppression pool as a qualified inventory source was allowed per Clintons Updated Safety Analysis Report (USAR) and Technical Specifications (TS).
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| The finding was greater than minor because if left uncorrected, could result in the HPCS system becoming inoperable due to air entrainment as the water level in the RCIC water tank decreased toward the swapover setpoint. This finding affected the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). This finding was determined to be greater than Green based on the preliminary results of the Phase 2 and Phase 3 analyses Inspection Report# : 2006011 (pdf)
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| Significance: Nov 17, 2006 Identified By: NRC Item Type: NCV NonCited Violation POTENTIAL INOPERABILITY OF RCIC DUE TO VORTEXING A finding of very low safety significance was identified by the inspectors for an Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control requirements. Specifically, in Calculation IP-M-0384, Evaluation of Vortex in the RCIC [Water] Storage Tank, Revisions 0 and 1, the licensee failed to adequately demonstrate that the RCIC pump would be capable of performing its safety function prior to swapping suction paths from the RCIC tank to the suppression pool. As an immediate corrective action, the licensee aligned the suction path of the RCIC system to the suppression pool.
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| The finding was greater than minor because the calculation of record was not adequate and there was reasonable doubt of the successful outcome of a re-analysis. The finding was determined to be of very low safety significance because the inspectors answered no to all five screening questions in the Phase 1 Screening Worksheet under the Mitigating Systems column. After further analysis, the inspectors concluded that the RCIC pump was operable.
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| Inspection Report# : 2006011 (pdf)
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| Significance: Mar 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE DESIGN CONTROL DURING REVIEW OF ENGINEERING CHANGE PACKAGE 356820 "SHUTDOWN COOLING HEADER LEAK-OFF LINE".
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| In February 2006, a finding of very low safety significance involving a Non-Cited Violation of 10 CFR 50, Appendix B, Criteria III, "Design Control," was identified. During a review of Engineering Change Package 356820, "Shutdown Cooling Header Leak-off line," the inspectors identified that the design change, as installed, would adversely impact the functionality of both the Division 2 residual heat removal system's water leg (keep-fill) pump and the C residual heat removal pump. This adverse condition would be caused by the introduction of high temperature water on the suction side of both pumps. The design change was being installed to prevent pressurization of the shutdown cooling header due to leakage through the reactor coolant system pressure isolation valves.
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| This issue was more than minor because the finding affected the Mitigating Systems cornerstone objective of ensuring the availability of mitigating systems to prevent undesirable consequences (Design Control attributes). The finding was of very low safety significance because, with the expected operator actions, this condition would not result in a loss of operability.
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| This conclusion was made based on the flow limiting characteristics of the leak-off line orifice with the suction cooling header volume at saturated conditions in conjunction with the subsequent operator alarm response requirements. Corrective actions by the licensee included procedure revisions and local monitoring of the C residual heat removal suction line temperature once the leak-off line was placed in service.
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| Inspection Report# : 2006002 (pdf)
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| 4Q/2006 Inspection Findings - Clinton Page 3 of 4 Significance: Mar 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE TEST CONTROL DURING THE REVIEW OF THE LICENSEE'S SURVEILLANCE TEST TO DETERMINE OPERABILITY OF THE SHUTDOWN SERVICE WATER SYSTEM.
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| On February 2, 2006, the inspectors identified a finding involving a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XI, "Test Controls." During a review of the licensee's surveillance test to determine the operability of the shutdown service water system, the inspectors identified that the system's leakage could exceed both the administrative and operability limits established by design basis documents, without the test detecting the actual leak rate. This condition was caused by an inadequate test connection.
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| This issue was more than minor because the finding affected the Mitigating Systems cornerstone objective of ensuring the availability of mitigating systems to prevent undesirable consequences. An adverse condition would have been masked by leakage that exceeded both administrative and operability limits, and would not have been identified under testing conditions mandated by the licensee's testing program. the finding was of very low safety significance because the actual measured leakage was well below the capability of accurately being measured, and this issue did not result in a system operability concern. As part of the corrective actions, the licensee planned to performed an extent of condition review to ensure that no other system leakage tests were affected by this issue.
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| Inspection Report# : 2006002 (pdf)
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| Barrier Integrity Emergency Preparedness Occupational Radiation Safety Significance: Mar 31, 2006 Identified By: NRC Item Type: FIN Finding FAILURE TO MAINTAIN COLLECTIVE RADIATION DOSE TO OCCUPATIONAL WORKERS INVOLVED IN REFUEL FLOOR WORK ALARA.
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| An inspector-identified finding of very low safety significance was identified for the failure to maintain the collective dose As-Low-As-Is-Reasonably-Achievable (ALARA) for refuel floor non-cavity work that was conducted during the February 2006 refueling outage. The additional, unintended dose was attributable to deficiencies in both work planning and work execution. The actual collective dose for this work activity was approximately 14 person-rem compared to the licensee's initial dose estimate of 4.4 person-rem. A revised dose estimate of about 7 person-rem was determined by the inspectors based on reasonably unexpected changes in radiological conditions and equipment problems. Consequently, the collective dose for this work exceeded 5 rem and exceeded the revised dose projection by more than 50 percent.
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| The issue was more than minor because it was associated with the Program/Process (ALARA planning) attribute of the Occupational Radiation Safety cornerstone and affected the cornerstone objective to ensure adequate protection of worker health and safety from exposure to radiation. This issue represents a finding of very low safety significance because it involved ALARA planning; however, the Clinton plant's current 3-year rolling average collective dose does not exceed 240 person-rem. The licensee entered this radiological work planning/dose performance problem into its outage lessons learned database to allow the development of measures to better plan and execute refuel floor work during future refueling outages.
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| Inspection Report# : 2006002 (pdf)
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| 4Q/2006 Inspection Findings - Clinton Page 4 of 4 Public Radiation Safety Physical Protection Physical Protection information not publicly available.
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| Miscellaneous Last modified : March 01, 2007
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| Clinton 1Q/2007 Plant Inspection Findings Initiating Events Significance: Sep 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation THE INSPECTORS DETERMINED THAT THE FAILURE TO APPROPRIATELY IDENTIFY AND CORRECT THE CAUSE OF THE DIVISION 4 NSPS INVERTER IN MARCH WAS A PERFROMANCE DEFICIENCY.
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| A finding of very low safety significance and a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, was self revealed following a reactor scram on August 27, 2006, due to the licensee's failure to identify and correct a condition adverse to quality in March 2006. The licensee determined and corrected the actual cause of the failure and revised procurement procedures to disallow purchase of parts manufactured under the same process as the failed board.
| |
| Additionally, the licensee commenced a common cause evaluation to assist in planning and developing additional corrective actions to address whether there are issues involving the licensee proficiency in identifying causes of operational occurrences.
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| The finding was more than minor because it resulted in a reactor scram and was associated with the equipment performance attribute of the initiating events cornerstone. The finding was of very low safety significance because it would not affect the availability of a mitigating system. The finding was also determined to affect the cross-cutting area of problem identification and resolution in that the actual cause of the March 26, 2006 failure was not properly identified, resulting in the corrective action not addressing the cause, and a more significant failure occurring in August 2006.
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| Inspection Report# : 2006007 (pdf)
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| Mitigating Systems Significance: Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation FAILURE OF THE ELECTRICAL CIRCUIT CARD RESULTED IN A LOSS OF SAFETY FUNCTION FOR THE MAIN TURBINE BYPASS VALVES.
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| A finding of very low safety significance involving a Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, was self revealed when a low main condenser vacuum alarm was recieved in the main control room. The alarm was caused by the failure of an electronic circuit card. This circuit card failure also resulted in the main turbine bypass valves being interlocked closed (loss of safety function). The inspectors determined that the cause of this issue was inadequate instructions contained in the licensee's Performance Centered Maintenance (PCM) process.
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| The finding was greater than minor because failure to have adequate instructions to implement an effective preventive maintenance program could be reasonably viewed as a precursor to a more significant event. Additionally, this finding could affect the mitigating systems cornerstone in that it is associated with a degraded condition that could concurrently influence mitigation equipment and the operator's response to an initiating event. This finding was of very low safety significance because the exposure time was of short duration, less than 3 days.
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| Inspection Report# : 2007002 (pdf)
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| Significance: Mar 31, 2007 Identified By: NRC Item Type: FIN Finding FAILURE TO PERFORM AN ADEQUATE CONFIGURATION CONTROL RISK EVALUATION WAS A
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| PERFORMANCE DEFICIENCY WARRANTING A SIGNIFICANCE EVALUATION.
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| A finding of very low safety significance was self-revealed following the loss of the division 3 shutdown service water (SX) system on August 17, 2006. The loss of division 3 of SX occurred when a security guard bumped an SX circuit breaker hand switch for the cross tie valve, 1SX014C, with a piece of protective equipment. This finding resulted from the licensee's failure to do an adequate inadvertent contact configuration control risk assessment during the implementation of a 2005 requirement for security personnel to carry new equipment on their person.
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| The finding was more than minor because it impacted the mitigating systems cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events. With the circuit breaker in the OFF position, 1SX014C would remain opend uring a loss of offsite power event. In this configuration, the SX system could not perform its safety function of supplying cooling water to both the division 3 diesel generator and the high pressure core spray pump room cooling system. This finding was of very low safety significance due to the short duration exposure time, less than three days, and credit for operator actions to restore the system back to service. This finding affected the work practices component of the cross-cutting area of human performance. Licensee management failed to ensure the proper management and oversight of security personnel rounds activities.
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| Inspection Report# : 2007002 (pdf)
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| Significance: Nov 17, 2006 Identified By: NRC Item Type: NCV NonCited Violation POTENTIAL INOPERABILITY OF RCIC DUE TO VORTEXING A finding of very low safety significance was identified by the inspectors for an Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control requirements. Specifically, in Calculation IP-M-0384, Evaluation of Vortex in the RCIC [Water] Storage Tank, Revisions 0 and 1, the licensee failed to adequately demonstrate that the RCIC pump would be capable of performing its safety function prior to swapping suction paths from the RCIC tank to the suppression pool. As an immediate corrective action, the licensee aligned the suction path of the RCIC system to the suppression pool.
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| The finding was greater than minor because the calculation of record was not adequate and there was reasonable doubt of the successful outcome of a re-analysis. The finding was determined to be of very low safety significance because the inspectors answered no to all five screening questions in the Phase 1 Screening Worksheet under the Mitigating Systems column. After further analysis, the inspectors concluded that the RCIC pump was operable.
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| Inspection Report# : 2006011 (pdf)
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| Significance: Aug 12, 2006 Identified By: NRC Item Type: VIO Violation HPCS OPERABILITY QUESTIONED DUE TO VORTEXING White. A finding of low to moderate safety significance was identified by the inspectors for a violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control requirements. Specifically, the licensee failed to adequately address vortexing in the reactor core isolation cooling (RCIC) water storage tank. As a result, the setpoint for the high pressure core spray (HPCS) pump suction source to swap from the RCIC tank to the suppression pool may be too low and result in significant air entrainment such that the HPCS pump would not be capable of completing its safety function. As a corrective action, on December 1, 2005, the licensee shifted the HPCS and RCIC inventory source to the suppression pool as a conservative measure. Vortexing from the suppression pool should not occur due to the depth of the HPCS and RCIC suction lines and the use of the suppression pool as a qualified inventory source was allowed per Clintons Updated Safety Analysis Report (USAR) and Technical Specifications (TS).
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| The finding was greater than minor because if left uncorrected, could result in the HPCS system becoming inoperable due to air entrainment as the water level in the RCIC water tank decreased toward the swapover setpoint. This finding affected the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage)
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| Based on the discussion during the regulatory conference, the NRC determined that operators would be directed to throttle HPCS in response to transient (i.e., non- Loss of Coolant Accidents and non- Anticipated Transient Without a Scram) scenarios. If operators successfully throttle the HPCS injection valve, the system flow rate will be low enough that air entrainment during suction swap-over to the suppression pool would no longer be a concern. For the final significance
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| determination, the NRC assumed that HPCS would fail in response to transient initiating events only if the operator failed to properly throttle the HPCS injection valve. For all other initiating events, HPCS was assumed to fail during the suction transfer, consistent with the assumption in the preliminary significance determination. Given the inherent uncertainty in estimating human error probabilities, the NRC used its best estimate of 2.6E-2 for the human error probability in the final significance determination.
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| Inspection Report# : 2007006 (pdf)
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| Inspection Report# : 2006011 (pdf)
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| Barrier Integrity Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Physical Protection information not publicly available.
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| Miscellaneous Last modified : June 01, 2007
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| Clinton 2Q/2007 Plant Inspection Findings Initiating Events Significance: Sep 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation THE INSPECTORS DETERMINED THAT THE FAILURE TO APPROPRIATELY IDENTIFY AND CORRECT THE CAUSE OF THE DIVISION 4 NSPS INVERTER IN MARCH WAS A PERFROMANCE DEFICIENCY.
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| A finding of very low safety significance and a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, was self revealed following a reactor scram on August 27, 2006, due to the licensee's failure to identify and correct a condition adverse to quality in March 2006. The licensee determined and corrected the actual cause of the failure and revised procurement procedures to disallow purchase of parts manufactured under the same process as the failed board. Additionally, the licensee commenced a common cause evaluation to assist in planning and developing additional corrective actions to address whether there are issues involving the licensee proficiency in identifying causes of operational occurrences.
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| The finding was more than minor because it resulted in a reactor scram and was associated with the equipment performance attribute of the initiating events cornerstone. The finding was of very low safety significance because it would not affect the availability of a mitigating system. The finding was also determined to affect the cross-cutting area of problem identification and resolution in that the actual cause of the March 26, 2006 failure was not properly identified, resulting in the corrective action not addressing the cause, and a more significant failure occurring in August 2006.
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| Inspection Report# : 2006007 (pdf)
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| Mitigating Systems Significance: Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation FAILURE OF THE ELECTRICAL CIRCUIT CARD RESULTED IN A LOSS OF SAFETY FUNCTION FOR THE MAIN TURBINE BYPASS VALVES.
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| A finding of very low safety significance involving a Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, was self revealed when a low main condenser vacuum alarm was recieved in the main control room. The alarm was caused by the failure of an electronic circuit card. This circuit card failure also resulted in the main turbine bypass valves being interlocked closed (loss of safety function). The inspectors determined that the cause of this issue was inadequate instructions contained in the licensee's Performance Centered Maintenance (PCM) process.
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| The finding was greater than minor because failure to have adequate instructions to implement an effective preventive maintenance program could be reasonably viewed as a precursor to a more significant event. Additionally, this finding could affect the mitigating systems cornerstone in that it is associated with a degraded condition that could concurrently influence mitigation equipment and the operator's response to an initiating event. This finding was of very low safety significance because the exposure time was of short duration, less than 3 days.
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| Inspection Report# : 2007002 (pdf)
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| Significance: Mar 31, 2007 Identified By: NRC
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| Item Type: FIN Finding FAILURE TO PERFORM AN ADEQUATE CONFIGURATION CONTROL RISK EVALUATION WAS A PERFORMANCE DEFICIENCY WARRANTING A SIGNIFICANCE EVALUATION.
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| A finding of very low safety significance was self-revealed following the loss of the division 3 shutdown service water (SX) system on August 17, 2006. The loss of division 3 of SX occurred when a security guard bumped an SX circuit breaker hand switch for the cross tie valve, 1SX014C, with a piece of protective equipment. This finding resulted from the licensee's failure to do an adequate inadvertent contact configuration control risk assessment during the implementation of a 2005 requirement for security personnel to carry new equipment on their person.
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| The finding was more than minor because it impacted the mitigating systems cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events. With the circuit breaker in the OFF position, 1SX014C would remain opend uring a loss of offsite power event. In this configuration, the SX system could not perform its safety function of supplying cooling water to both the division 3 diesel generator and the high pressure core spray pump room cooling system. This finding was of very low safety significance due to the short duration exposure time, less than three days, and credit for operator actions to restore the system back to service. This finding affected the work practices component of the cross-cutting area of human performance. Licensee management failed to ensure the proper management and oversight of security personnel rounds activities.
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| Inspection Report# : 2007002 (pdf)
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| Significance: Nov 17, 2006 Identified By: NRC Item Type: NCV NonCited Violation POTENTIAL INOPERABILITY OF RCIC DUE TO VORTEXING A finding of very low safety significance was identified by the inspectors for an Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control requirements. Specifically, in Calculation IP-M-0384, Evaluation of Vortex in the RCIC [Water] Storage Tank, Revisions 0 and 1, the licensee failed to adequately demonstrate that the RCIC pump would be capable of performing its safety function prior to swapping suction paths from the RCIC tank to the suppression pool. As an immediate corrective action, the licensee aligned the suction path of the RCIC system to the suppression pool.
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| The finding was greater than minor because the calculation of record was not adequate and there was reasonable doubt of the successful outcome of a re-analysis. The finding was determined to be of very low safety significance because the inspectors answered no to all five screening questions in the Phase 1 Screening Worksheet under the Mitigating Systems column. After further analysis, the inspectors concluded that the RCIC pump was operable.
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| Inspection Report# : 2006011 (pdf)
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| Significance: Aug 12, 2006 Identified By: NRC Item Type: VIO Violation HPCS OPERABILITY QUESTIONED DUE TO VORTEXING White. A finding of low to moderate safety significance was identified by the inspectors for a violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control requirements. Specifically, the licensee failed to adequately address vortexing in the reactor core isolation cooling (RCIC) water storage tank. As a result, the setpoint for the high pressure core spray (HPCS) pump suction source to swap from the RCIC tank to the suppression pool may be too low and result in significant air entrainment such that the HPCS pump would not be capable of completing its safety function. As a corrective action, on December 1, 2005, the licensee shifted the HPCS and RCIC inventory source to the suppression pool as a conservative measure. Vortexing from the suppression pool should not occur due to the depth of the HPCS and RCIC suction lines and the use of the suppression pool as a qualified inventory source was allowed per Clintons Updated Safety Analysis Report (USAR) and Technical Specifications (TS).
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| The finding was greater than minor because if left uncorrected, could result in the HPCS system becoming inoperable due to air entrainment as the water level in the RCIC water tank decreased toward the swapover setpoint. This finding affected the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage)
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| Based on the discussion during the regulatory conference, the NRC determined that operators would be directed to
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| throttle HPCS in response to transient (i.e., non- Loss of Coolant Accidents and non- Anticipated Transient Without a Scram) scenarios. If operators successfully throttle the HPCS injection valve, the system flow rate will be low enough that air entrainment during suction swap-over to the suppression pool would no longer be a concern. For the final significance determination, the NRC assumed that HPCS would fail in response to transient initiating events only if the operator failed to properly throttle the HPCS injection valve. For all other initiating events, HPCS was assumed to fail during the suction transfer, consistent with the assumption in the preliminary significance determination. Given the inherent uncertainty in estimating human error probabilities, the NRC used its best estimate of 2.6E-2 for the human error probability in the final significance determination.
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| Inspection Report# : 2007006 (pdf)
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| Inspection Report# : 2006011 (pdf)
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| Barrier Integrity Significance: Mar 23, 2007 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW CLINOTN PROCEDURE 1019.05, "TRANSIENT EQUIPMENT/MATERIALS,"
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| RESULTED IN VIOLATION OF 10 CFR PART 50, APPENDIX B, CRITERION V.
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| The inspectors identified an NCV of 10 CFR Part 50, Appendix V, "Instructions, Procedures, and Drawings," for failure to assure that activites affecting quality be accomplished in accordance with prescribed documented instructions, procedures, or drawings. Contrary to CPS procedure 1019.05, "Transient Equipment/Materials," step 8.5.3, four radiation protection stanchions were secured to the 755' elevation in the containment building with ty-raps instead of metal grating clips. The licensee removed the stanchions, performed a walkdown of containment to ensure there were no other improperly installed stanchions, and entered the performance deficiency into the CAP for resolution.
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| The finding was associated with the Barrier Integrity Cornerstone. The finding was more than minor because the finding was viewed as a presursor to a significant event. If left uncorrected, the stanchions could become missiles during a suppression pool swell event, potentially damaging containment isolation valves. The inspectors assessed the significance of this finding as very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of the reactor containment. The finding was associated with cross-cutting aspect P.1c, thoroughly Evaluate Problems, of the problem identification and resolution cross-cutting area, in that, the licensee's initial reviews of the issue failed to evaluate the potential design basis impact.
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| Inspection Report# : 2007007 (pdf)
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| Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings
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| pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.
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| Miscellaneous Last modified : August 24, 2007
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| Clinton 3Q/2007 Plant Inspection Findings Initiating Events Mitigating Systems Significance: Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation FAILURE OF THE ELECTRICAL CIRCUIT CARD RESULTED IN A LOSS OF SAFETY FUNCTION FOR THE MAIN TURBINE BYPASS VALVES.
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| A finding of very low safety significance involving a Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, was self revealed when a low main condenser vacuum alarm was recieved in the main control room. The alarm was caused by the failure of an electronic circuit card. This circuit card failure also resulted in the main turbine bypass valves being interlocked closed (loss of safety function). The inspectors determined that the cause of this issue was inadequate instructions contained in the licensee's Performance Centered Maintenance (PCM) process.
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| The finding was greater than minor because failure to have adequate instructions to implement an effective preventive maintenance program could be reasonably viewed as a precursor to a more significant event. Additionally, this finding could affect the mitigating systems cornerstone in that it is associated with a degraded condition that could concurrently influence mitigation equipment and the operator's response to an initiating event. This finding was of very low safety significance because the exposure time was of short duration, less than 3 days.
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| Inspection Report# : 2007002 (pdf)
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| Significance: Mar 31, 2007 Identified By: NRC Item Type: FIN Finding FAILURE TO PERFORM AN ADEQUATE CONFIGURATION CONTROL RISK EVALUATION WAS A PERFORMANCE DEFICIENCY WARRANTING A SIGNIFICANCE EVALUATION.
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| A finding of very low safety significance was self-revealed following the loss of the division 3 shutdown service water (SX) system on August 17, 2006. The loss of division 3 of SX occurred when a security guard bumped an SX circuit breaker hand switch for the cross tie valve, 1SX014C, with a piece of protective equipment. This finding resulted from the licensee's failure to do an adequate inadvertent contact configuration control risk assessment during the implementation of a 2005 requirement for security personnel to carry new equipment on their person.
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| The finding was more than minor because it impacted the mitigating systems cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events. With the circuit breaker in the OFF position, 1SX014C would remain opend uring a loss of offsite power event. In this configuration, the SX system could not perform its safety function of supplying cooling water to both the division 3 diesel generator and the high pressure core spray pump room cooling system. This finding was of very low safety significance due to the short duration exposure time, less than three days, and credit for operator actions to restore the system back to service. This finding affected the work practices component of the cross-cutting area of human performance. Licensee management failed to ensure the proper management and oversight of security personnel rounds activities.
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| Inspection Report# : 2007002 (pdf)
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| Significance: Nov 17, 2006 Identified By: NRC Item Type: NCV NonCited Violation
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| POTENTIAL INOPERABILITY OF RCIC DUE TO VORTEXING A finding of very low safety significance was identified by the inspectors for an Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control requirements. Specifically, in Calculation IP-M-0384, Evaluation of Vortex in the RCIC [Water] Storage Tank, Revisions 0 and 1, the licensee failed to adequately demonstrate that the RCIC pump would be capable of performing its safety function prior to swapping suction paths from the RCIC tank to the suppression pool. As an immediate corrective action, the licensee aligned the suction path of the RCIC system to the suppression pool.
| |
| The finding was greater than minor because the calculation of record was not adequate and there was reasonable doubt of the successful outcome of a re-analysis. The finding was determined to be of very low safety significance because the inspectors answered no to all five screening questions in the Phase 1 Screening Worksheet under the Mitigating Systems column. After further analysis, the inspectors concluded that the RCIC pump was operable.
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| Inspection Report# : 2006011 (pdf)
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| Significance: Aug 12, 2006 Identified By: NRC Item Type: VIO Violation HPCS OPERABILITY QUESTIONED DUE TO VORTEXING White. A finding of low to moderate safety significance was identified by the inspectors for a violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control requirements. Specifically, the licensee failed to adequately address vortexing in the reactor core isolation cooling (RCIC) water storage tank. As a result, the setpoint for the high pressure core spray (HPCS) pump suction source to swap from the RCIC tank to the suppression pool may be too low and result in significant air entrainment such that the HPCS pump would not be capable of completing its safety function. As a corrective action, on December 1, 2005, the licensee shifted the HPCS and RCIC inventory source to the suppression pool as a conservative measure. Vortexing from the suppression pool should not occur due to the depth of the HPCS and RCIC suction lines and the use of the suppression pool as a qualified inventory source was allowed per Clintons Updated Safety Analysis Report (USAR) and Technical Specifications (TS).
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| The finding was greater than minor because if left uncorrected, could result in the HPCS system becoming inoperable due to air entrainment as the water level in the RCIC water tank decreased toward the swapover setpoint. This finding affected the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage)
| |
| Based on the discussion during the regulatory conference, the NRC determined that operators would be directed to throttle HPCS in response to transient (i.e., non- Loss of Coolant Accidents and non- Anticipated Transient Without a Scram) scenarios. If operators successfully throttle the HPCS injection valve, the system flow rate will be low enough that air entrainment during suction swap-over to the suppression pool would no longer be a concern. For the final significance determination, the NRC assumed that HPCS would fail in response to transient initiating events only if the operator failed to properly throttle the HPCS injection valve. For all other initiating events, HPCS was assumed to fail during the suction transfer, consistent with the assumption in the preliminary significance determination. Given the inherent uncertainty in estimating human error probabilities, the NRC used its best estimate of 2.6E-2 for the human error probability in the final significance determination.
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| Inspection Report# : 2006011 (pdf)
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| Inspection Report# : 2007006 (pdf)
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| Barrier Integrity Significance: Sep 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE PROCUREMENT SPECIFICATION FOR CHARCOAL RESULTS IN INOPERABLE CONTROL ROOM VENTILATION SUBSYSTEM.
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| A performance deficiency involving a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion IV,
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| "Procurement Document Control," was self revealed following receipt of laboratory results that showed that Division 1 control room ventilation system charcoal filter penetration values were higher than allowed by Clinton's Technical Specifications. This issue occurred because the licensee failed to establish proper purchase specifications for charcoal used in the control room ventilation system. Additionally, this issue led to Division 1 control room ventilation subsystem being inoperable from May 9 through May 14, 2005. Licensee corrective actions included entering the issue into the corrective action program, revising the charcoal purchase specifications, and adding limitations to work orders to prevent scheduling work that could impact the operability of redundant systems.
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| This issue was more than minor because it affected the objective of the Barrier Integrity cornerstone of assuring that physical design barriers protect the public from radionuclide releases caused by accident or events. Additionally, this issue is associated with the barrier perfromance attribute of maintaining Radiological Barrier functionality of the control room. Failure to ensure adequate purchase specifications resulted in there being a period where both trains of control room ventilation were inoperable without the knowledge of the operators. The issue was of very low safety significance because it only represented a degradation of the radiological barrier function provided for the control room.
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| Inspection Report# : 2007004 (pdf)
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| Significance: Sep 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO COMPLY WITH TECHNICAL SPECIFICATION 3.4.5 FOR RCS PRESSURE BOUNDARY LEAK The inspectors identified a performance deficiency involving a Non-Cited Violation of Technical Specifications when the licensee failed to meet the required completion time for an action statement in Technical Specification 3.4.5.
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| Specifically, Technical Specification 3.4.5 does not allow reactor coolant system pressure boundary leakage and requires a shutdown to Mode 3 within 12 hours if pressure boundary leakage is discovered. Upon entry into the drywell following a shutdown of the reactor on June 19, 2007, the licensee discovered the existence of reactor coolant system pressure boundary leakage. Indications of the leakage had been discovered at 0433 on June 18, 2007, but the plant was not placed in Mode 3 until approximately 31 hours later at 1125 on June 19, 2007. Licensee corrective actions included replacing the leaking flexible hose, scheduling replacement of other flexible hoses, and establishing a preventive maintenance replacement frequency for the flexible hoses.
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| This issue was more than minor because oeprating with a degraded pressure boundary affected the reactor coolant system equipment and barrier performance attribute of the Barrier Integrity cornerstone, in that, reactor coolant system pressure boundary leakage results in a reduction in the reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The issue was of very low safety significance because the potential maximum size of the leak was well within the capability of the available mitigating equipment. The finding is related to the cross-cutting area of Human Performance (Decision Making) in the operators had initially entered TS 3.4.5 for pressure boundary leakage, but later chose not to treat the leakage as pressure boundary leakage, and treat it as unidentified leakage until the actual location could be determined.
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| Inspection Report# : 2007004 (pdf)
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| Significance: Sep 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE PROCEDURE RESULTS IN SPENT FUEL BUNDLE INCIDENT A performance deficiency involving a Non-Cited Violation of 10 CFR Part 50 Appendix B, Criteria V, "Instructions, Procedures, and Drawings," was self-revealed following an event on August 17, 2007, where a spent fuel bundle being moved to a temporary storage location came in contact with and rested upon another fuel bundle seated in its storage location. The licensee procedure that governs spent fuel pool movement failed to provide adequate guidance on how high to lift the fuel bundle prior to traversing across the spent fuel pool. Licensee corrective actions included revising the fuel handling procedure to provide specific instructions regarding how high to lift a fuel bundle during spent fuel pool movements.
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| This issue was more than minor because it affected the barrier integrity objective of assuring that physcial design barriers protect the public from radionuclide releases caused by accidents or events. The inspectors determined that
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| this issue only degraded the Fuel Cladding Barrier and its associated cornerstone, therefore, this issue was of very low safety significance. This finding is related to the cross-cutting are of Human Performance (Resources) because the licensee did not provide complete and accurate procedures. Specifically, the procedure relied on the skills of the operator, did not provided specific values on how high to life a fuel bundle, and did not require independent verification Inspection Report# : 2007004 (pdf)
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| Significance: Mar 23, 2007 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW CLINOTN PROCEDURE 1019.05, "TRANSIENT EQUIPMENT/MATERIALS,"
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| RESULTED IN VIOLATION OF 10 CFR PART 50, APPENDIX B, CRITERION V.
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| The inspectors identified an NCV of 10 CFR Part 50, Appendix V, "Instructions, Procedures, and Drawings," for failure to assure that activites affecting quality be accomplished in accordance with prescribed documented instructions, procedures, or drawings. Contrary to CPS procedure 1019.05, "Transient Equipment/Materials," step 8.5.3, four radiation protection stanchions were secured to the 755' elevation in the containment building with ty-raps instead of metal grating clips. The licensee removed the stanchions, performed a walkdown of containment to ensure there were no other improperly installed stanchions, and entered the performance deficiency into the CAP for resolution.
| |
| The finding was associated with the Barrier Integrity Cornerstone. The finding was more than minor because the finding was viewed as a presursor to a significant event. If left uncorrected, the stanchions could become missiles during a suppression pool swell event, potentially damaging containment isolation valves. The inspectors assessed the significance of this finding as very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of the reactor containment. The finding was associated with cross-cutting aspect P.1c, thoroughly Evaluate Problems, of the problem identification and resolution cross-cutting area, in that, the licensee's initial reviews of the issue failed to evaluate the potential design basis impact.
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| Inspection Report# : 2007007 (pdf)
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| Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.
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| Miscellaneous Last modified : December 07, 2007
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| Clinton 4Q/2007 Plant Inspection Findings Initiating Events Mitigating Systems Significance: Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation FAILURE OF THE ELECTRICAL CIRCUIT CARD RESULTED IN A LOSS OF SAFETY FUNCTION FOR THE MAIN TURBINE BYPASS VALVES.
| |
| A finding of very low safety significance involving a Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, was self revealed when a low main condenser vacuum alarm was recieved in the main control room. The alarm was caused by the failure of an electronic circuit card. This circuit card failure also resulted in the main turbine bypass valves being interlocked closed (loss of safety function). The inspectors determined that the cause of this issue was inadequate instructions contained in the licensee's Performance Centered Maintenance (PCM) process.
| |
| The finding was greater than minor because failure to have adequate instructions to implement an effective preventive maintenance program could be reasonably viewed as a precursor to a more significant event. Additionally, this finding could affect the mitigating systems cornerstone in that it is associated with a degraded condition that could concurrently influence mitigation equipment and the operator's response to an initiating event. This finding was of very low safety significance because the exposure time was of short duration, less than 3 days.
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| Inspection Report# : 2007002 (pdf)
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| Significance: Mar 31, 2007 Identified By: NRC Item Type: FIN Finding FAILURE TO PERFORM AN ADEQUATE CONFIGURATION CONTROL RISK EVALUATION WAS A PERFORMANCE DEFICIENCY WARRANTING A SIGNIFICANCE EVALUATION.
| |
| A finding of very low safety significance was self-revealed following the loss of the division 3 shutdown service water (SX) system on August 17, 2006. The loss of division 3 of SX occurred when a security guard bumped an SX circuit breaker hand switch for the cross tie valve, 1SX014C, with a piece of protective equipment. This finding resulted from the licensee's failure to do an adequate inadvertent contact configuration control risk assessment during the implementation of a 2005 requirement for security personnel to carry new equipment on their person.
| |
| The finding was more than minor because it impacted the mitigating systems cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events. With the circuit breaker in the OFF position, 1SX014C would remain opend uring a loss of offsite power event. In this configuration, the SX system could not perform its safety function of supplying cooling water to both the division 3 diesel generator and the high pressure core spray pump room cooling system. This finding was of very low safety significance due to the short duration exposure time, less than three days, and credit for operator actions to restore the system back to service. This finding affected the work practices component of the cross-cutting area of human performance. Licensee management failed to ensure the proper management and oversight of security personnel rounds activities.
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| Inspection Report# : 2007002 (pdf)
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| Barrier Integrity
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| Significance: Sep 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE PROCUREMENT SPECIFICATION FOR CHARCOAL RESULTS IN INOPERABLE CONTROL ROOM VENTILATION SUBSYSTEM.
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| A performance deficiency involving a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion IV, "Procurement Document Control," was self revealed following receipt of laboratory results that showed that Division 1 control room ventilation system charcoal filter penetration values were higher than allowed by Clinton's Technical Specifications. This issue occurred because the licensee failed to establish proper purchase specifications for charcoal used in the control room ventilation system. Additionally, this issue led to Division 1 control room ventilation subsystem being inoperable from May 9 through May 14, 2005. Licensee corrective actions included entering the issue into the corrective action program, revising the charcoal purchase specifications, and adding limitations to work orders to prevent scheduling work that could impact the operability of redundant systems.
| |
| This issue was more than minor because it affected the objective of the Barrier Integrity cornerstone of assuring that physical design barriers protect the public from radionuclide releases caused by accident or events. Additionally, this issue is associated with the barrier perfromance attribute of maintaining Radiological Barrier functionality of the control room. Failure to ensure adequate purchase specifications resulted in there being a period where both trains of control room ventilation were inoperable without the knowledge of the operators. The issue was of very low safety significance because it only represented a degradation of the radiological barrier function provided for the control room.
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| Inspection Report# : 2007004 (pdf)
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| Significance: Sep 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO COMPLY WITH TECHNICAL SPECIFICATION 3.4.5 FOR RCS PRESSURE BOUNDARY LEAK The inspectors identified a performance deficiency involving a Non-Cited Violation of Technical Specifications when the licensee failed to meet the required completion time for an action statement in Technical Specification 3.4.5.
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| Specifically, Technical Specification 3.4.5 does not allow reactor coolant system pressure boundary leakage and requires a shutdown to Mode 3 within 12 hours if pressure boundary leakage is discovered. Upon entry into the drywell following a shutdown of the reactor on June 19, 2007, the licensee discovered the existence of reactor coolant system pressure boundary leakage. Indications of the leakage had been discovered at 0433 on June 18, 2007, but the plant was not placed in Mode 3 until approximately 31 hours later at 1125 on June 19, 2007. Licensee corrective actions included replacing the leaking flexible hose, scheduling replacement of other flexible hoses, and establishing a preventive maintenance replacement frequency for the flexible hoses.
| |
| This issue was more than minor because oeprating with a degraded pressure boundary affected the reactor coolant system equipment and barrier performance attribute of the Barrier Integrity cornerstone, in that, reactor coolant system pressure boundary leakage results in a reduction in the reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The issue was of very low safety significance because the potential maximum size of the leak was well within the capability of the available mitigating equipment. The finding is related to the cross-cutting area of Human Performance (Decision Making) in the operators had initially entered TS 3.4.5 for pressure boundary leakage, but later chose not to treat the leakage as pressure boundary leakage, and treat it as unidentified leakage until the actual location could be determined.
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| Inspection Report# : 2007004 (pdf)
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| Significance: Sep 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE PROCEDURE RESULTS IN SPENT FUEL BUNDLE INCIDENT A performance deficiency involving a Non-Cited Violation of 10 CFR Part 50 Appendix B, Criteria V, "Instructions, Procedures, and Drawings," was self-revealed following an event on August 17, 2007, where a spent fuel bundle being moved to a temporary storage location came in contact with and rested upon another fuel bundle seated in its
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| storage location. The licensee procedure that governs spent fuel pool movement failed to provide adequate guidance on how high to lift the fuel bundle prior to traversing across the spent fuel pool. Licensee corrective actions included revising the fuel handling procedure to provide specific instructions regarding how high to lift a fuel bundle during spent fuel pool movements.
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| This issue was more than minor because it affected the barrier integrity objective of assuring that physcial design barriers protect the public from radionuclide releases caused by accidents or events. The inspectors determined that this issue only degraded the Fuel Cladding Barrier and its associated cornerstone, therefore, this issue was of very low safety significance. This finding is related to the cross-cutting are of Human Performance (Resources) because the licensee did not provide complete and accurate procedures. Specifically, the procedure relied on the skills of the operator, did not provided specific values on how high to life a fuel bundle, and did not require independent verification Inspection Report# : 2007004 (pdf)
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| Significance: Mar 23, 2007 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW CLINOTN PROCEDURE 1019.05, "TRANSIENT EQUIPMENT/MATERIALS,"
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| RESULTED IN VIOLATION OF 10 CFR PART 50, APPENDIX B, CRITERION V.
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| The inspectors identified an NCV of 10 CFR Part 50, Appendix V, "Instructions, Procedures, and Drawings," for failure to assure that activites affecting quality be accomplished in accordance with prescribed documented instructions, procedures, or drawings. Contrary to CPS procedure 1019.05, "Transient Equipment/Materials," step 8.5.3, four radiation protection stanchions were secured to the 755' elevation in the containment building with ty-raps instead of metal grating clips. The licensee removed the stanchions, performed a walkdown of containment to ensure there were no other improperly installed stanchions, and entered the performance deficiency into the CAP for resolution.
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| The finding was associated with the Barrier Integrity Cornerstone. The finding was more than minor because the finding was viewed as a presursor to a significant event. If left uncorrected, the stanchions could become missiles during a suppression pool swell event, potentially damaging containment isolation valves. The inspectors assessed the significance of this finding as very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of the reactor containment. The finding was associated with cross-cutting aspect P.1c, thoroughly Evaluate Problems, of the problem identification and resolution cross-cutting area, in that, the licensee's initial reviews of the issue failed to evaluate the potential design basis impact.
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| Inspection Report# : 2007007 (pdf)
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| Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.
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| Miscellaneous Last modified : February 04, 2008
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| Clinton 1Q/2008 Plant Inspection Findings Initiating Events Significance: Mar 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW APPROVED FIRE PROTECTION PROGRAM PROCEDURES CONCERNING CONTROL OF TRANSIENT COMBUSTIBLE MATERIAL.
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| The inspectors identified a performance deficiency involving a Non-Cited Violation (NCV) of Clinton Power Station Operating License NPF-62, Section 2.F for failure to implement the fire protection program in accordance with program requirements. The inspectors identified multiple instances of the licensee's failure to follow approved fire protection program procedures concerning control of transient combustible material. Corrective actions for this issue included removing the unattended combustible material, initiating transient combustible permits, and/or initiating compensatory measures.
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| The inspectors determined that this issue was more than minor because the identified transient combustibles were in a combustible free zone required for separation of redundant trains. This finding was of very low safety significance because the transient combustible materials identified by the inspectors were not combustibles of significance. The inspectors determined that this finding was cross-cutting in the area of Problem Identification and Resolution.
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| Specifically, the licensee implements a corrective action program with a low treshold for identifying issues. The licensee identifies such issues completely, accurately, and in a timely manner commensurate with their safety significance (P.1(a)).
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| Inspection Report# : 2008002 (pdf)
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| Significance: Mar 31, 2008 Identified By: NRC Item Type: FIN Finding THE LICENSEE DISCOVERED THAT THE WRONG COMPONENT WAS INSTALLED IN THE B TURBINE DRIVEN REACTOR FEED PUMP OIL PRESSURE SENSING LOGIC.
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| A finding of very low safety significance was self-revealed by the automatic runback of the turbine driven reactor feed pump during post outage power ascension. The licensee discovered that the wrong component was installed in the B turbine driven reactor feed pump oil pressure sensing logic. The inspectors determined that the licensee failed to perform an adequate post maintenance test in accordance with procedures. This issue resulted in an unexpected power change from 54 percent power to 46 percent power. The licensee entered the issue into the corrective action program, performed tailgate discussions with technicians and work planners on the oil pressure switches were up to date in the materials and work management computer system.
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| The inspectors determined this issue was more than minor because it was associated with the Human Performance attribute of the Initiating Events Cornerstone and affected the cornerstone objective of limiting the frequency of those events that upset plant stability. Specifically, the failure to perform adequate post maintenance testing of pressure switch 1PS-FW 135 permitted the wrong component to be installed and placed in service. This deficiency ultimately resulted in an unplanned plant transient. The finding was of very low safety significance because this issue did not increase the likelihood that mitigation equipment or fundctions would not be available. The inspectors also concluded that the failure of the technician to properly follow calibration procedure 8801.01 during the initial calibration of this switch represented a cross-cutting issue in the area of Human Performance, Work Practices (H.4(b)), because licensee personnel failed to follow procedures in regard to pressure switch calibration.
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| Inspection Report# : 2008002 (pdf)
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| Significance: Mar 31, 2008 Identified By: NRC
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| Item Type: NCV NonCited Violation DURING THE PERFORMANCE OF NRC FINAL DRYWELL CLOSEOUT, THE INSPECTORS NOTED THAT FOREIGN MATERIAL/HOUSEKEEPING SOCK HAD NOT BEEN REMOVED FROM THE DRYWELL FLOOR DRAINS.
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| The inspectors identified a finding and an associated NCV of 10 CFR Part 50, Appendix B, Criteria V, "Instructions, Procedures, and Drawings," having a very low safety significance during drywell closeout inspections. Specifically, during the performance of the NRC final drywell closeout, the inspectors noted that foreigh material/houskeeping socks had not been removed from the drywell floor drains. This issue could have resulted in the drywell leak detection system being inoperable following a reactor event. The licensee proecdures for drywell closeout directed licensee staff to remove all loose material and devices associated with the licensee material condition and housekeeping program.
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| The licensee's corrective actions for this issue included removing the floor drain socks and incorporating a work activities item for sock removal in the outage schedule template.
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| The inspectors determined that this issue was more than minor because, if left uncorrected, it could result in a more significant safety concern. Failure to remove drain socks from drywell floor drains could result in the inability to readily detect and tract unidentified leakage following a reactor event. the finding was of very low safety significance because this finding didi not result in exceeding the Technical Specification limit for reactor coolant system (RCS) leakage nor did it affect other mitigating systems resulting in a ttal loss of their safety function. The inspectors also concluded this this issue was a result of no work tiem in the outage schedule to remove the socks, and therefore represented a cross-cutting issue in the are of Human Performance, Work Control (H.3(b)).
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| Inspection Report# : 2008002 (pdf)
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| Significance: Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement fire protection program in accordance with program requirements Identified a performance deficiency involving a NCV of Clinton Power Station Operating License NPF-62, Section 2.F for failure to implement fire protection program in accordance with program requirements. Inspectors identified multiple instances of the licensee failure to follow program procedures concerning control of Transient Combustible Material and Fire Protection Impairment Reporting. Corrective actions included removing the unattended combustible material and repairing latches on the fire doors.
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| This issue was more than minor because it could be a precursor to a significant event. A fire had potential of impacting safety related equipment used for safe shutdown purposes. This finding was of very low safety significance because the transient material identified by the inspectors were not combustibles of significance, and the licensee maintained fire suppression systems in the areas where the fire door latches were not functional. This finding was cross-cutting in the area of P.1(a) because the licensee failed to identify these issues in their corrective action program.
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| Inspection Report# : 2007005 (pdf)
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| Mitigating Systems Significance: Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PERFORM POST MODIFICATION TESTING TO SHUTDOWN SERVICE WATER VACUUM BREAKERS Identified a NCV of 10 CFR Part 50, App B, Crit XI, Test Controls, having a low safety significance for failure to properly test a permanent plant modification to the Div 1 & 2 SX. This resulted in two of four vacuum breakers that failed the minimal design specification during testing.
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| It was determined that the issue was more than minor because it is viewed as a precursor to a significant event. Failure to perform modification testing could lead to components within safety-related systems that do not work as designed.
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| Through detailed analysis the licensee concluded that the hydraulic experience with the vacuum breaker not meeting the minimal design specification would not make the shutdown service water system inoperable. This finding had a cross-cutting aspect in the area of H.4(c) because there were multiple opportunities for the licensee engineering staff to identify the need for this testing.
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| Inspection Report# : 2007005 (pdf)
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| Significance: Dec 19, 2007 Identified By: NRC Item Type: NCV NonCited Violation Continuously Submerged Cables Design Deficiency The team identified an NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, having very low safety significance involving inadequate cable design. Specifically, the team identified that the licensee failed to incorporate appropriate licensing and design basis requirements reflecting worst case environmental conditions for power and control safety related cables. Incorporation of these requirements would have ensured that the cables were designed for the continuous submerged conditions that are experienced at Clinton. The issue was entered into the licensees corrective action program to initiate a review of the current cable monitoring programs, and to initiate long-term corrective actions. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, because the licensee did not thoroughly evaluate problems such as the resolutions, address causes, and extent of condition (P.1 (c)). (Section 1R21.3.b.1)
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| Inspection Report# : 2007008 (pdf)
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| Significance: Dec 19, 2007 Identified By: NRC Item Type: NCV NonCited Violation Division 3 Emergency Diesel Generator Neutral Ground Resistor Design Inadequacy The team identified an NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, having very low safety significance involving inadequate equipment design. Specifically, the Division 3, emergency diesel generator (EDG) neutral ground resistor was found to be in a non-ventilated enclosure contrary to the USAR, which called for a ventilated housing. The issue was entered into the licensees corrective action program to address this non-conforming condition and develop a design change to enhance ventilation for the resistor. The team determined that there was no cross-cutting aspect to this finding. (Section 1R21.3.b.2)
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| Inspection Report# : 2007008 (pdf)
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| Significance: Dec 19, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Design of Emergency Diesel Generator Exhaust The team identified an NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, having very low safety significance involving inadequate design of the emergency diesel generator (EDG) exhaust sub-systems. Specifically, the licensee failed to properly account for severe weather in the design of the exhaust ducts for the EDGs.
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| Consequently, during severe weather conditions, icing or glazing could potentially result in blockage of the exhaust ducts screens located at the duct outlet and in exceeding the backpressure requirements of the ducts. Once identified, the licensee initiated a prompt operability evaluation to verify system operability and an Issue Report which included appropriate compensatory actions. The team determined that there was no cross-cutting aspect to this finding. (Section 1R21.3.b.3)
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| Inspection Report# : 2007008 (pdf)
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| Significance: Dec 19, 2007 Identified By: NRC Item Type: NCV NonCited Violation Residual Heat Removal Pipe Support Calculation Deficiencies The team identified an NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, having very low safety significance involving a temporary installation that added lead shielding to the Unit 1 residual heat removal (RHR) piping. Specifically, the team identified numerous non-conservative technical errors and calculation omissions in seismic design basis analysis calculations that supported this temporary installation. Once identified, the licensee initiated a prompt operability evaluation to verify system operability and an Issue Report which included appropriate
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| compensatory actions. The cause of the finding is related to the cross-cutting element of Human Performance Resources, because the licensee did not provide complete, accurate and up-to-date design documentation to assure nuclear safety (H.2(c)). Specifically, the licensee had the temporary installation of lead shielding in place since 2002 and did not formally update the associated pipe support calculations in a timely manner. (Section 1R21.3.b.4)
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| Inspection Report# : 2007008 (pdf)
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| Significance: Dec 19, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inappropriate SX Pump Test Acceptance Criteria The team identified an NCV of 10 CFR Part 50, Appendix B, Criterion XI, ATest Control,@ having very low safety significance, in that, the shutdown service water (SX) pump tests conducted did not appropriately demonstrate that the SX pumps met design basis accident requirements. Specifically, the pump test acceptance criteria allowed the pump performance to degrade below the performance assumed by the design analysis. Once identified, this finding was entered into the licensees corrective action program and the licensee completed an evaluation and retesting that demonstrated the pumps capacity to perform required safety functions. The team determined that there was no cross-cutting aspect to this finding. (Section 1R21.3.b.5)
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| Inspection Report# : 2007008 (pdf)
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| Barrier Integrity Significance: Mar 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO EVALUATE HYDRAULIC POWER UNIT PIPING FOR IMPACT WITH CONTAINMENT ATMOSPHERE MONITORING LINE.
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| The inspectors identified a finding and an associated NCV of 10 CFR Part 50, Appendix b, Criterion XVI, "Corrective Actions," having very low safety significance, in that, inevaluating whether the reactor recirculation flow control valve "A" hydraulic power unit (HPU) piping was adequately supported in response to concerns raised in two condition reports, the licensee did not adequately address that the as-build support configuration had not been properly verified from a design standpoint. In particular, the licensee did not consider the safety related classification of nearby containment/drywell atmosphere monitoring tubing and that this tubing could be impacged if the HPU piping failed during a postulated design basis seismic event. Hence, the licensee did not implement the additonal evaluation/calculations required to demonstrate the HPU piping met more stringent design requirements and was adequately supported. The primary cause of the violation was related to the cross-cutting component of Human Performance, Resources (H.2(c)) because the licensee failed to maintain complete, accurate, and up-to-date design documentation. Subsequently, the licensee performed evaluations/calculations demonstrating that the HPU piping will not adversely impact the safety related containment monitoring tubing during a design basis seismic event. The licensee entered the finding in the corrective action program as Action Request 723620.
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| The finding was more than minor because it was associated with the Barrier Integrity Cornerstone and affected the cornerstone objective of maintaining functionality of containment due to the potential impact on the safety related containment atmosphere monitoring system which was needed to monitor and to take actions to mitigate challenges to containment integrity. The finding was of very low safety significance because the licensee's preliminary results based on sonservative calculation indicated that the design basis requirements were met, and hence field modifications were not necessary.
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| Inspection Report# : 2008002 (pdf)
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| Significance: Sep 30, 2007 Identified By: NRC
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| Item Type: NCV NonCited Violation INADEQUATE PROCUREMENT SPECIFICATION FOR CHARCOAL RESULTS IN INOPERABLE CONTROL ROOM VENTILATION SUBSYSTEM.
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| A performance deficiency involving a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion IV, "Procurement Document Control," was self revealed following receipt of laboratory results that showed that Division 1 control room ventilation system charcoal filter penetration values were higher than allowed by Clinton's Technical Specifications. This issue occurred because the licensee failed to establish proper purchase specifications for charcoal used in the control room ventilation system. Additionally, this issue led to Division 1 control room ventilation subsystem being inoperable from May 9 through May 14, 2005. Licensee corrective actions included entering the issue into the corrective action program, revising the charcoal purchase specifications, and adding limitations to work orders to prevent scheduling work that could impact the operability of redundant systems.
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| This issue was more than minor because it affected the objective of the Barrier Integrity cornerstone of assuring that physical design barriers protect the public from radionuclide releases caused by accident or events. Additionally, this issue is associated with the barrier perfromance attribute of maintaining Radiological Barrier functionality of the control room. Failure to ensure adequate purchase specifications resulted in there being a period where both trains of control room ventilation were inoperable without the knowledge of the operators. The issue was of very low safety significance because it only represented a degradation of the radiological barrier function provided for the control room.
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| Inspection Report# : 2007004 (pdf)
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| Significance: Sep 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO COMPLY WITH TECHNICAL SPECIFICATION 3.4.5 FOR RCS PRESSURE BOUNDARY LEAK The inspectors identified a performance deficiency involving a Non-Cited Violation of Technical Specifications when the licensee failed to meet the required completion time for an action statement in Technical Specification 3.4.5.
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| Specifically, Technical Specification 3.4.5 does not allow reactor coolant system pressure boundary leakage and requires a shutdown to Mode 3 within 12 hours if pressure boundary leakage is discovered. Upon entry into the drywell following a shutdown of the reactor on June 19, 2007, the licensee discovered the existence of reactor coolant system pressure boundary leakage. Indications of the leakage had been discovered at 0433 on June 18, 2007, but the plant was not placed in Mode 3 until approximately 31 hours later at 1125 on June 19, 2007. Licensee corrective actions included replacing the leaking flexible hose, scheduling replacement of other flexible hoses, and establishing a preventive maintenance replacement frequency for the flexible hoses.
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| This issue was more than minor because oeprating with a degraded pressure boundary affected the reactor coolant system equipment and barrier performance attribute of the Barrier Integrity cornerstone, in that, reactor coolant system pressure boundary leakage results in a reduction in the reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The issue was of very low safety significance because the potential maximum size of the leak was well within the capability of the available mitigating equipment. The finding is related to the cross-cutting area of Human Performance (Decision Making) in the operators had initially entered TS 3.4.5 for pressure boundary leakage, but later chose not to treat the leakage as pressure boundary leakage, and treat it as unidentified leakage until the actual location could be determined.
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| Inspection Report# : 2007004 (pdf)
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| Significance: Sep 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE PROCEDURE RESULTS IN SPENT FUEL BUNDLE INCIDENT A performance deficiency involving a Non-Cited Violation of 10 CFR Part 50 Appendix B, Criteria V, "Instructions, Procedures, and Drawings," was self-revealed following an event on August 17, 2007, where a spent fuel bundle being moved to a temporary storage location came in contact with and rested upon another fuel bundle seated in its storage location. The licensee procedure that governs spent fuel pool movement failed to provide adequate guidance on how high to lift the fuel bundle prior to traversing across the spent fuel pool. Licensee corrective actions included revising the fuel handling procedure to provide specific instructions regarding how high to lift a fuel bundle during spent fuel pool movements.
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| This issue was more than minor because it affected the barrier integrity objective of assuring that physcial design barriers protect the public from radionuclide releases caused by accidents or events. The inspectors determined that this issue only degraded the Fuel Cladding Barrier and its associated cornerstone, therefore, this issue was of very low safety significance. This finding is related to the cross-cutting are of Human Performance (Resources) because the licensee did not provide complete and accurate procedures. Specifically, the procedure relied on the skills of the operator, did not provided specific values on how high to life a fuel bundle, and did not require independent verification Inspection Report# : 2007004 (pdf)
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| Emergency Preparedness Occupational Radiation Safety Significance: Mar 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO BARRICADE AND LOCK A LOCKED HIGH RADIATION ARE.
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| The inspectors identified a finding of very low safety significance and an associated NCV of Technical Specification 5.7.2 for failure to barricade, lock, or continuously guard a high radiation area with dose rates greater than 1000 millirem per hour. On January 24, 2008, licensee staff failed to properly barricade and lock or guard three entrances to the under vessel area of the drywell. As corrective actions, the licensee suspended access to the Radiologically controlled Area (RCA) for the personnel involved and initiated a prompt investigation, including assessment of the extent of conditon plant-wide. The licensee entered the issued into the corrective action program as IR 726499.
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| The finding was more than minor because it was associated with the Program/Process attribute of the Occupational Radiation Safety Cornerstone and affected the cornerstone objective to ensure worker health and safety from exposure to radiation, in that, failure to follow procedures for control of locked high radiation areas could result in unplannec exposure. the finding was determined to be of vey low safety significance because the finding did not involve As-Low-As-Is-Reasonably-Achievable (ALARA) planning, collective dose was not a factor, it did not involve an overexposure, there was not a substantial potential for a worker overexposure, and the licensee's ability to assess worker dose was not compromised. Additionally, this finding has a cross cutting aspect in the the area of Human Performance because radiation protection staff did not appropriately follow procedures (H.4(b)) which governed control of access into locked high radiation areas.
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| Inspection Report# : 2008002 (pdf)
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| Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.
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| Miscellaneous
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| Last modified : June 05, 2008 Clinton 2Q/2008 Plant Inspection Findings Initiating Events Significance: Mar 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW APPROVED FIRE PROTECTION PROGRAM PROCEDURES CONCERNING CONTROL OF TRANSIENT COMBUSTIBLE MATERIAL.
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| The inspectors identified a performance deficiency involving a Non-Cited Violation (NCV) of Clinton Power Station Operating License NPF-62, Section 2.F for failure to implement the fire protection program in accordance with program requirements. The inspectors identified multiple instances of the licensee's failure to follow approved fire protection program procedures concerning control of transient combustible material. Corrective actions for this issue included removing the unattended combustible material, initiating transient combustible permits, and/or initiating compensatory measures.
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| The inspectors determined that this issue was more than minor because the identified transient combustibles were in a combustible free zone required for separation of redundant trains. This finding was of very low safety significance because the transient combustible materials identified by the inspectors were not combustibles of significance. The inspectors determined that this finding was cross-cutting in the area of Problem Identification and Resolution.
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| Specifically, the licensee implements a corrective action program with a low treshold for identifying issues. The licensee identifies such issues completely, accurately, and in a timely manner commensurate with their safety significance (P.1(a)).
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| Inspection Report# : 2008002 (pdf)
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| Significance: Mar 31, 2008 Identified By: NRC Item Type: FIN Finding THE LICENSEE DISCOVERED THAT THE WRONG COMPONENT WAS INSTALLED IN THE B TURBINE DRIVEN REACTOR FEED PUMP OIL PRESSURE SENSING LOGIC.
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| A finding of very low safety significance was self-revealed by the automatic runback of the turbine driven reactor feed pump during post outage power ascension. The licensee discovered that the wrong component was installed in the B turbine driven reactor feed pump oil pressure sensing logic. The inspectors determined that the licensee failed to perform an adequate post maintenance test in accordance with procedures. This issue resulted in an unexpected power change from 54 percent power to 46 percent power. The licensee entered the issue into the corrective action program, performed tailgate discussions with technicians and work planners on the oil pressure switches were up to date in the materials and work management computer system.
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| The inspectors determined this issue was more than minor because it was associated with the Human Performance attribute of the Initiating Events Cornerstone and affected the cornerstone objective of limiting the frequency of those events that upset plant stability. Specifically, the failure to perform adequate post maintenance testing of pressure switch 1PS-FW 135 permitted the wrong component to be installed and placed in service. This deficiency ultimately resulted in an unplanned plant transient. The finding was of very low safety significance because this issue did not increase the likelihood that mitigation equipment or fundctions would not be available. The inspectors also concluded that the failure of the technician to properly follow calibration procedure 8801.01 during the initial calibration of this switch represented a cross-cutting issue in the area of Human Performance, Work Practices (H.4(b)), because licensee personnel failed to follow procedures in regard to pressure switch calibration.
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| Inspection Report# : 2008002 (pdf)
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| Significance: Mar 31, 2008 Identified By: NRC
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| Item Type: NCV NonCited Violation DURING THE PERFORMANCE OF NRC FINAL DRYWELL CLOSEOUT, THE INSPECTORS NOTED THAT FOREIGN MATERIAL/HOUSEKEEPING SOCK HAD NOT BEEN REMOVED FROM THE DRYWELL FLOOR DRAINS.
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| The inspectors identified a finding and an associated NCV of 10 CFR Part 50, Appendix B, Criteria V, "Instructions, Procedures, and Drawings," having a very low safety significance during drywell closeout inspections. Specifically, during the performance of the NRC final drywell closeout, the inspectors noted that foreigh material/houskeeping socks had not been removed from the drywell floor drains. This issue could have resulted in the drywell leak detection system being inoperable following a reactor event. The licensee proecdures for drywell closeout directed licensee staff to remove all loose material and devices associated with the licensee material condition and housekeeping program.
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| The licensee's corrective actions for this issue included removing the floor drain socks and incorporating a work activities item for sock removal in the outage schedule template.
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| The inspectors determined that this issue was more than minor because, if left uncorrected, it could result in a more significant safety concern. Failure to remove drain socks from drywell floor drains could result in the inability to readily detect and tract unidentified leakage following a reactor event. the finding was of very low safety significance because this finding didi not result in exceeding the Technical Specification limit for reactor coolant system (RCS) leakage nor did it affect other mitigating systems resulting in a ttal loss of their safety function. The inspectors also concluded this this issue was a result of no work tiem in the outage schedule to remove the socks, and therefore represented a cross-cutting issue in the are of Human Performance, Work Control (H.3(b)).
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| Inspection Report# : 2008002 (pdf)
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| Significance: Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement fire protection program in accordance with program requirements Identified a performance deficiency involving a NCV of Clinton Power Station Operating License NPF-62, Section 2.F for failure to implement fire protection program in accordance with program requirements. Inspectors identified multiple instances of the licensee failure to follow program procedures concerning control of Transient Combustible Material and Fire Protection Impairment Reporting. Corrective actions included removing the unattended combustible material and repairing latches on the fire doors.
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| This issue was more than minor because it could be a precursor to a significant event. A fire had potential of impacting safety related equipment used for safe shutdown purposes. This finding was of very low safety significance because the transient material identified by the inspectors were not combustibles of significance, and the licensee maintained fire suppression systems in the areas where the fire door latches were not functional. This finding was cross-cutting in the area of P.1(a) because the licensee failed to identify these issues in their corrective action program.
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| Inspection Report# : 2007005 (pdf)
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| Mitigating Systems Significance: Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PERFORM POST MODIFICATION TESTING TO SHUTDOWN SERVICE WATER VACUUM BREAKERS Identified a NCV of 10 CFR Part 50, App B, Crit XI, Test Controls, having a low safety significance for failure to properly test a permanent plant modification to the Div 1 & 2 SX. This resulted in two of four vacuum breakers that failed the minimal design specification during testing.
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| It was determined that the issue was more than minor because it is viewed as a precursor to a significant event. Failure to perform modification testing could lead to components within safety-related systems that do not work as designed.
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| Through detailed analysis the licensee concluded that the hydraulic experience with the vacuum breaker not meeting the minimal design specification would not make the shutdown service water system inoperable. This finding had a cross-cutting aspect in the area of H.4(c) because there were multiple opportunities for the licensee engineering staff to identify the need for this testing.
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| Inspection Report# : 2007005 (pdf)
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| Significance: Dec 19, 2007 Identified By: NRC Item Type: NCV NonCited Violation Continuously Submerged Cables Design Deficiency The team identified an NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, having very low safety significance involving inadequate cable design. Specifically, the team identified that the licensee failed to incorporate appropriate licensing and design basis requirements reflecting worst case environmental conditions for power and control safety related cables. Incorporation of these requirements would have ensured that the cables were designed for the continuous submerged conditions that are experienced at Clinton. The issue was entered into the licensees corrective action program to initiate a review of the current cable monitoring programs, and to initiate long-term corrective actions. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, because the licensee did not thoroughly evaluate problems such as the resolutions, address causes, and extent of condition (P.1 (c)). (Section 1R21.3.b.1)
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| Inspection Report# : 2007008 (pdf)
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| Significance: Dec 19, 2007 Identified By: NRC Item Type: NCV NonCited Violation Division 3 Emergency Diesel Generator Neutral Ground Resistor Design Inadequacy The team identified an NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, having very low safety significance involving inadequate equipment design. Specifically, the Division 3, emergency diesel generator (EDG) neutral ground resistor was found to be in a non-ventilated enclosure contrary to the USAR, which called for a ventilated housing. The issue was entered into the licensees corrective action program to address this non-conforming condition and develop a design change to enhance ventilation for the resistor. The team determined that there was no cross-cutting aspect to this finding. (Section 1R21.3.b.2)
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| Inspection Report# : 2007008 (pdf)
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| Significance: Dec 19, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Design of Emergency Diesel Generator Exhaust The team identified an NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, having very low safety significance involving inadequate design of the emergency diesel generator (EDG) exhaust sub-systems. Specifically, the licensee failed to properly account for severe weather in the design of the exhaust ducts for the EDGs.
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| Consequently, during severe weather conditions, icing or glazing could potentially result in blockage of the exhaust ducts screens located at the duct outlet and in exceeding the backpressure requirements of the ducts. Once identified, the licensee initiated a prompt operability evaluation to verify system operability and an Issue Report which included appropriate compensatory actions. The team determined that there was no cross-cutting aspect to this finding. (Section 1R21.3.b.3)
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| Inspection Report# : 2007008 (pdf)
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| Significance: Dec 19, 2007 Identified By: NRC Item Type: NCV NonCited Violation Residual Heat Removal Pipe Support Calculation Deficiencies The team identified an NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, having very low safety significance involving a temporary installation that added lead shielding to the Unit 1 residual heat removal (RHR) piping. Specifically, the team identified numerous non-conservative technical errors and calculation omissions in seismic design basis analysis calculations that supported this temporary installation. Once identified, the licensee initiated a prompt operability evaluation to verify system operability and an Issue Report which included appropriate
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| compensatory actions. The cause of the finding is related to the cross-cutting element of Human Performance Resources, because the licensee did not provide complete, accurate and up-to-date design documentation to assure nuclear safety (H.2(c)). Specifically, the licensee had the temporary installation of lead shielding in place since 2002 and did not formally update the associated pipe support calculations in a timely manner. (Section 1R21.3.b.4)
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| Inspection Report# : 2007008 (pdf)
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| Significance: Dec 19, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inappropriate SX Pump Test Acceptance Criteria The team identified an NCV of 10 CFR Part 50, Appendix B, Criterion XI, ATest Control,@ having very low safety significance, in that, the shutdown service water (SX) pump tests conducted did not appropriately demonstrate that the SX pumps met design basis accident requirements. Specifically, the pump test acceptance criteria allowed the pump performance to degrade below the performance assumed by the design analysis. Once identified, this finding was entered into the licensees corrective action program and the licensee completed an evaluation and retesting that demonstrated the pumps capacity to perform required safety functions. The team determined that there was no cross-cutting aspect to this finding. (Section 1R21.3.b.5)
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| Inspection Report# : 2007008 (pdf)
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| Barrier Integrity Significance: Mar 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO EVALUATE HYDRAULIC POWER UNIT PIPING FOR IMPACT WITH CONTAINMENT ATMOSPHERE MONITORING LINE.
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| The inspectors identified a finding and an associated NCV of 10 CFR Part 50, Appendix b, Criterion XVI, "Corrective Actions," having very low safety significance, in that, inevaluating whether the reactor recirculation flow control valve "A" hydraulic power unit (HPU) piping was adequately supported in response to concerns raised in two condition reports, the licensee did not adequately address that the as-build support configuration had not been properly verified from a design standpoint. In particular, the licensee did not consider the safety related classification of nearby containment/drywell atmosphere monitoring tubing and that this tubing could be impacged if the HPU piping failed during a postulated design basis seismic event. Hence, the licensee did not implement the additonal evaluation/calculations required to demonstrate the HPU piping met more stringent design requirements and was adequately supported. The primary cause of the violation was related to the cross-cutting component of Human Performance, Resources (H.2(c)) because the licensee failed to maintain complete, accurate, and up-to-date design documentation. Subsequently, the licensee performed evaluations/calculations demonstrating that the HPU piping will not adversely impact the safety related containment monitoring tubing during a design basis seismic event. The licensee entered the finding in the corrective action program as Action Request 723620.
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| The finding was more than minor because it was associated with the Barrier Integrity Cornerstone and affected the cornerstone objective of maintaining functionality of containment due to the potential impact on the safety related containment atmosphere monitoring system which was needed to monitor and to take actions to mitigate challenges to containment integrity. The finding was of very low safety significance because the licensee's preliminary results based on sonservative calculation indicated that the design basis requirements were met, and hence field modifications were not necessary.
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| Inspection Report# : 2008002 (pdf)
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| Significance: Sep 30, 2007 Identified By: NRC
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| Item Type: NCV NonCited Violation INADEQUATE PROCUREMENT SPECIFICATION FOR CHARCOAL RESULTS IN INOPERABLE CONTROL ROOM VENTILATION SUBSYSTEM.
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| A performance deficiency involving a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion IV, "Procurement Document Control," was self revealed following receipt of laboratory results that showed that Division 1 control room ventilation system charcoal filter penetration values were higher than allowed by Clinton's Technical Specifications. This issue occurred because the licensee failed to establish proper purchase specifications for charcoal used in the control room ventilation system. Additionally, this issue led to Division 1 control room ventilation subsystem being inoperable from May 9 through May 14, 2005. Licensee corrective actions included entering the issue into the corrective action program, revising the charcoal purchase specifications, and adding limitations to work orders to prevent scheduling work that could impact the operability of redundant systems.
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| This issue was more than minor because it affected the objective of the Barrier Integrity cornerstone of assuring that physical design barriers protect the public from radionuclide releases caused by accident or events. Additionally, this issue is associated with the barrier perfromance attribute of maintaining Radiological Barrier functionality of the control room. Failure to ensure adequate purchase specifications resulted in there being a period where both trains of control room ventilation were inoperable without the knowledge of the operators. The issue was of very low safety significance because it only represented a degradation of the radiological barrier function provided for the control room.
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| Inspection Report# : 2007004 (pdf)
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| Significance: Sep 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO COMPLY WITH TECHNICAL SPECIFICATION 3.4.5 FOR RCS PRESSURE BOUNDARY LEAK The inspectors identified a performance deficiency involving a Non-Cited Violation of Technical Specifications when the licensee failed to meet the required completion time for an action statement in Technical Specification 3.4.5.
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| Specifically, Technical Specification 3.4.5 does not allow reactor coolant system pressure boundary leakage and requires a shutdown to Mode 3 within 12 hours if pressure boundary leakage is discovered. Upon entry into the drywell following a shutdown of the reactor on June 19, 2007, the licensee discovered the existence of reactor coolant system pressure boundary leakage. Indications of the leakage had been discovered at 0433 on June 18, 2007, but the plant was not placed in Mode 3 until approximately 31 hours later at 1125 on June 19, 2007. Licensee corrective actions included replacing the leaking flexible hose, scheduling replacement of other flexible hoses, and establishing a preventive maintenance replacement frequency for the flexible hoses.
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| This issue was more than minor because oeprating with a degraded pressure boundary affected the reactor coolant system equipment and barrier performance attribute of the Barrier Integrity cornerstone, in that, reactor coolant system pressure boundary leakage results in a reduction in the reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The issue was of very low safety significance because the potential maximum size of the leak was well within the capability of the available mitigating equipment. The finding is related to the cross-cutting area of Human Performance (Decision Making) in the operators had initially entered TS 3.4.5 for pressure boundary leakage, but later chose not to treat the leakage as pressure boundary leakage, and treat it as unidentified leakage until the actual location could be determined.
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| Inspection Report# : 2007004 (pdf)
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| Significance: Sep 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE PROCEDURE RESULTS IN SPENT FUEL BUNDLE INCIDENT A performance deficiency involving a Non-Cited Violation of 10 CFR Part 50 Appendix B, Criteria V, "Instructions, Procedures, and Drawings," was self-revealed following an event on August 17, 2007, where a spent fuel bundle being moved to a temporary storage location came in contact with and rested upon another fuel bundle seated in its storage location. The licensee procedure that governs spent fuel pool movement failed to provide adequate guidance on how high to lift the fuel bundle prior to traversing across the spent fuel pool. Licensee corrective actions included revising the fuel handling procedure to provide specific instructions regarding how high to lift a fuel bundle during spent fuel pool movements.
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| This issue was more than minor because it affected the barrier integrity objective of assuring that physcial design barriers protect the public from radionuclide releases caused by accidents or events. The inspectors determined that this issue only degraded the Fuel Cladding Barrier and its associated cornerstone, therefore, this issue was of very low safety significance. This finding is related to the cross-cutting are of Human Performance (Resources) because the licensee did not provide complete and accurate procedures. Specifically, the procedure relied on the skills of the operator, did not provided specific values on how high to life a fuel bundle, and did not require independent verification Inspection Report# : 2007004 (pdf)
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| Emergency Preparedness Occupational Radiation Safety Significance: Mar 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO BARRICADE AND LOCK A LOCKED HIGH RADIATION ARE.
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| The inspectors identified a finding of very low safety significance and an associated NCV of Technical Specification 5.7.2 for failure to barricade, lock, or continuously guard a high radiation area with dose rates greater than 1000 millirem per hour. On January 24, 2008, licensee staff failed to properly barricade and lock or guard three entrances to the under vessel area of the drywell. As corrective actions, the licensee suspended access to the Radiologically controlled Area (RCA) for the personnel involved and initiated a prompt investigation, including assessment of the extent of conditon plant-wide. The licensee entered the issued into the corrective action program as IR 726499.
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| The finding was more than minor because it was associated with the Program/Process attribute of the Occupational Radiation Safety Cornerstone and affected the cornerstone objective to ensure worker health and safety from exposure to radiation, in that, failure to follow procedures for control of locked high radiation areas could result in unplannec exposure. the finding was determined to be of vey low safety significance because the finding did not involve As-Low-As-Is-Reasonably-Achievable (ALARA) planning, collective dose was not a factor, it did not involve an overexposure, there was not a substantial potential for a worker overexposure, and the licensee's ability to assess worker dose was not compromised. Additionally, this finding has a cross cutting aspect in the the area of Human Performance because radiation protection staff did not appropriately follow procedures (H.4(b)) which governed control of access into locked high radiation areas.
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| Inspection Report# : 2008002 (pdf)
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| Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.
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| Miscellaneous
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| Last modified : August 29, 2008 Clinton 3Q/2008 Plant Inspection Findings Initiating Events Significance: Sep 30, 2008 Identified By: Self-Revealing Item Type: FIN Finding FAILURE TO PERFORM ADEQUATE POST MAINTENANCE TESTING RESULTED IN HIGH REACTOR VESSEL WATER LEVEL (LEVEL 8) SCRAM The inspectors identified a finding of very low safety significance associated with a self-revealed event that resulted in a Unit 1 reactor scram.
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| The licensee failed to perform adequate post maintenance testing following replacement of the feedwater level control system dynamic compensator card during the Cycle 10 refueling outage that concluded in February 2006. This resulted in ineffective response from the feedwater level control system and a subsequent reactor scram following the unexpected loss of a reactor recirculation pump. The ineffective feedwater level control system response has not been corrected; however, the licensee entered this issue into its corrective action program for evaluation. No violation of regulatory requirements was identified.
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| The finding was of more than minor significance because this issue was associated with the Equipment Performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during power operations. Specifically, inadequate post maintenance testing resulted in ineffective response from the feedwater level control system during a loss of a reactor recirculation pump transient and caused a reactor scram. The finding was of very low safety significance because the issue: (1) did not contribute to the likelihood of a primary or secondary system loss-of-coolant-accident initiator, (2) did not contribute to both the likelihood of a reactor trip AND the likelihood that mitigation equipment or functions would not be available, and (3) did not increase the likelihood of a fire or internal/external flooding event. The inspectors did not identify a cross-cutting area component related to this finding.
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| Inspection Report# : 2008004 (pdf)
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| Significance: Sep 30, 2008 Identified By: Self-Revealing Item Type: FIN Finding FAILURE TO EVALUATE AN UNEXPECTED AND UNKNOWN CAUSE FOR STRAY VOLTAGE IN THE END-OF-CYCLE RECIRCULATION PUMP TRIP CIRCUIT DURING POST MODIFICATION TESTING RESULTED IN A REACTOR RECIRCULATION PUMP TRIP The inspectors identified a finding of very low safety significance associated with a self-revealed event that resulted in the unexpected loss of a reactor recirculation pump. The licensee failed to evaluate an unexpected and unknown cause for stray voltage in the End-of-Cycle Recirculation Pump Trip (EOC-RPT) circuit during post modification testing during the Cycle 11 refueling outage that concluded in February 2008. This resulted in the unexpected loss of a reactor recirculation pump and the subsequent plant transient that led to a reactor scram. As an immediate and interim corrective action, the licensee implemented a design change to the EOC-RPT circuitry that should prevent inadvertent relay actuation causing recirculation pumps trips due to the stray voltage problem. No violation of regulatory requirements was identified.
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| The finding was of more than minor significance because this issue was associated with the Equipment Performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during power operations. Specifically, the failure to evaluate an unexpected and unknown cause for stray voltage in the EOC-RPT circuit during post modification testing resulted in the unexpected loss of a reactor recirculation pump and the subsequent plant transient that led to a reactor scram. The finding was of very low safety significance because the issue: (1) did not contribute to the likelihood of a primary or secondary system loss-of-coolant-accident initiator, (2) did not contribute to both the likelihood of a reactor trip AND the likelihood that mitigation equipment for functions would not be available, and (3) did not increase the likelihood of a fire or internal/external flooding event. The inspectors concluded that this finding affected the cross-cutting area of human performance. Specifically, the licensee failed to appropriately incorporate risk insights in investigating and resolving an unexplained source of voltage in a circuit that had a high risk consequence (i.e., reactor recirculation pump trip). (IMC 0305 H.3(a))
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| Inspection Report# : 2008004 (pdf)
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| Significance: Mar 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW APPROVED FIRE PROTECTION PROGRAM PROCEDURES CONCERNING CONTROL OF TRANSIENT COMBUSTIBLE MATERIAL.
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| The inspectors identified a performance deficiency involving a Non-Cited Violation (NCV) of Clinton Power Station Operating License NPF-62, Section 2.F for failure to implement the fire protection program in accordance with program requirements. The inspectors identified
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| multiple instances of the licensee's failure to follow approved fire protection program procedures concerning control of transient combustible material. Corrective actions for this issue included removing the unattended combustible material, initiating transient combustible permits, and/or initiating compensatory measures.
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| The inspectors determined that this issue was more than minor because the identified transient combustibles were in a combustible free zone required for separation of redundant trains. This finding was of very low safety significance because the transient combustible materials identified by the inspectors were not combustibles of significance. The inspectors determined that this finding was cross-cutting in the area of Problem Identification and Resolution. Specifically, the licensee implements a corrective action program with a low treshold for identifying issues. The licensee identifies such issues completely, accurately, and in a timely manner commensurate with their safety significance (P.1 (a)).
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| Inspection Report# : 2008002 (pdf)
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| Significance: Mar 31, 2008 Identified By: NRC Item Type: FIN Finding THE LICENSEE DISCOVERED THAT THE WRONG COMPONENT WAS INSTALLED IN THE B TURBINE DRIVEN REACTOR FEED PUMP OIL PRESSURE SENSING LOGIC.
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| A finding of very low safety significance was self-revealed by the automatic runback of the turbine driven reactor feed pump during post outage power ascension. The licensee discovered that the wrong component was installed in the B turbine driven reactor feed pump oil pressure sensing logic. The inspectors determined that the licensee failed to perform an adequate post maintenance test in accordance with procedures. This issue resulted in an unexpected power change from 54 percent power to 46 percent power. The licensee entered the issue into the corrective action program, performed tailgate discussions with technicians and work planners on the oil pressure switches were up to date in the materials and work management computer system.
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| The inspectors determined this issue was more than minor because it was associated with the Human Performance attribute of the Initiating Events Cornerstone and affected the cornerstone objective of limiting the frequency of those events that upset plant stability. Specifically, the failure to perform adequate post maintenance testing of pressure switch 1PS-FW 135 permitted the wrong component to be installed and placed in service. This deficiency ultimately resulted in an unplanned plant transient. The finding was of very low safety significance because this issue did not increase the likelihood that mitigation equipment or fundctions would not be available. The inspectors also concluded that the failure of the technician to properly follow calibration procedure 8801.01 during the initial calibration of this switch represented a cross-cutting issue in the area of Human Performance, Work Practices (H.4(b)), because licensee personnel failed to follow procedures in regard to pressure switch calibration.
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| Inspection Report# : 2008002 (pdf)
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| Significance: Mar 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation DURING THE PERFORMANCE OF NRC FINAL DRYWELL CLOSEOUT, THE INSPECTORS NOTED THAT FOREIGN MATERIAL/HOUSEKEEPING SOCK HAD NOT BEEN REMOVED FROM THE DRYWELL FLOOR DRAINS.
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| The inspectors identified a finding and an associated NCV of 10 CFR Part 50, Appendix B, Criteria V, "Instructions, Procedures, and Drawings," having a very low safety significance during drywell closeout inspections. Specifically, during the performance of the NRC final drywell closeout, the inspectors noted that foreigh material/houskeeping socks had not been removed from the drywell floor drains. This issue could have resulted in the drywell leak detection system being inoperable following a reactor event. The licensee proecdures for drywell closeout directed licensee staff to remove all loose material and devices associated with the licensee material condition and housekeeping program. The licensee's corrective actions for this issue included removing the floor drain socks and incorporating a work activities item for sock removal in the outage schedule template.
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| The inspectors determined that this issue was more than minor because, if left uncorrected, it could result in a more significant safety concern.
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| Failure to remove drain socks from drywell floor drains could result in the inability to readily detect and track unidentified leakage following a reactor event. The finding was of very low safety significance because this finding did not result in exceeding the Technical Specification limit for reactor coolant system (RCS) leakage nor did it affect other mitigating systems resulting in a total loss of their safety function. The inspectors also concluded that this issue was a result of no work item in the outage schedule to remove the socks, and therefore represented a cross-cutting issue in the are of Human Performance, Work Control (H.3(b)).
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| Inspection Report# : 2008002 (pdf)
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| Significance: Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to implement fire protection program in accordance with program requirements Identified a performance deficiency involving a NCV of Clinton Power Station Operating License NPF-62, Section 2.F for failure to implement fire protection program in accordance with program requirements. Inspectors identified multiple instances of the licensee failure to follow program procedures concerning control of Transient Combustible Material and Fire Protection Impairment Reporting. Corrective actions included removing the unattended combustible material and repairing latches on the fire doors.
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| This issue was more than minor because it could be a precursor to a significant event. A fire had potential of impacting safety related equipment used for safe shutdown purposes. This finding was of very low safety significance because the transient material identified by the inspectors were not combustibles of significance, and the licensee maintained fire suppression systems in the areas where the fire door latches were not functional. This finding was cross-cutting in the area of P.1(a) because the licensee failed to identify these issues in their corrective action program.
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| Inspection Report# : 2007005 (pdf)
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| Mitigating Systems Significance: Sep 30, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation FAILURE TO PERFORM ADEQUATE PREVENTIVE MAINTENANCE ON SHUTDOWN SERVICE WATER VALVE 1SX014A RESULTED IN SIGNIFICANT DEGRADATION AND GROSS SEAT LEAKAGE A finding of very low safety significance with an associated Non-Cited Violation of Technical Specification (TS) 5.4.1.a was self-revealed.
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| The licensee failed to perform adequate preventive maintenance on shutdown service water system valve 1SX014A. This resulted in significant degradation of the valve body by corrosion due to prolonged exposure to raw service water that went undetected until gross seat leakage was discovered while attempting to establish conditions for surveillance testing. The licensee replaced the valve and has established a preventive maintenance schedule for internal valve inspections.
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| The finding would become a more significant safety convern if left uncorrected and was therefore more than a minor concern. Specifically, the failure to adequately perform preventive maintenance could reasonably result in significantly degraded or inoperable safety related equipment. Because the shutdown service water system was primarily associated with long term decay heat removal following certain design basis accidents, the inspectors concluded that this issue was associated with the Mitigating Systems cornerstone. The finding was of very low safety significance because the issue: (1) was not a design or qualification deficiency; (2) did not represent an actual loss of safety function of a system; (3) did not represent an actual loss of safety function of a single train for greater than its TS allowed outage time; (4) did not represent an actual loss of safety function of one or more non-TS trains of equipment designated as risk significant; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors concluded that this finding affected the cross-cutting area of human performance. Specifically, the licensee's investigation determined that internal valve inspections were not performed because the component category was incorrectly classified. (IMC 0305 H.3(b))
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| Inspection Report# : 2008004 (pdf)
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| Significance: Sep 30, 2008 Identified By: NRC Item Type: FIN Finding FAILURE TO RECOGNIZE THE SAFETY RELATED SYSTEM FUNCTION OF THE 1B RESIDUAL HEAT REMOVAL PUMP SEAL COOLER WHEN EVALUATING PAST OPERABILITY OF THE PUMP.
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| The inspectors identified a finding of very low safety significance associated with the licensee's failure to recognize the safety related system function of the 1B residual heat removal pump seal cooler when initially evaluating the past operability of the pump after unacceptable results were obtained during service water system flow balance testing. No analysis was performed to ensure that the pump's safety function would be fulfilled with less than minimum design flow to the cooler until the inspectors challenged the licensee's original conclusion. The licensee re-performed the past operability evaluation and determined that sufficient margin existed such that the pump would have been able to fulfill its safety function with significantly less than design flow to the seal cooler as measured during the test. No violation of regulatory requirements was identified.
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| The finding would become a more significant safety concern if left uncorrected and was therefore more than a minor concern. Specifically, the failure to correctly recognize the safety related functions of systems or components when performing operability or past operability evaluations could reasonably result in an unrecognized condition of a system failing to fulfill its safety related function. In addition, based on review of examples of minor issues in Inspection Manual Chapter 0612, "Power Reactor Inspection Reports," Appendix E, "Examples of Minor Issues," evaluation errors resulting in a reasonable doubt about the operability of a system or component are generally not considered to be of minor significance. Because the residual heat removal system was primarily associated with long term decay heat removal following certain design basis accidents, the inspectors condcluded that this issue was associated with the Mitigating Systems cornerstone. The finding was of very low safety significance because the issue: (1) was not a design or qualification deficienty; (2) did not represent an actual loss of safety function of a system; (3) did not represent an actual loss of safety function of a single train for greater than its Technical Specification (TS) allowed outage time; (4) did not represent an actual loss of safety function of one or more non-TS trains of equipment designated as risk signficiant; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. Subsequent evaluation was able to determine that sufficient margin in flow existed for the time period in question. The inspectors did not identify a corss-cutting area component related to this finding.
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| Inspection Report# : 2008004 (pdf)
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| Significance: Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PERFORM POST MODIFICATION TESTING TO SHUTDOWN SERVICE WATER VACUUM BREAKERS Identified a NCV of 10 CFR Part 50, App B, Crit XI, Test Controls, having a low safety significance for failure to properly test a permanent plant modification to the Div 1 & 2 SX. This resulted in two of four vacuum breakers that failed the minimal design specification during testing.
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| It was determined that the issue was more than minor because it is viewed as a precursor to a significant event. Failure to perform modification testing could lead to components within safety-related systems that do not work as designed. Through detailed analysis the licensee concluded that the hydraulic experience with the vacuum breaker not meeting the minimal design specification would not make the shutdown service water system inoperable. This finding had a cross-cutting aspect in the area of H.4(c) because there were multiple opportunities for the licensee engineering staff to identify the need for this testing.
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| Inspection Report# : 2007005 (pdf)
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| Significance: Dec 19, 2007 Identified By: NRC Item Type: NCV NonCited Violation Continuously Submerged Cables Design Deficiency The team identified an NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, having very low safety significance involving inadequate cable design. Specifically, the team identified that the licensee failed to incorporate appropriate licensing and design basis requirements reflecting worst case environmental conditions for power and control safety related cables. Incorporation of these requirements would have ensured that the cables were designed for the continuous submerged conditions that are experienced at Clinton. The issue was entered into the licensees corrective action program to initiate a review of the current cable monitoring programs, and to initiate long-term corrective actions. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, because the licensee did not thoroughly evaluate problems such as the resolutions, address causes, and extent of condition (P.1 (c)).
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| Inspection Report# : 2007008 (pdf)
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| Significance: Dec 19, 2007 Identified By: NRC Item Type: NCV NonCited Violation Division 3 Emergency Diesel Generator Neutral Ground Resistor Design Inadequacy The team identified an NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, having very low safety significance involving inadequate equipment design. Specifically, the Division 3, emergency diesel generator (EDG) neutral ground resistor was found to be in a non-ventilated enclosure contrary to the USAR, which called for a ventilated housing. The issue was entered into the licensees corrective action program to address this non-conforming condition and develop a design change to enhance ventilation for the resistor. The team determined that there was no cross-cutting aspect to this finding.
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| Inspection Report# : 2007008 (pdf)
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| Significance: Dec 19, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Design of Emergency Diesel Generator Exhaust The team identified an NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, having very low safety significance involving inadequate design of the emergency diesel generator (EDG) exhaust sub-systems. Specifically, the licensee failed to properly account for severe weather in the design of the exhaust ducts for the EDGs. Consequently, during severe weather conditions, icing or glazing could potentially result in blockage of the exhaust ducts screens located at the duct outlet and in exceeding the backpressure requirements of the ducts. Once identified, the licensee initiated a prompt operability evaluation to verify system operability and an Issue Report which included appropriate compensatory actions. The team determined that there was no cross-cutting aspect to this finding.
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| Inspection Report# : 2007008 (pdf)
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| Significance: Dec 19, 2007 Identified By: NRC Item Type: NCV NonCited Violation Residual Heat Removal Pipe Support Calculation Deficiencies The team identified an NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, having very low safety significance involving a temporary installation that added lead shielding to the Unit 1 residual heat removal (RHR) piping. Specifically, the team identified numerous non-conservative technical errors and calculation omissions in seismic design basis analysis calculations that supported this temporary
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| installation. Once identified, the licensee initiated a prompt operability evaluation to verify system operability and an Issue Report which included appropriate compensatory actions. The cause of the finding is related to the cross-cutting element of Human Performance Resources, because the licensee did not provide complete, accurate and up-to-date design documentation to assure nuclear safety (H.2(c)). Specifically, the licensee had the temporary installation of lead shielding in place since 2002 and did not formally update the associated pipe support calculations in a timely manner.
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| Inspection Report# : 2007008 (pdf)
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| Significance: Dec 19, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inappropriate SX Pump Test Acceptance Criteria The team identified an NCV of 10 CFR Part 50, Appendix B, Criterion XI, "Test Control," having very low safety significance, in that, the shutdown service water (SX) pump tests conducted did not appropriately demonstrate that the SX pumps met design basis accident requirements. Specifically, the pump test acceptance criteria allowed the pump performance to degrade below the performance assumed by the design analysis. Once identified, this finding was entered into the licensees corrective action program and the licensee completed an evaluation and retesting that demonstrated the pumps capacity to perform required safety functions. The team determined that there was no cross-cutting aspect to this finding.
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| Inspection Report# : 2007008 (pdf)
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| Barrier Integrity Significance: Mar 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO EVALUATE HYDRAULIC POWER UNIT PIPING FOR IMPACT WITH CONTAINMENT ATMOSPHERE MONITORING LINE.
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| The inspectors identified a finding and an associated NCV of 10 CFR Part 50, Appendix b, Criterion XVI, "Corrective Actions," having very low safety significance, in that, inevaluating whether the reactor recirculation flow control valve "A" hydraulic power unit (HPU) piping was adequately supported in response to concerns raised in two condition reports, the licensee did not adequately address that the as-build support configuration had not been properly verified from a design standpoint. In particular, the licensee did not consider the safety related classification of nearby containment/drywell atmosphere monitoring tubing and that this tubing could be impacged if the HPU piping failed during a postulated design basis seismic event. Hence, the licensee did not implement the additonal evaluation/calculations required to demonstrate the HPU piping met more stringent design requirements and was adequately supported. The primary cause of the violation was related to the cross-cutting component of Human Performance, Resources (H.2(c)) because the licensee failed to maintain complete, accurate, and up-to-date design documentation. Subsequently, the licensee performed evaluations/calculations demonstrating that the HPU piping will not adversely impact the safety related containment monitoring tubing during a design basis seismic event. The licensee entered the finding in the corrective action program as Action Request 723620.
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| The finding was more than minor because it was associated with the Barrier Integrity Cornerstone and affected the cornerstone objective of maintaining functionality of containment due to the potential impact on the safety related containment atmosphere monitoring system which was needed to monitor and to take actions to mitigate challenges to containment integrity. The finding was of very low safety significance because the licensee's preliminary results based on sonservative calculation indicated that the design basis requirements were met, and hence field modifications were not necessary.
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| Inspection Report# : 2008002 (pdf)
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| Emergency Preparedness Occupational Radiation Safety Significance: Mar 31, 2008 Identified By: NRC
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| Item Type: NCV NonCited Violation FAILURE TO BARRICADE AND LOCK A LOCKED HIGH RADIATION AREA.
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| The inspectors identified a finding of very low safety significance and an associated NCV of Technical Specification 5.7.2 for failure to barricade, lock, or continuously guard a high radiation area with dose rates greater than 1000 millirem per hour. On January 24, 2008, licensee staff failed to properly barricade and lock or guard three entrances to the under vessel area of the drywell. As corrective actions, the licensee suspended access to the Radiologically controlled Area (RCA) for the personnel involved and initiated a prompt investigation, including assessment of the extent of conditon plant-wide. The licensee entered the issued into the corrective action program as IR 726499.
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| The finding was more than minor because it was associated with the Program/Process attribute of the Occupational Radiation Safety Cornerstone and affected the cornerstone objective to ensure worker health and safety from exposure to radiation, in that, failure to follow procedures for control of locked high radiation areas could result in unplannec exposure. the finding was determined to be of vey low safety significance because the finding did not involve As-Low-As-Is-Reasonably-Achievable (ALARA) planning, collective dose was not a factor, it did not involve an overexposure, there was not a substantial potential for a worker overexposure, and the licensee's ability to assess worker dose was not compromised. Additionally, this finding has a cross cutting aspect in the the area of Human Performance because radiation protection staff did not appropriately follow procedures (H.4(b)) which governed control of access into locked high radiation areas.
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| Inspection Report# : 2008002 (pdf)
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| Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.
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| Miscellaneous Last modified : November 26, 2008
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| Clinton 4Q/2008 Plant Inspection Findings Initiating Events Significance: Dec 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO CONTROL TRANSIENT COMBUSTIBLE MATERIALS IN ACCORDANCE WITH FIRE PROTECTION PROGRAM.
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| GREEN. The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of the Clinton Power Station Unit 1 Operating License (NPF-62, Section 2.F). The licensee failed to implement the fire protection program in accordance with program requirements by failing to follow approved fire protection program procedures for the control of transient combustible matrials. The licensee promptly removed transient combustible materials found by the inspectors and subsequently completed a detailed walk down of the plant's transient combustible free zones to identify and remove any additional transient combustible materials.
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| The inspectors concluded that this finding could be reasonably viewed as a precursor to a significant event (i.e., a fire affecting more than one train of safe shutdown equipment). Specifically, the presence of transient combustible materials in a combustible free zone could reasonably result in degradation of the fire protection defense-in-depth elements in place to prevent fires from starting and mitigate the consequences of fires. In addition, based on review of Example 4k in Inspection Manual Chapter (IMC) 0612, "Power Reactor Inspection Reports," Appendix E, "Examples of Minor Issues," the issue would not be considered to be of minor significance because the identified transient combustibles were found in a combustible free zone required for separation of redundant trains. The finding was of very low safety significance because the items found in the combustible free zone would not be considered transient combustibles of significance as defined in IMC 0609, Appendix F, "Fire Protection Significance Determination Process," Attachment 2, "Degradation Rating Guidance Specific to Various Fire Protection Program Elements," and therefore the issue was assigned a "low degradation" rating. The inspectors concluded that this finding affected the cross-cutting area of problem identification and resolution. Specifically, the licensee missed an opportunity to identify and remove the transient combustible materials while implementing corrective actions for previously inspector identified findings involving the control of transient combustible materials. (IMC 0305 P.1 (a))
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| Inspection Report# : 2008005 (pdf)
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| Significance: Sep 30, 2008 Identified By: Self-Revealing Item Type: FIN Finding FAILURE TO PERFORM ADEQUATE POST MAINTENANCE TESTING RESULTED IN HIGH REACTOR VESSEL WATER LEVEL (LEVEL 8) SCRAM The inspectors identified a finding of very low safety significance associated with a self-revealed event that resulted in a Unit 1 reactor scram. The licensee failed to perform adequate post maintenance testing following replacement of the feedwater level control system dynamic compensator card during the Cycle 10 refueling outage that concluded in February 2006. This resulted in ineffective response from the feedwater level control system and a subsequent reactor scram following the unexpected loss of a reactor recirculation pump. The ineffective feedwater level control system response has not been corrected; however, the licensee entered this issue into its corrective action program for evaluation. No violation of regulatory requirements was identified.
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| The finding was of more than minor significance because this issue was associated with the Equipment Performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during power operations.
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| Specifically, inadequate post maintenance testing resulted in ineffective response from the feedwater level control system during a loss of a reactor recirculation pump transient and caused a reactor scram. The finding was of very low
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| safety significance because the issue: (1) did not contribute to the likelihood of a primary or secondary system loss-of-coolant-accident initiator, (2) did not contribute to both the likelihood of a reactor trip AND the likelihood that mitigation equipment or functions would not be available, and (3) did not increase the likelihood of a fire or internal/external flooding event. The inspectors did not identify a cross-cutting area component related to this finding.
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| Inspection Report# : 2008004 (pdf)
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| Significance: Sep 30, 2008 Identified By: Self-Revealing Item Type: FIN Finding FAILURE TO EVALUATE AN UNEXPECTED AND UNKNOWN CAUSE FOR STRAY VOLTAGE IN THE END-OF-CYCLE RECIRCULATION PUMP TRIP CIRCUIT DURING POST MODIFICATION TESTING RESULTED IN A REACTOR RECIRCULATION PUMP TRIP The inspectors identified a finding of very low safety significance associated with a self-revealed event that resulted in the unexpected loss of a reactor recirculation pump. The licensee failed to evaluate an unexpected and unknown cause for stray voltage in the End-of-Cycle Recirculation Pump Trip (EOC-RPT) circuit during post modification testing during the Cycle 11 refueling outage that concluded in February 2008. This resulted in the unexpected loss of a reactor recirculation pump and the subsequent plant transient that led to a reactor scram. As an immediate and interim corrective action, the licensee implemented a design change to the EOC-RPT circuitry that should prevent inadvertent relay actuation causing recirculation pumps trips due to the stray voltage problem. No violation of regulatory requirements was identified.
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| The finding was of more than minor significance because this issue was associated with the Equipment Performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during power operations.
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| Specifically, the failure to evaluate an unexpected and unknown cause for stray voltage in the EOC-RPT circuit during post modification testing resulted in the unexpected loss of a reactor recirculation pump and the subsequent plant transient that led to a reactor scram. The finding was of very low safety significance because the issue: (1) did not contribute to the likelihood of a primary or secondary system loss-of-coolant-accident initiator, (2) did not contribute to both the likelihood of a reactor trip AND the likelihood that mitigation equipment for functions would not be available, and (3) did not increase the likelihood of a fire or internal/external flooding event. The inspectors concluded that this finding affected the cross-cutting area of human performance. Specifically, the licensee failed to appropriately incorporate risk insights in investigating and resolving an unexplained source of voltage in a circuit that had a high risk consequence (i.e., reactor recirculation pump trip). (IMC 0305 H.3(a))
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| Inspection Report# : 2008004 (pdf)
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| Significance: Mar 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW APPROVED FIRE PROTECTION PROGRAM PROCEDURES CONCERNING CONTROL OF TRANSIENT COMBUSTIBLE MATERIAL.
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| The inspectors identified a performance deficiency involving a Non-Cited Violation (NCV) of Clinton Power Station Operating License NPF-62, Section 2.F for failure to implement the fire protection program in accordance with program requirements. The inspectors identified multiple instances of the licensee's failure to follow approved fire protection program procedures concerning control of transient combustible material. Corrective actions for this issue included removing the unattended combustible material, initiating transient combustible permits, and/or initiating compensatory measures.
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| The inspectors determined that this issue was more than minor because the identified transient combustibles were in a combustible free zone required for separation of redundant trains. This finding was of very low safety significance because the transient combustible materials identified by the inspectors were not combustibles of significance. The inspectors determined that this finding was cross-cutting in the area of Problem Identification and Resolution.
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| Specifically, the licensee implements a corrective action program with a low treshold for identifying issues. The licensee identifies such issues completely, accurately, and in a timely manner commensurate with their safety significance (P.1(a)).
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| Inspection Report# : 2008002 (pdf)
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| Significance: Mar 31, 2008 Identified By: NRC Item Type: FIN Finding THE LICENSEE DISCOVERED THAT THE WRONG COMPONENT WAS INSTALLED IN THE B TURBINE DRIVEN REACTOR FEED PUMP OIL PRESSURE SENSING LOGIC.
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| A finding of very low safety significance was self-revealed by the automatic runback of the turbine driven reactor feed pump during post outage power ascension. The licensee discovered that the wrong component was installed in the B turbine driven reactor feed pump oil pressure sensing logic. The inspectors determined that the licensee failed to perform an adequate post maintenance test in accordance with procedures. This issue resulted in an unexpected power change from 54 percent power to 46 percent power. The licensee entered the issue into the corrective action program, performed tailgate discussions with technicians and work planners on the oil pressure switches were up to date in the materials and work management computer system.
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| The inspectors determined this issue was more than minor because it was associated with the Human Performance attribute of the Initiating Events Cornerstone and affected the cornerstone objective of limiting the frequency of those events that upset plant stability. Specifically, the failure to perform adequate post maintenance testing of pressure switch 1PS-FW 135 permitted the wrong component to be installed and placed in service. This deficiency ultimately resulted in an unplanned plant transient. The finding was of very low safety significance because this issue did not increase the likelihood that mitigation equipment or fundctions would not be available. The inspectors also concluded that the failure of the technician to properly follow calibration procedure 8801.01 during the initial calibration of this switch represented a cross-cutting issue in the area of Human Performance, Work Practices (H.4(b)), because licensee personnel failed to follow procedures in regard to pressure switch calibration.
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| Inspection Report# : 2008002 (pdf)
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| Significance: Mar 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation DURING THE PERFORMANCE OF NRC FINAL DRYWELL CLOSEOUT, THE INSPECTORS NOTED THAT FOREIGN MATERIAL/HOUSEKEEPING SOCK HAD NOT BEEN REMOVED FROM THE DRYWELL FLOOR DRAINS.
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| The inspectors identified a finding and an associated NCV of 10 CFR Part 50, Appendix B, Criteria V, "Instructions, Procedures, and Drawings," having a very low safety significance during drywell closeout inspections. Specifically, during the performance of the NRC final drywell closeout, the inspectors noted that foreigh material/houskeeping socks had not been removed from the drywell floor drains. This issue could have resulted in the drywell leak detection system being inoperable following a reactor event. The licensee proecdures for drywell closeout directed licensee staff to remove all loose material and devices associated with the licensee material condition and housekeeping program.
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| The licensee's corrective actions for this issue included removing the floor drain socks and incorporating a work activities item for sock removal in the outage schedule template.
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| The inspectors determined that this issue was more than minor because, if left uncorrected, it could result in a more significant safety concern. Failure to remove drain socks from drywell floor drains could result in the inability to readily detect and track unidentified leakage following a reactor event. The finding was of very low safety significance because this finding did not result in exceeding the Technical Specification limit for reactor coolant system (RCS) leakage nor did it affect other mitigating systems resulting in a total loss of their safety function. The inspectors also concluded that this issue was a result of no work item in the outage schedule to remove the socks, and therefore represented a cross-cutting issue in the are of Human Performance, Work Control (H.3(b)).
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| Inspection Report# : 2008002 (pdf)
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| Mitigating Systems
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| Significance: Sep 30, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation FAILURE TO PERFORM ADEQUATE PREVENTIVE MAINTENANCE ON SHUTDOWN SERVICE WATER VALVE 1SX014A RESULTED IN SIGNIFICANT DEGRADATION AND GROSS SEAT LEAKAGE A finding of very low safety significance with an associated Non-Cited Violation of Technical Specification (TS) 5.4.1.a was self-revealed. The licensee failed to perform adequate preventive maintenance on shutdown service water system valve 1SX014A. This resulted in significant degradation of the valve body by corrosion due to prolonged exposure to raw service water that went undetected until gross seat leakage was discovered while attempting to establish conditions for surveillance testing. The licensee replaced the valve and has established a preventive maintenance schedule for internal valve inspections.
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| The finding would become a more significant safety convern if left uncorrected and was therefore more than a minor concern. Specifically, the failure to adequately perform preventive maintenance could reasonably result in significantly degraded or inoperable safety related equipment. Because the shutdown service water system was primarily associated with long term decay heat removal following certain design basis accidents, the inspectors concluded that this issue was associated with the Mitigating Systems cornerstone. The finding was of very low safety significance because the issue: (1) was not a design or qualification deficiency; (2) did not represent an actual loss of safety function of a system; (3) did not represent an actual loss of safety function of a single train for greater than its TS allowed outage time; (4) did not represent an actual loss of safety function of one or more non-TS trains of equipment designated as risk significant; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors concluded that this finding affected the cross-cutting area of human performance. Specifically, the licensee's investigation determined that internal valve inspections were not performed because the component category was incorrectly classified. (IMC 0305 H.3(b))
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| Inspection Report# : 2008004 (pdf)
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| Significance: Sep 30, 2008 Identified By: NRC Item Type: FIN Finding FAILURE TO RECOGNIZE THE SAFETY RELATED SYSTEM FUNCTION OF THE 1B RESIDUAL HEAT REMOVAL PUMP SEAL COOLER WHEN EVALUATING PAST OPERABILITY OF THE PUMP.
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| The inspectors identified a finding of very low safety significance associated with the licensee's failure to recognize the safety related system function of the 1B residual heat removal pump seal cooler when initially evaluating the past operability of the pump after unacceptable results were obtained during service water system flow balance testing. No analysis was performed to ensure that the pump's safety function would be fulfilled with less than minimum design flow to the cooler until the inspectors challenged the licensee's original conclusion. The licensee re-performed the past operability evaluation and determined that sufficient margin existed such that the pump would have been able to fulfill its safety function with significantly less than design flow to the seal cooler as measured during the test. No violation of regulatory requirements was identified.
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| The finding would become a more significant safety concern if left uncorrected and was therefore more than a minor concern. Specifically, the failure to correctly recognize the safety related functions of systems or components when performing operability or past operability evaluations could reasonably result in an unrecognized condition of a system failing to fulfill its safety related function. In addition, based on review of examples of minor issues in Inspection Manual Chapter 0612, "Power Reactor Inspection Reports," Appendix E, "Examples of Minor Issues,"
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| evaluation errors resulting in a reasonable doubt about the operability of a system or component are generally not considered to be of minor significance. Because the residual heat removal system was primarily associated with long term decay heat removal following certain design basis accidents, the inspectors condcluded that this issue was associated with the Mitigating Systems cornerstone. The finding was of very low safety significance because the issue: (1) was not a design or qualification deficienty; (2) did not represent an actual loss of safety function of a system; (3) did not represent an actual loss of safety function of a single train for greater than its Technical Specification (TS) allowed outage time; (4) did not represent an actual loss of safety function of one or more non-TS trains of equipment designated as risk signficiant; and (5) did not screen as potentially risk significant due to a
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| seismic, flooding, or severe weather initiating event. Subsequent evaluation was able to determine that sufficient margin in flow existed for the time period in question. The inspectors did not identify a corss-cutting area component related to this finding.
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| Inspection Report# : 2008004 (pdf)
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| Barrier Integrity Significance: Mar 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO EVALUATE HYDRAULIC POWER UNIT PIPING FOR IMPACT WITH CONTAINMENT ATMOSPHERE MONITORING LINE.
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| The inspectors identified a finding and an associated NCV of 10 CFR Part 50, Appendix b, Criterion XVI, "Corrective Actions," having very low safety significance, in that, inevaluating whether the reactor recirculation flow control valve "A" hydraulic power unit (HPU) piping was adequately supported in response to concerns raised in two condition reports, the licensee did not adequately address that the as-build support configuration had not been properly verified from a design standpoint. In particular, the licensee did not consider the safety related classification of nearby containment/drywell atmosphere monitoring tubing and that this tubing could be impacged if the HPU piping failed during a postulated design basis seismic event. Hence, the licensee did not implement the additonal evaluation/calculations required to demonstrate the HPU piping met more stringent design requirements and was adequately supported. The primary cause of the violation was related to the cross-cutting component of Human Performance, Resources (H.2(c)) because the licensee failed to maintain complete, accurate, and up-to-date design documentation. Subsequently, the licensee performed evaluations/calculations demonstrating that the HPU piping will not adversely impact the safety related containment monitoring tubing during a design basis seismic event. The licensee entered the finding in the corrective action program as Action Request 723620.
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| The finding was more than minor because it was associated with the Barrier Integrity Cornerstone and affected the cornerstone objective of maintaining functionality of containment due to the potential impact on the safety related containment atmosphere monitoring system which was needed to monitor and to take actions to mitigate challenges to containment integrity. The finding was of very low safety significance because the licensee's preliminary results based on sonservative calculation indicated that the design basis requirements were met, and hence field modifications were not necessary.
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| Inspection Report# : 2008002 (pdf)
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| Emergency Preparedness Occupational Radiation Safety Significance: Mar 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO BARRICADE AND LOCK A LOCKED HIGH RADIATION AREA.
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| The inspectors identified a finding of very low safety significance and an associated NCV of Technical Specification 5.7.2 for failure to barricade, lock, or continuously guard a high radiation area with dose rates greater than 1000 millirem per hour. On January 24, 2008, licensee staff failed to properly barricade and lock or guard three entrances to the under vessel area of the drywell. As corrective actions, the licensee suspended access to the Radiologically controlled Area (RCA) for the personnel involved and initiated a prompt investigation, including assessment of the
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| extent of conditon plant-wide. The licensee entered the issued into the corrective action program as IR 726499.
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| The finding was more than minor because it was associated with the Program/Process attribute of the Occupational Radiation Safety Cornerstone and affected the cornerstone objective to ensure worker health and safety from exposure to radiation, in that, failure to follow procedures for control of locked high radiation areas could result in unplannec exposure. the finding was determined to be of vey low safety significance because the finding did not involve As-Low-As-Is-Reasonably-Achievable (ALARA) planning, collective dose was not a factor, it did not involve an overexposure, there was not a substantial potential for a worker overexposure, and the licensee's ability to assess worker dose was not compromised. Additionally, this finding has a cross cutting aspect in the the area of Human Performance because radiation protection staff did not appropriately follow procedures (H.4(b)) which governed control of access into locked high radiation areas.
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| Inspection Report# : 2008002 (pdf)
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| Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.
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| Miscellaneous Last modified : April 07, 2009
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| Clinton 1Q/2009 Plant Inspection Findings Initiating Events Significance: Dec 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO CONTROL TRANSIENT COMBUSTIBLE MATERIALS IN ACCORDANCE WITH FIRE PROTECTION PROGRAM.
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| GREEN. The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of the Clinton Power Station Unit 1 Operating License (NPF-62, Section 2.F). The licensee failed to implement the fire protection program in accordance with program requirements by failing to follow approved fire protection program procedures for the control of transient combustible matrials. The licensee promptly removed transient combustible materials found by the inspectors and subsequently completed a detailed walk down of the plant's transient combustible free zones to identify and remove any additional transient combustible materials.
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| The inspectors concluded that this finding could be reasonably viewed as a precursor to a significant event (i.e., a fire affecting more than one train of safe shutdown equipment). Specifically, the presence of transient combustible materials in a combustible free zone could reasonably result in degradation of the fire protection defense-in-depth elements in place to prevent fires from starting and mitigate the consequences of fires. In addition, based on review of Example 4k in Inspection Manual Chapter (IMC) 0612, "Power Reactor Inspection Reports," Appendix E, "Examples of Minor Issues," the issue would not be considered to be of minor significance because the identified transient combustibles were found in a combustible free zone required for separation of redundant trains. The finding was of very low safety significance because the items found in the combustible free zone would not be considered transient combustibles of significance as defined in IMC 0609, Appendix F, "Fire Protection Significance Determination Process," Attachment 2, "Degradation Rating Guidance Specific to Various Fire Protection Program Elements," and therefore the issue was assigned a "low degradation" rating. The inspectors concluded that this finding affected the cross-cutting area of problem identification and resolution. Specifically, the licensee missed an opportunity to identify and remove the transient combustible materials while implementing corrective actions for previously inspector identified findings involving the control of transient combustible materials. (IMC 0305 P.1 (a))
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| Inspection Report# : 2008005 (pdf)
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| Significance: Sep 30, 2008 Identified By: Self-Revealing Item Type: FIN Finding FAILURE TO PERFORM ADEQUATE POST MAINTENANCE TESTING RESULTED IN HIGH REACTOR VESSEL WATER LEVEL (LEVEL 8) SCRAM The inspectors identified a finding of very low safety significance associated with a self-revealed event that resulted in a Unit 1 reactor scram. The licensee failed to perform adequate post maintenance testing following replacement of the feedwater level control system dynamic compensator card during the Cycle 10 refueling outage that concluded in February 2006. This resulted in ineffective response from the feedwater level control system and a subsequent reactor scram following the unexpected loss of a reactor recirculation pump. The ineffective feedwater level control system response has not been corrected; however, the licensee entered this issue into its corrective action program for evaluation. No violation of regulatory requirements was identified.
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| The finding was of more than minor significance because this issue was associated with the Equipment Performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during power operations.
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| Specifically, inadequate post maintenance testing resulted in ineffective response from the feedwater level control system during a loss of a reactor recirculation pump transient and caused a reactor scram. The finding was of very low
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| safety significance because the issue: (1) did not contribute to the likelihood of a primary or secondary system loss-of-coolant-accident initiator, (2) did not contribute to both the likelihood of a reactor trip AND the likelihood that mitigation equipment or functions would not be available, and (3) did not increase the likelihood of a fire or internal/external flooding event. The inspectors did not identify a cross-cutting area component related to this finding.
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| Inspection Report# : 2008004 (pdf)
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| Significance: Sep 30, 2008 Identified By: Self-Revealing Item Type: FIN Finding FAILURE TO EVALUATE AN UNEXPECTED AND UNKNOWN CAUSE FOR STRAY VOLTAGE IN THE END-OF-CYCLE RECIRCULATION PUMP TRIP CIRCUIT DURING POST MODIFICATION TESTING RESULTED IN A REACTOR RECIRCULATION PUMP TRIP The inspectors identified a finding of very low safety significance associated with a self-revealed event that resulted in the unexpected loss of a reactor recirculation pump. The licensee failed to evaluate an unexpected and unknown cause for stray voltage in the End-of-Cycle Recirculation Pump Trip (EOC-RPT) circuit during post modification testing during the Cycle 11 refueling outage that concluded in February 2008. This resulted in the unexpected loss of a reactor recirculation pump and the subsequent plant transient that led to a reactor scram. As an immediate and interim corrective action, the licensee implemented a design change to the EOC-RPT circuitry that should prevent inadvertent relay actuation causing recirculation pumps trips due to the stray voltage problem. No violation of regulatory requirements was identified.
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| The finding was of more than minor significance because this issue was associated with the Equipment Performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during power operations.
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| Specifically, the failure to evaluate an unexpected and unknown cause for stray voltage in the EOC-RPT circuit during post modification testing resulted in the unexpected loss of a reactor recirculation pump and the subsequent plant transient that led to a reactor scram. The finding was of very low safety significance because the issue: (1) did not contribute to the likelihood of a primary or secondary system loss-of-coolant-accident initiator, (2) did not contribute to both the likelihood of a reactor trip AND the likelihood that mitigation equipment for functions would not be available, and (3) did not increase the likelihood of a fire or internal/external flooding event. The inspectors concluded that this finding affected the cross-cutting area of human performance. Specifically, the licensee failed to appropriately incorporate risk insights in investigating and resolving an unexplained source of voltage in a circuit that had a high risk consequence (i.e., reactor recirculation pump trip). (IMC 0305 H.3(a))
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| Inspection Report# : 2008004 (pdf)
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| Mitigating Systems Significance: Sep 30, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation FAILURE TO PERFORM ADEQUATE PREVENTIVE MAINTENANCE ON SHUTDOWN SERVICE WATER VALVE 1SX014A RESULTED IN SIGNIFICANT DEGRADATION AND GROSS SEAT LEAKAGE A finding of very low safety significance with an associated Non-Cited Violation of Technical Specification (TS) 5.4.1.a was self-revealed. The licensee failed to perform adequate preventive maintenance on shutdown service water system valve 1SX014A. This resulted in significant degradation of the valve body by corrosion due to prolonged exposure to raw service water that went undetected until gross seat leakage was discovered while attempting to establish conditions for surveillance testing. The licensee replaced the valve and has established a preventive maintenance schedule for internal valve inspections.
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| The finding would become a more significant safety convern if left uncorrected and was therefore more than a minor concern. Specifically, the failure to adequately perform preventive maintenance could reasonably result in
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| significantly degraded or inoperable safety related equipment. Because the shutdown service water system was primarily associated with long term decay heat removal following certain design basis accidents, the inspectors concluded that this issue was associated with the Mitigating Systems cornerstone. The finding was of very low safety significance because the issue: (1) was not a design or qualification deficiency; (2) did not represent an actual loss of safety function of a system; (3) did not represent an actual loss of safety function of a single train for greater than its TS allowed outage time; (4) did not represent an actual loss of safety function of one or more non-TS trains of equipment designated as risk significant; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors concluded that this finding affected the cross-cutting area of human performance. Specifically, the licensee's investigation determined that internal valve inspections were not performed because the component category was incorrectly classified. (IMC 0305 H.3(b))
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| Inspection Report# : 2008004 (pdf)
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| Significance: Sep 30, 2008 Identified By: NRC Item Type: FIN Finding FAILURE TO RECOGNIZE THE SAFETY RELATED SYSTEM FUNCTION OF THE 1B RESIDUAL HEAT REMOVAL PUMP SEAL COOLER WHEN EVALUATING PAST OPERABILITY OF THE PUMP.
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| The inspectors identified a finding of very low safety significance associated with the licensee's failure to recognize the safety related system function of the 1B residual heat removal pump seal cooler when initially evaluating the past operability of the pump after unacceptable results were obtained during service water system flow balance testing. No analysis was performed to ensure that the pump's safety function would be fulfilled with less than minimum design flow to the cooler until the inspectors challenged the licensee's original conclusion. The licensee re-performed the past operability evaluation and determined that sufficient margin existed such that the pump would have been able to fulfill its safety function with significantly less than design flow to the seal cooler as measured during the test. No violation of regulatory requirements was identified.
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| The finding would become a more significant safety concern if left uncorrected and was therefore more than a minor concern. Specifically, the failure to correctly recognize the safety related functions of systems or components when performing operability or past operability evaluations could reasonably result in an unrecognized condition of a system failing to fulfill its safety related function. In addition, based on review of examples of minor issues in Inspection Manual Chapter 0612, "Power Reactor Inspection Reports," Appendix E, "Examples of Minor Issues,"
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| evaluation errors resulting in a reasonable doubt about the operability of a system or component are generally not considered to be of minor significance. Because the residual heat removal system was primarily associated with long term decay heat removal following certain design basis accidents, the inspectors condcluded that this issue was associated with the Mitigating Systems cornerstone. The finding was of very low safety significance because the issue: (1) was not a design or qualification deficienty; (2) did not represent an actual loss of safety function of a system; (3) did not represent an actual loss of safety function of a single train for greater than its Technical Specification (TS) allowed outage time; (4) did not represent an actual loss of safety function of one or more non-TS trains of equipment designated as risk signficiant; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. Subsequent evaluation was able to determine that sufficient margin in flow existed for the time period in question. The inspectors did not identify a corss-cutting area component related to this finding.
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| Inspection Report# : 2008004 (pdf)
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| Barrier Integrity Emergency Preparedness
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| Occupational Radiation Safety Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.
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| Miscellaneous Last modified : May 28, 2009
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| Clinton 2Q/2009 Plant Inspection Findings Initiating Events Significance: Dec 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO CONTROL TRANSIENT COMBUSTIBLE MATERIALS IN ACCORDANCE WITH FIRE PROTECTION PROGRAM.
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| GREEN. The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of the Clinton Power Station Unit 1 Operating License (NPF-62, Section 2.F). The licensee failed to implement the fire protection program in accordance with program requirements by failing to follow approved fire protection program procedures for the control of transient combustible matrials. The licensee promptly removed transient combustible materials found by the inspectors and subsequently completed a detailed walk down of the plant's transient combustible free zones to identify and remove any additional transient combustible materials.
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| The inspectors concluded that this finding could be reasonably viewed as a precursor to a significant event (i.e., a fire affecting more than one train of safe shutdown equipment). Specifically, the presence of transient combustible materials in a combustible free zone could reasonably result in degradation of the fire protection defense-in-depth elements in place to prevent fires from starting and mitigate the consequences of fires. In addition, based on review of Example 4k in Inspection Manual Chapter (IMC) 0612, "Power Reactor Inspection Reports," Appendix E, "Examples of Minor Issues," the issue would not be considered to be of minor significance because the identified transient combustibles were found in a combustible free zone required for separation of redundant trains. The finding was of very low safety significance because the items found in the combustible free zone would not be considered transient combustibles of significance as defined in IMC 0609, Appendix F, "Fire Protection Significance Determination Process," Attachment 2, "Degradation Rating Guidance Specific to Various Fire Protection Program Elements," and therefore the issue was assigned a "low degradation" rating. The inspectors concluded that this finding affected the cross-cutting area of problem identification and resolution. Specifically, the licensee missed an opportunity to identify and remove the transient combustible materials while implementing corrective actions for previously inspector identified findings involving the control of transient combustible materials. (IMC 0305 P.1 (a))
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| Inspection Report# : 2008005 (pdf)
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| Significance: Sep 30, 2008 Identified By: Self-Revealing Item Type: FIN Finding FAILURE TO PERFORM ADEQUATE POST MAINTENANCE TESTING RESULTED IN HIGH REACTOR VESSEL WATER LEVEL (LEVEL 8) SCRAM The inspectors identified a finding of very low safety significance associated with a self-revealed event that resulted in a Unit 1 reactor scram. The licensee failed to perform adequate post maintenance testing following replacement of the feedwater level control system dynamic compensator card during the Cycle 10 refueling outage that concluded in February 2006. This resulted in ineffective response from the feedwater level control system and a subsequent reactor scram following the unexpected loss of a reactor recirculation pump. The ineffective feedwater level control system response has not been corrected; however, the licensee entered this issue into its corrective action program for evaluation. No violation of regulatory requirements was identified.
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| The finding was of more than minor significance because this issue was associated with the Equipment Performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during power operations.
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| Specifically, inadequate post maintenance testing resulted in ineffective response from the feedwater level control system during a loss of a reactor recirculation pump transient and caused a reactor scram. The finding was of very low
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| safety significance because the issue: (1) did not contribute to the likelihood of a primary or secondary system loss-of-coolant-accident initiator, (2) did not contribute to both the likelihood of a reactor trip AND the likelihood that mitigation equipment or functions would not be available, and (3) did not increase the likelihood of a fire or internal/external flooding event. The inspectors did not identify a cross-cutting area component related to this finding.
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| Inspection Report# : 2008004 (pdf)
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| Significance: Sep 30, 2008 Identified By: Self-Revealing Item Type: FIN Finding FAILURE TO EVALUATE AN UNEXPECTED AND UNKNOWN CAUSE FOR STRAY VOLTAGE IN THE END-OF-CYCLE RECIRCULATION PUMP TRIP CIRCUIT DURING POST MODIFICATION TESTING RESULTED IN A REACTOR RECIRCULATION PUMP TRIP The inspectors identified a finding of very low safety significance associated with a self-revealed event that resulted in the unexpected loss of a reactor recirculation pump. The licensee failed to evaluate an unexpected and unknown cause for stray voltage in the End-of-Cycle Recirculation Pump Trip (EOC-RPT) circuit during post modification testing during the Cycle 11 refueling outage that concluded in February 2008. This resulted in the unexpected loss of a reactor recirculation pump and the subsequent plant transient that led to a reactor scram. As an immediate and interim corrective action, the licensee implemented a design change to the EOC-RPT circuitry that should prevent inadvertent relay actuation causing recirculation pumps trips due to the stray voltage problem. No violation of regulatory requirements was identified.
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| The finding was of more than minor significance because this issue was associated with the Equipment Performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during power operations.
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| Specifically, the failure to evaluate an unexpected and unknown cause for stray voltage in the EOC-RPT circuit during post modification testing resulted in the unexpected loss of a reactor recirculation pump and the subsequent plant transient that led to a reactor scram. The finding was of very low safety significance because the issue: (1) did not contribute to the likelihood of a primary or secondary system loss-of-coolant-accident initiator, (2) did not contribute to both the likelihood of a reactor trip AND the likelihood that mitigation equipment for functions would not be available, and (3) did not increase the likelihood of a fire or internal/external flooding event. The inspectors concluded that this finding affected the cross-cutting area of human performance. Specifically, the licensee failed to appropriately incorporate risk insights in investigating and resolving an unexplained source of voltage in a circuit that had a high risk consequence (i.e., reactor recirculation pump trip). (IMC 0305 H.3(a))
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| Inspection Report# : 2008004 (pdf)
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| Mitigating Systems Significance: Jun 30, 2009 Identified By: NRC Item Type: FIN Finding FAILURE TO EVALUTE SAFETY FUNCTION OF SUPPRESSION POOL MAKEUP SYSTEM The inspectors identified a finding of very low safety significance associated with the licensee's failure to recognize a potential loss of safety function for the suppression pool makeup system following the loss of upper containment pool inventory when spent fuel pool cooling system flow control valve 1FC004A failed closed. No evaluation was performed to ensure that the suppression pool makeup system's safety function would be fulfilled with less than Technical Specification (TS) minimum containment upper pool level. The licensee subsequently performed an evaluation and determined that sufficient margin existed such that the system would have been able to fulfill its safety function with limited margin. Corrective actions to address the inadequate reportability review included training for licensed senior reactor operators and development of a formal operability/reportability review process template. No violation of regulatory requirements was identified.
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| The finding would become a more significant safety concern if left uncorrected and was therefore, more than a minor
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| concern. Specifically, the failure to correctly recognize and evaluate a potential loss of a safety function of systems, structures, and components when performing operability or past operability evaluations could reasonably result in an unrecognized condition of a system failing to fulfill its safety-related function. Because the suppression pool makeup system was primarily associated with long term decay heat removal following certain design basis accidents, the inspectors concluded that this issue was associated with the Mitigating Systems cornerstone. The finding was of very low safety significance because the issue: (1) was not a design or qualification deficiency; (2) did not represent an actual loss of safety function of a system; (3) did not represent an actual loss of safety function of a single train for greater than its TS allowed outage time; (4) did not represent an actual loss of safety function of one or more non-TS trains of equipment designated as risk significant; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors concluded that this finding affected the cross-cutting area of human performance because the licensee did not have a formal process in place with adequate guidance and training to enable licensed senior reactor operators, whose resonsibility it was to evaluate a potential loss of safety function, to correctly do so. As a result, senior reactor operators did not adequately review the TS Bases to understand and evaluate whether the system was able to fullfill its safety function. (IMC 0305 H.1(a))
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| Inspection Report# : 2009003 (pdf)
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| Significance: Jun 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PERFROM SURVEILLANCE TESTING ON THE DIVISION 3 SHUTDOWN SERVICE WATER PUMP WITH ADEQUATE MEASURING AND TEST EQUIPMENT.
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| The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criteria XII, "Control of Measuring and Test Equipment," and 10 CFR 50, Appendix B, Criteria SI, "Test Control." The licensee failed to perform surveillance testing on the Division 3 shutdown service water pump with a lake level gage that was properly controlled and adjusted to ensure that is was readable within the range it was used. The licensee subsequently replaced the unreadable lake level gage section with one that was readable and implemented additional corrective actions to address a lapse in operations standards.
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| The inspectors concluded that this finding would become a more significant safety concern if left uncorrected and it was therefore more than a minor concern. Specifically, the failure to perform surveillance testing with properly controlled and accurate measuring and test equipment could reasonably result in the failure to identify degraded or inoperable safety-related components. Because the shutdown service water system was primarily associated with long term decay heat removal following certain design basis accidents, the inspectors concluded that this issue was associated with the Mitigating Systems Cornerstone. The finding was of very low safety significance because the issue was a design or qualification deficiency confirmed not to result in loss of operability or availability. The inspectors concluded that this finding affected the cross-cutting area of problem identification and resolution because the licensee was not properly maintaining the lake level gage to ensure that it would remain usable and did not correct the degraded level gage in a timely manner after it was identified. As a result, operators accepted the degraded level gage for continued use. (IMC 0305 P.1(d))
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| Inspection Report# : 2009003 (pdf)
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| Significance: Sep 30, 2008 Identified By: Self-Revealing Item Type: NCV NonCited Violation FAILURE TO PERFORM ADEQUATE PREVENTIVE MAINTENANCE ON SHUTDOWN SERVICE WATER VALVE 1SX014A RESULTED IN SIGNIFICANT DEGRADATION AND GROSS SEAT LEAKAGE A finding of very low safety significance with an associated Non-Cited Violation of Technical Specification (TS) 5.4.1.a was self-revealed. The licensee failed to perform adequate preventive maintenance on shutdown service water system valve 1SX014A. This resulted in significant degradation of the valve body by corrosion due to prolonged exposure to raw service water that went undetected until gross seat leakage was discovered while attempting to establish conditions for surveillance testing. The licensee replaced the valve and has established a preventive maintenance schedule for internal valve inspections.
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| The finding would become a more significant safety convern if left uncorrected and was therefore more than a minor concern. Specifically, the failure to adequately perform preventive maintenance could reasonably result in significantly degraded or inoperable safety related equipment. Because the shutdown service water system was primarily associated with long term decay heat removal following certain design basis accidents, the inspectors concluded that this issue was associated with the Mitigating Systems cornerstone. The finding was of very low safety significance because the issue: (1) was not a design or qualification deficiency; (2) did not represent an actual loss of safety function of a system; (3) did not represent an actual loss of safety function of a single train for greater than its TS allowed outage time; (4) did not represent an actual loss of safety function of one or more non-TS trains of equipment designated as risk significant; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors concluded that this finding affected the cross-cutting area of human performance. Specifically, the licensee's investigation determined that internal valve inspections were not performed because the component category was incorrectly classified. (IMC 0305 H.3(b))
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| Inspection Report# : 2008004 (pdf)
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| Significance: Sep 30, 2008 Identified By: NRC Item Type: FIN Finding FAILURE TO RECOGNIZE THE SAFETY RELATED SYSTEM FUNCTION OF THE 1B RESIDUAL HEAT REMOVAL PUMP SEAL COOLER WHEN EVALUATING PAST OPERABILITY OF THE PUMP.
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| The inspectors identified a finding of very low safety significance associated with the licensee's failure to recognize the safety related system function of the 1B residual heat removal pump seal cooler when initially evaluating the past operability of the pump after unacceptable results were obtained during service water system flow balance testing. No analysis was performed to ensure that the pump's safety function would be fulfilled with less than minimum design flow to the cooler until the inspectors challenged the licensee's original conclusion. The licensee re-performed the past operability evaluation and determined that sufficient margin existed such that the pump would have been able to fulfill its safety function with significantly less than design flow to the seal cooler as measured during the test. No violation of regulatory requirements was identified.
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| The finding would become a more significant safety concern if left uncorrected and was therefore more than a minor concern. Specifically, the failure to correctly recognize the safety related functions of systems or components when performing operability or past operability evaluations could reasonably result in an unrecognized condition of a system failing to fulfill its safety related function. In addition, based on review of examples of minor issues in Inspection Manual Chapter 0612, "Power Reactor Inspection Reports," Appendix E, "Examples of Minor Issues,"
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| evaluation errors resulting in a reasonable doubt about the operability of a system or component are generally not considered to be of minor significance. Because the residual heat removal system was primarily associated with long term decay heat removal following certain design basis accidents, the inspectors condcluded that this issue was associated with the Mitigating Systems cornerstone. The finding was of very low safety significance because the issue: (1) was not a design or qualification deficienty; (2) did not represent an actual loss of safety function of a system; (3) did not represent an actual loss of safety function of a single train for greater than its Technical Specification (TS) allowed outage time; (4) did not represent an actual loss of safety function of one or more non-TS trains of equipment designated as risk signficiant; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. Subsequent evaluation was able to determine that sufficient margin in flow existed for the time period in question. The inspectors did not identify a corss-cutting area component related to this finding.
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| Inspection Report# : 2008004 (pdf)
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| Barrier Integrity Emergency Preparedness
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| Occupational Radiation Safety Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.
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| Miscellaneous Significance: N/A Apr 17, 2009 Identified By: NRC Item Type: FIN Finding Biennial PI&R Inspection Summary The inspectors concluded that the implementation of the corrective action program (CAP) at Clinton was generally good. The licensee had a low threshold for identifying station problems and entering them into the CAP. In addition, the station was effective at incorporating operating experience reports into the CAP. The inspectors determined that issues were generally effectively screened and prioritized in a timely manner using established criteria based on plant risk and uncertainty. Casual evaluations sampled were of sufficient depth, considered extent of condition, generic issues, and previous occurrences. Corrective actions program assignments were generally completed in a timely and accurate manner. The team noted that station effectiveness reviews, audits, and self assessment were generally thorough and effective at identifying unrecognized weakness. The inspectors concluded that station employees appeared to be willing to express safety concerns through established processes and a healthy safety conscious work environment (SCWE) existed at the station.
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| Inspection Report# : 2009007 (pdf)
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| Last modified : August 31, 2009
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| Clinton 3Q/2009 Plant Inspection Findings Initiating Events Significance: Dec 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO CONTROL TRANSIENT COMBUSTIBLE MATERIALS IN ACCORDANCE WITH FIRE PROTECTION PROGRAM.
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| The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of the Clinton Power Station Unit 1 Operating License (NPF-62, Section 2.F). The licensee failed to implement the fire protection program in accordance with program requirements by failing to follow approved fire protection program procedures for the control of transient combustible matrials. The licensee promptly removed transient combustible materials found by the inspectors and subsequently completed a detailed walk down of the plant's transient combustible free zones to identify and remove any additional transient combustible materials.
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| The inspectors concluded that this finding could be reasonably viewed as a precursor to a significant event (i.e., a fire affecting more than one train of safe shutdown equipment). Specifically, the presence of transient combustible materials in a combustible free zone could reasonably result in degradation of the fire protection defense-in-depth elements in place to prevent fires from starting and mitigate the consequences of fires. In addition, based on review of Example 4k in Inspection Manual Chapter (IMC) 0612, "Power Reactor Inspection Reports," Appendix E, "Examples of Minor Issues," the issue would not be considered to be of minor significance because the identified transient combustibles were found in a combustible free zone required for separation of redundant trains. The finding was of very low safety significance because the items found in the combustible free zone would not be considered transient combustibles of significance as defined in IMC 0609, Appendix F, "Fire Protection Significance Determination Process," Attachment 2, "Degradation Rating Guidance Specific to Various Fire Protection Program Elements," and therefore the issue was assigned a "low degradation" rating. The inspectors concluded that this finding affected the cross-cutting area of problem identification and resolution. Specifically, the licensee missed an opportunity to identify and remove the transient combustible materials while implementing corrective actions for previously inspector identified findings involving the control of transient combustible materials. (IMC 0305 P.1 (a))
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| Inspection Report# : 2008005 (pdf)
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| Mitigating Systems Significance: Sep 30, 2009 Identified By: Self-Revealing Item Type: FIN Finding INEFFECTIVE CORRECTIVE ACTIONS FOR VIBRATION INDUCED STEM/DISC SEPARATION OF FUEL POOL COOLING SYSTEM TRAIN 'A' FLOW CONTROL VALVE 1FC004A.
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| A finding of very low safety significance was self-revealed on May 27, 2009, when fuel pool cooling system flow control valve 1FC004A failed closed. The licensee failed to implement effective corrective actions in response to the same failure mode for the valve that occurred on November 21, 2005. This resulted in the failure of 1FC0014A once again and the subsequent loss of inventory from the containment upper pool and inoperability of the suppression pool makeup system. The licensee entered this issue into its corrective action program to investigate the cause and to identify appropriate corrective actions. No violation of regulatory requirements was identified.
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| The finding was of more than minor significance because it was associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone and directly affected the cornerstone objective to ensure the availability,
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| reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.
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| Specifically, the May 2009 valve failure resulted in a loss of inventory from the containment upper pool and inoperability of the suppression pool makeup system, therefore impacting its availability for certain initiating events.
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| The finding was of very low safety significance because the issue: (1) was not a design or qualification deficiency; (2) did not represent an actual loss of safety function of a system; (3) did not represent an actual loss of safety function of a single train for greater than its Technical Specification (TS) allowed outage time; (4) did not represent an actual loss of safety function of one or more non-TS trains of equipment designated as risk significant; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors did not identify a cross-cutting area component related to this finding.
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| Inspection Report# : 2009004 (pdf)
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| Significance: Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ADEQUATELY IMPLEMENT REQUIREMENTS OF THE LEAKAGE REDUCTION AND MONITORING PROGRAM.
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| The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criteria V, "Instructions, Procedures, and Drawings," regarding the licensee's failure to adequately implement periodic visual inspection requirements to monitor and minimize leakage from piping systems connecting to the reactor coolant system. The licensee entered this violation into its corrective action program to investigate the cause and to identify appropriate corrective actions.
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| The finding would become a more significant safety concern if left uncorrected and was therefore more than a minor concern. Specifically, the failure to adequately implement required leakage reduction and monitoring program controls to minimize leakage from reactor coolant sources outside of containment that could contain highly radioactive fluids during a serious transient or accident could reasonably result in higher doses to plant workers and higher potential offsite release levels. Because the leakage reducation and monitoring program is intended to contain highly radioactive fluids within piping systems outside containment, which supports the radiological barrier functions to protect plant workers and the public following serious transients or accidents, the inspectors concluded that this issue was associated with the Barrier Integrity Cornerstone. The finding was of very low safety significance because it invovled only a degradation of the radiological barrier function provided for the Auxiliary Building. The inspectors concluded that this finding affected the cross-cutting area of human performance because the licensee did not provide adequate procedural guidance and training to enable operators to correctly perform and document piping system visual inspections to implement its leakage reduction and monitoring program. As a result, the licensee did not have appropriate objective quality evidence to demonstrate that the program requirements were met. (IMC 0305 H.2(c))
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| Inspection Report# : 2009004 (pdf)
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| Significance: Jun 30, 2009 Identified By: NRC Item Type: FIN Finding FAILURE TO EVALUTE SAFETY FUNCTION OF SUPPRESSION POOL MAKEUP SYSTEM The inspectors identified a finding of very low safety significance associated with the licensee's failure to recognize a potential loss of safety function for the suppression pool makeup system following the loss of upper containment pool inventory when spent fuel pool cooling system flow control valve 1FC004A failed closed. No evaluation was performed to ensure that the suppression pool makeup system's safety function would be fulfilled with less than Technical Specification (TS) minimum containment upper pool level. The licensee subsequently performed an evaluation and determined that sufficient margin existed such that the system would have been able to fulfill its safety function with limited margin. Corrective actions to address the inadequate reportability review included training for licensed senior reactor operators and development of a formal operability/reportability review process template. No violation of regulatory requirements was identified.
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| The finding would become a more significant safety concern if left uncorrected and was therefore, more than a minor concern. Specifically, the failure to correctly recognize and evaluate a potential loss of a safety function of systems, structures, and components when performing operability or past operability evaluations could reasonably result in an unrecognized condition of a system failing to fulfill its safety-related function. Because the suppression pool makeup
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| system was primarily associated with long term decay heat removal following certain design basis accidents, the inspectors concluded that this issue was associated with the Mitigating Systems cornerstone. The finding was of very low safety significance because the issue: (1) was not a design or qualification deficiency; (2) did not represent an actual loss of safety function of a system; (3) did not represent an actual loss of safety function of a single train for greater than its TS allowed outage time; (4) did not represent an actual loss of safety function of one or more non-TS trains of equipment designated as risk significant; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors concluded that this finding affected the cross-cutting area of human performance because the licensee did not have a formal process in place with adequate guidance and training to enable licensed senior reactor operators, whose resonsibility it was to evaluate a potential loss of safety function, to correctly do so. As a result, senior reactor operators did not adequately review the TS Bases to understand and evaluate whether the system was able to fullfill its safety function. (IMC 0305 H.1(a))
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| Inspection Report# : 2009003 (pdf)
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| Significance: Jun 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PERFROM SURVEILLANCE TESTING ON THE DIVISION 3 SHUTDOWN SERVICE WATER PUMP WITH ADEQUATE MEASURING AND TEST EQUIPMENT.
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| The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criteria XII, "Control of Measuring and Test Equipment," and 10 CFR 50, Appendix B, Criteria SI, "Test Control." The licensee failed to perform surveillance testing on the Division 3 shutdown service water pump with a lake level gage that was properly controlled and adjusted to ensure that is was readable within the range it was used. The licensee subsequently replaced the unreadable lake level gage section with one that was readable and implemented additional corrective actions to address a lapse in operations standards.
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| The inspectors concluded that this finding would become a more significant safety concern if left uncorrected and it was therefore more than a minor concern. Specifically, the failure to perform surveillance testing with properly controlled and accurate measuring and test equipment could reasonably result in the failure to identify degraded or inoperable safety-related components. Because the shutdown service water system was primarily associated with long term decay heat removal following certain design basis accidents, the inspectors concluded that this issue was associated with the Mitigating Systems Cornerstone. The finding was of very low safety significance because the issue was a design or qualification deficiency confirmed not to result in loss of operability or availability. The inspectors concluded that this finding affected the cross-cutting area of problem identification and resolution because the licensee was not properly maintaining the lake level gage to ensure that it would remain usable and did not correct the degraded level gage in a timely manner after it was identified. As a result, operators accepted the degraded level gage for continued use. (IMC 0305 P.1(d))
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| Inspection Report# : 2009003 (pdf)
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| Barrier Integrity Significance: SL-IV Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO UPDATE THE FINAL SAFETY ANALYSIS REPORT.
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| The inspectors identified a Non-Cited Violation of 10 CFR 50.71, "Maintenance of Records, Making of Reports,"
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| associated with the licensee's failure to correctly update the Updated Fainal Safety Analysis Report (UFSAR) when modifying Technical Specification (TS) requirements for the Control Room ventilation system during implementation of Improved Standard Technical Specifications. Specifically, the licensee failed to change the specified safety function description for the system to maintain positive pressure within the Control Room envelope with respect to adjacent areas during all operating modes except when the system is in the recirculation mode or when the system is in the maximum outside air purge mode. This directly contributed to the licensee's failure to correctly evaluate the operability of Control Room ventilation system Train 'B' when the system was unable to maintain the Control Room envelope at a positive pressure relative to adjacent areas while operating in the normal mode. Subsequent evaluation by the inspectors determined that the safety function description in the UFSAR was inaccurate and the system was
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| operable with the degraded/nonconforming condition. The licensee entered this violation into its corrective action program to investigate the cause and to identify appropriate corrective actions.
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| Because the issue affected the NRC's ability to perform its regulatory function, the violation was reviewed under the traditional enforcement process; however, the underlying technical issue was evaluated using the Significance Determination Process. The finding would become a more significant safety concern if left uncorrected and was therefore more than a minor concern. Specifically, the failure to correctly evaluate a degraded/nonconforming condition potentially affecting the operability of a system, structure, or component (SSC) required to be operable by TS could reasonably result in an unrecognized condition of an SSC failing to fulfill a safety-related function. Because the Control Room ventilation system supports the radiological barrier function to protect operators inside the Control Room following certain design basis accidents, the inspectors concluded that this issue was associated with the Barrier Integrity Cornerstone. The finding was of very low safety significance because it involved only a degradation of the radiological barrier function provided for the Control Room. The inspectors did not identify a cross-cutting aspect related to this finding.
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| Inspection Report# : 2009004 (pdf)
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| Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.
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| Miscellaneous Significance: N/A Apr 17, 2009 Identified By: NRC Item Type: FIN Finding Biennial PI&R Inspection Summary The inspectors concluded that the implementation of the corrective action program (CAP) at Clinton was generally good. The licensee had a low threshold for identifying station problems and entering them into the CAP. In addition, the station was effective at incorporating operating experience reports into the CAP. The inspectors determined that issues were generally effectively screened and prioritized in a timely manner using established criteria based on plant risk and uncertainty. Casual evaluations sampled were of sufficient depth, considered extent of condition, generic issues, and previous occurrences. Corrective actions program assignments were generally completed in a timely and accurate manner. The team noted that station effectiveness reviews, audits, and self assessment were generally thorough and effective at identifying unrecognized weakness. The inspectors concluded that station employees appeared to be willing to express safety concerns through established processes and a healthy safety conscious work environment (SCWE) existed at the station.
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| Inspection Report# : 2009007 (pdf)
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| Last modified : December 10, 2009
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| Clinton 4Q/2009 Plant Inspection Findings Initiating Events Significance: Dec 31, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation FAILURE TO CORRECTLY TORQUE VALVE PACKING GLAND NUTS RESULTED IN VALVE PACKING FAILURE AND UNPLANNED PLANT SHUTDOWN A finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criteria V, "Instructions, Procedures, and Drawings," was self-revealed on September 29, 2009, when a steam leak developed from the reactor core isolation cooling (RCIC) system inboard steam isolation valve (1E51F0063) stem packing. This resulted in a plant shutdown due to a greater than 2 gallon-per-minute increase in unidentified reactor coolant system (RCS) leakage within the previous 24 hours. The licensee failed to correctly tighten the valve packing gland nuts to the as-left torque valve from original packing installation when performing scheduled maintenance to verify the as-found torque value. The licensee replaced the 1E51F0063 valve stem packing during the subsequent forced outage and tightened the gland nuts to the correct torque value.
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| The finding was of more than minor significance because it was associated with the Equipment Performance attribute of the Initiaging Events Cornerstone and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the failure to correctly tighten the valve stem packing gland nuts resulted in stem packing failure and a subsequent plant shutdown due to exceeding the Technical Specification (TS) limit for an increase in unidentified RCS leakage.
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| Although the finding resulted in exceeding the TS limit for RCS leakage, it was determined to be of very low safety significance during a Phase 2 Significance Determination Process review because there was no loss of mitigation capability for any safety system and therefore no resultant change in core damage frequency. Because the performance issue was associated with maintenance performed in February 2006, it did not necessarily reflect current licensee performance and no cross-cutting aspect was identified.
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| Inspection Report# : 2009005 (pdf)
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| Mitigating Systems Significance: Dec 31, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation FAILURE TO CORRECTLY INSTALL RELAYS INSIDE OF THE DIVISION 3 DIESEL GENERATOR CONTROL PANEL A finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criteria V, "Instructions, Procedures, and Drawings," was self-revealed on September 23, 2009, when the Division 3 diesel generator (DG) was found to have had two components installed incorrectly. Electrical maintenance technicians had incorrectly replaced time delay relays K-8A and K-32 on September 24, 2007, essentially swapping the locations of the two relays. This rendered the Division 3 DG inoperable for about 2 years and resulted in a loss of safety function for the Division 3 DG and high pressure core spray system under a certain sequence of initiating events. The licensee restored the two time delay relays in the correct configuration and immediately verified that the remaining time delay relays inside the Division 3 DG Control Panel were in their proper locations.
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| The finding was of more than minor significance because, if left uncorrected,it would potentially lead to a more significant safety concern (i.e., the inoperability of risk-significant plant safety systems). In addition, based on review
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| of Example 5c in IMC 0612, "Power Reactor Inspection Reports," Appendix E, "Examples of Minor Issues," the issue would not be considered to be of minor significance because the incorrect relays were installed in the control panel.
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| Although the finding resulted in a loss of safetyfunction for the Division 3 DG and high pressure core spray system, it was determined to be of very low safety significance during a Phase 2 Significance Determination Process Review considering the very limited conditions (i.e., only 45 seconds following shutdown of the engine concurrent with a design basis accident) when the Division 3 DG was incapable of performing its safety function. The resultant exposure time was estimated to be about 27 minutes during the 2-year period. The inspectors concluded that this finding affected the cross-cutting area of human performance because the licensee did not effectively communicate expectations regarding procedural compliance and, as a result, maintenance technicians did not follow their procedures by installing nonconforming components and restoring the safety system to service. (IMC 0305 H.4(b))
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| Inspection Report# : 2009005 (pdf)
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| Significance: Sep 30, 2009 Identified By: Self-Revealing Item Type: FIN Finding INEFFECTIVE CORRECTIVE ACTIONS FOR VIBRATION INDUCED STEM/DISC SEPARATION OF FUEL POOL COOLING SYSTEM TRAIN 'A' FLOW CONTROL VALVE 1FC004A.
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| A finding of very low safety significance was self-revealed on May 27, 2009, when fuel pool cooling system flow control valve 1FC004A failed closed. The licensee failed to implement effective corrective actions in response to the same failure mode for the valve that occurred on November 21, 2005. This resulted in the failure of 1FC0014A once again and the subsequent loss of inventory from the containment upper pool and inoperability of the suppression pool makeup system. The licensee entered this issue into its corrective action program to investigate the cause and to identify appropriate corrective actions. No violation of regulatory requirements was identified.
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| The finding was of more than minor significance because it was associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone and directly affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.
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| Specifically, the May 2009 valve failure resulted in a loss of inventory from the containment upper pool and inoperability of the suppression pool makeup system, therefore impacting its availability for certain initiating events.
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| The finding was of very low safety significance because the issue: (1) was not a design or qualification deficiency; (2) did not represent an actual loss of safety function of a system; (3) did not represent an actual loss of safety function of a single train for greater than its Technical Specification (TS) allowed outage time; (4) did not represent an actual loss of safety function of one or more non-TS trains of equipment designated as risk significant; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors did not identify a cross-cutting area component related to this finding.
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| Inspection Report# : 2009004 (pdf)
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| Significance: Jun 30, 2009 Identified By: NRC Item Type: FIN Finding FAILURE TO EVALUTE SAFETY FUNCTION OF SUPPRESSION POOL MAKEUP SYSTEM The inspectors identified a finding of very low safety significance associated with the licensee's failure to recognize a potential loss of safety function for the suppression pool makeup system following the loss of upper containment pool inventory when spent fuel pool cooling system flow control valve 1FC004A failed closed. No evaluation was performed to ensure that the suppression pool makeup system's safety function would be fulfilled with less than Technical Specification (TS) minimum containment upper pool level. The licensee subsequently performed an evaluation and determined that sufficient margin existed such that the system would have been able to fulfill its safety function with limited margin. Corrective actions to address the inadequate reportability review included training for licensed senior reactor operators and development of a formal operability/reportability review process template. No violation of regulatory requirements was identified.
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| The finding would become a more significant safety concern if left uncorrected and was therefore, more than a minor concern. Specifically, the failure to correctly recognize and evaluate a potential loss of a safety function of systems, structures, and components when performing operability or past operability evaluations could reasonably result in an unrecognized condition of a system failing to fulfill its safety-related function. Because the suppression pool makeup
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| system was primarily associated with long term decay heat removal following certain design basis accidents, the inspectors concluded that this issue was associated with the Mitigating Systems cornerstone. The finding was of very low safety significance because the issue: (1) was not a design or qualification deficiency; (2) did not represent an actual loss of safety function of a system; (3) did not represent an actual loss of safety function of a single train for greater than its TS allowed outage time; (4) did not represent an actual loss of safety function of one or more non-TS trains of equipment designated as risk significant; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors concluded that this finding affected the cross-cutting area of human performance because the licensee did not have a formal process in place with adequate guidance and training to enable licensed senior reactor operators, whose resonsibility it was to evaluate a potential loss of safety function, to correctly do so. As a result, senior reactor operators did not adequately review the TS Bases to understand and evaluate whether the system was able to fullfill its safety function. (IMC 0305 H.1(a))
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| Inspection Report# : 2009003 (pdf)
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| Inspection Report# : 2009005 (pdf)
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| Significance: Jun 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PERFROM SURVEILLANCE TESTING ON THE DIVISION 3 SHUTDOWN SERVICE WATER PUMP WITH ADEQUATE MEASURING AND TEST EQUIPMENT.
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| The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criteria XII, "Control of Measuring and Test Equipment," and 10 CFR 50, Appendix B, Criteria XI, "Test Control." The licensee failed to perform surveillance testing on the Division 3 shutdown service water pump with a lake level gage that was properly controlled and adjusted to ensure that it was readable within the range it was used. The licensee subsequently replaced the unreadable lake level gage section with one that was readable and implemented additional corrective actions to address a lapse in operations standards.
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| The inspectors concluded that this finding would become a more significant safety concern if left uncorrected and it was therefore more than a minor concern. Specifically, the failure to perform surveillance testing with properly controlled and accurate measuring and test equipment could reasonably result in the failure to identify degraded or inoperable safety-related components. Because the shutdown service water system was primarily associated with long term decay heat removal following certain design basis accidents, the inspectors concluded that this issue was associated with the Mitigating Systems Cornerstone. The finding was of very low safety significance because the issue was a design or qualification deficiency confirmed not to result in loss of operability or availability. The inspectors concluded that this finding affected the cross-cutting area of problem identification and resolution because the licensee was not properly maintaining the lake level gage to ensure that it would remain usable and did not correct the degraded level gage in a timely manner after it was identified. As a result, operators accepted the degraded level gage for continued use. (IMC 0305 P.1(d))
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| Inspection Report# : 2009003 (pdf)
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| Barrier Integrity Significance: SL-IV Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO UPDATE THE FINAL SAFETY ANALYSIS REPORT.
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| The inspectors identified a Non-Cited Violation of 10 CFR 50.71, "Maintenance of Records, Making of Reports,"
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| associated with the licensee's failure to correctly update the Updated Fainal Safety Analysis Report (UFSAR) when modifying Technical Specification (TS) requirements for the Control Room ventilation system during implementation of Improved Standard Technical Specifications. Specifically, the licensee failed to change the specified safety function description for the system to maintain positive pressure within the Control Room envelope with respect to adjacent areas during all operating modes except when the system is in the recirculation mode or when the system is in the maximum outside air purge mode. This directly contributed to the licensee's failure to correctly evaluate the operability of Control Room ventilation system Train 'B' when the system was unable to maintain the Control Room envelope at a positive pressure relative to adjacent areas while operating in the normal mode. Subsequent evaluation
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| by the inspectors determined that the safety function description in the UFSAR was inaccurate and the system was operable with the degraded/nonconforming condition. The licensee entered this violation into its corrective action program to investigate the cause and to identify appropriate corrective actions.
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| Because the issue affected the NRC's ability to perform its regulatory function, the violation was reviewed under the traditional enforcement process; however, the underlying technical issue was evaluated using the Significance Determination Process. The finding would become a more significant safety concern if left uncorrected and was therefore more than a minor concern. Specifically, the failure to correctly evaluate a degraded/nonconforming condition potentially affecting the operability of a system, structure, or component (SSC) required to be operable by TS could reasonably result in an unrecognized condition of an SSC failing to fulfill a safety-related function. Because the Control Room ventilation system supports the radiological barrier function to protect operators inside the Control Room following certain design basis accidents, the inspectors concluded that this issue was associated with the Barrier Integrity Cornerstone. The finding was of very low safety significance because it involved only a degradation of the radiological barrier function provided for the Control Room. The inspectors did not identify a cross-cutting aspect related to this finding.
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| Inspection Report# : 2009004 (pdf)
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| Significance: Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ADEQUATELY IMPLEMENT REQUIREMENTS OF THE LEAKAGE REDUCTION AND MONITORING PROGRAM.
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| The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criteria V, "Instructions, Procedures, and Drawings," regarding the licensee's failure to adequately implement periodic visual inspection requirements to monitor and minimize leakage from piping systems connecting to the reactor coolant system. The licensee entered this violation into its corrective action program to investigate the cause and to identify appropriate corrective actions.
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| The finding would become a more significant safety concern if left uncorrected and was therefore more than a minor concern. Specifically, the failure to adequately implement required leakage reduction and monitoring program controls to minimize leakage from reactor coolant sources outside of containment that could contain highly radioactive fluids during a serious transient or accident could reasonably result in higher doses to plant workers and higher potential offsite release levels. Because the leakage reducation and monitoring program is intended to contain highly radioactive fluids within piping systems outside containment, which supports the radiological barrier functions to protect plant workers and the public following serious transients or accidents, the inspectors concluded that this issue was associated with the Barrier Integrity Cornerstone. The finding was of very low safety significance because it invovled only a degradation of the radiological barrier function provided for the Auxiliary Building. The inspectors concluded that this finding affected the cross-cutting area of human performance because the licensee did not provide adequate procedural guidance and training to enable operators to correctly perform and document piping system visual inspections to implement its leakage reduction and monitoring program. As a result, the licensee did not have appropriate objective quality evidence to demonstrate that the program requirements were met. (IMC 0305 H.2(c))
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| Inspection Report# : 2009004 (pdf)
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| Emergency Preparedness Significance: SL-IV Nov 20, 2009 Identified By: NRC Item Type: NCV NonCited Violation Implementation of a Change which Decreased the Effectiveness of the Emergency Plan The inspectors identified a NCV of 10 CFR 50.54(q) associated with 10 CFR 50.47(b)(2) because the licensee failed to obtain prior NRC approval for a change made to its emergency plan that decreased the effectiveness of the plan.
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| Specifically, the licensee removed staffing and capabilities from the minimum on-shift emergency response staffing requirements from the Clinton Power Station Emergency Plan Annex, Section 2, Table B-1. The licensee entered this issue into their corrective action program and replaced staffing back on-shift as required by the 1998 emergency plan
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| annex.
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| This finding was more than minor and of very low safety-significance using IMC 0609, Appendix B, because the finding was associated with the Emergency Preparedness Cornerstone attribute of emergency response organization readiness for minimum on shift emergency response staffing. Because the finding affected the NRC's ability to perform its regulatory function, the inspectors evaluated the significance using the traditional enforcement process.
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| This finding was determined to be a Severity Level IV violation because the licensee failed to meet an emergency planning requirement not directly related to assessment and notification. The inspectors determined that this finding had a cross-cutting aspect in the area of Human Performance, decision making because the licensee did not initially recognize that the removal of minimum on-shift emergency response staffing decreased the effectiveness of the emergency plan (H.1.(b)).
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| Inspection Report# : 2009502 (pdf)
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| Occupational Radiation Safety Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.
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| Miscellaneous Significance: N/A Apr 17, 2009 Identified By: NRC Item Type: FIN Finding Biennial PI&R Inspection Summary The inspectors concluded that the implementation of the corrective action program (CAP) at Clinton was generally good. The licensee had a low threshold for identifying station problems and entering them into the CAP. In addition, the station was effective at incorporating operating experience reports into the CAP. The inspectors determined that issues were generally effectively screened and prioritized in a timely manner using established criteria based on plant risk and uncertainty. Casual evaluations sampled were of sufficient depth, considered extent of condition, generic issues, and previous occurrences. Corrective actions program assignments were generally completed in a timely and accurate manner. The team noted that station effectiveness reviews, audits, and self assessment were generally thorough and effective at identifying unrecognized weakness. The inspectors concluded that station employees appeared to be willing to express safety concerns through established processes and a healthy safety conscious work environment (SCWE) existed at the station.
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| Inspection Report# : 2009007 (pdf)
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| Last modified : March 01, 2010
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| Clinton 1Q/2010 Plant Inspection Findings Initiating Events Significance: Mar 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO CONTROL TRANSIENT COMBUSTIBLE MATERIALS IN ACCORDANCE WITH FIRE PROTECTION PROGRAM The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of the Clinton Power Station Unit 1 Operating License (NPF-62, Section 2.F). The licensee failed to implement the Fire Protection Program in accordance with program requirements by failing to follow approved Fire Protection Program procedures for the control of transient combustible materials. The licensee promptly removed the transient combustible materials found by the inspectors.
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| The inspectors concluded that this finding could be reasonably viewed as a precursor to a significant event (i.e., a fire affecting more than one train of safe shutdown equipment). Specifically, the presence of transient combustible materials in a combustible free zone could reasonably result in degradation of the fire protection defense-in-depth elements in place to prevent fires from starting and mitigate the consequences of fires. In addition, based on review of Example 4k in IMC 0612, "Power Reactor Inspection Reports," Appendix E, "Examples of Minor Issues," the issue would not be considered to be of minor significance because the identified transient combustibles were found in a combustible free zone required for separation of redundant trains. The finding was of very low safety significance because the items found in the combustible free zone would not be considered transient combustibles of significance as defined in IMC 0609, Appendix F, "Fire Protection Significance Determination Process," Attachment 2, "Degradation Rating Guidance Specific to Various Fire Protection Program Elements," and therefore the issue was assigned a "low degradation" rating. The inspectors concluded that this finding affected the cross-cutting area of problem identification and resolution. Specifically, the licensee missed an opportunity to identify and remove the transient combustible materials while implementing corrective actions for previous inspector identified findings involving the control of transient combustible materials.
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| Inspection Report# : 2010002 (pdf)
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| Significance: Mar 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO CONTROL COMBUSTIBLE GAS CYLINDERS IN ACCORDANCE WITH FIRE PROTECTION PROGRAM The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of the Clinton Power Station Unit 1 Operating License (NPF-62, Section 2.F). The licensee failed to implement the Fire Protection Program in accordance with program requirements by failing to follow approved Fire Protection Program procedures for the control of combustible gas cylinders in the plant. The licensee promptly removed the combustible gas cylinders found by the inspectors.
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| The inspectors concluded that this finding was associated with the Protection Against External Factors attribute of the Initiating Events Cornerstone and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the fire hazard for the affected area was increased by the uncontrolled presence of the compressed gas cylinders. In addition, based on review of Example 4k in Inspection Manual Chapter 0612, "Power Reactor Inspection Reports,"
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| Appendix E, "Examples of Minor Issues," the issue would not be considered to be of minor significance because a credible fire scenario involving the identified transient combustibles could affect equipment important to safety. The finding was detemined to be of very low safety significance during a Phase 3 Significance Determination Process review since the delta core damage frequency was determined to be negligible. Because a postulated fire in the area
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| where the combustible gas cylinders were found could affect only one train of safe shutdown equipment, the safe shutdown path was not affected by the finding. The inspectors concluded that this finding affected the cross-cutting area of human performance. Specifically, the licensee did not adquately ensure that supervisory and management oversight of work activities involving contractors supported nuclear safety.
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| Inspection Report# : 2010002 (pdf)
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| Significance: Mar 31, 2010 Identified By: Self-Revealing Item Type: FIN Finding FAILURE TO CORRECT INADEQUATE FWLCS RESPONSE RESULTED IN HIGH REACTOR VESSEL WATER LEVEL (LEVEL 8 ) SCRAM A finding of very low safety significance was self-revealed from an event that resulted in a Unit 1 reactor scram. The licensee failed to correct a non-conforming condition with inadequate response from the feedwater level control system (FWLCS) that caused an automatic reactor scram on Febraury 10, 2008, following an unexpected loss of a reactor recirculation pump. This resulted in a second reactor scram for the same cause on October 15, 2009, following the unexpected loss of a reactor recirculatio pump. Because the FWLCS is not safety-related, no violation of regulatory requirement was identified. The FWLCS response was corrected in January 2010 and proper system response was verified by the licensee upon start up from the January - February 2010 refueling outage.
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| The finding was of more than minor significance because this issue was associated with the Equipment Performance attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during power operations.
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| Specifically, inadequate FWLCS response resulted in a reactor scram following the unexpected loss of a reactor recirculation pump. The finding was of very low safety significance because the issue: (1) did not contribute to the likelihood of a primary or secondary system loss-of-collant-accident initiator, (2) did not contribute to both the likelihood of a ractor trip AND the likelihood that mitigation equipment or functions would not be available, and (3) did not increase the likelihood of a fire or internal/external flooding event. The inspectors did not identify a cross-cutting aspect related to this finding.
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| Inspection Report# : 2010002 (pdf)
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| Significance: Dec 31, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation FAILURE TO CORRECTLY TORQUE VALVE PACKING GLAND NUTS RESULTED IN VALVE PACKING FAILURE AND UNPLANNED PLANT SHUTDOWN A finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criteria V, "Instructions, Procedures, and Drawings," was self-revealed on September 29, 2009, when a steam leak developed from the reactor core isolation cooling (RCIC) system inboard steam isolation valve (1E51F0063) stem packing. This resulted in a plant shutdown due to a greater than 2 gallon-per-minute increase in unidentified reactor coolant system (RCS) leakage within the previous 24 hours. The licensee failed to correctly tighten the valve packing gland nuts to the as-left torque valve from original packing installation when performing scheduled maintenance to verify the as-found torque value. The licensee replaced the 1E51F0063 valve stem packing during the subsequent forced outage and tightened the gland nuts to the correct torque value.
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| The finding was of more than minor significance because it was associated with the Equipment Performance attribute of the Initiating Events Cornerstone and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the failure to correctly tighten the valve stem packing gland nuts resulted in stem packing failure and a subsequent plant shutdown due to exceeding the Technical Specification (TS) limit for an increase in unidentified RCS leakage.
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| Although the finding resulted in exceeding the TS limit for RCS leakage, it was determined to be of very low safety significance during a Phase 2 Significance Determination Process review because there was no loss of mitigation capability for any safety system and therefore no resultant change in core damage frequency. Because the performance issue was associated with maintenance performed in February 2006, it did not necessarily reflect current licensee performance and no cross-cutting aspect was identified.
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| Inspection Report# : 2009005 (pdf)
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| Mitigating Systems Significance: Mar 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation INTERCONNECTING FLOOR DRAINS BETWEEN THE RESIDUAL HEAT REMOVAL 'A' PUMP ROOM AND RADWASTE PIPE TUNNEL The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criteria III, "Design Control," regarding the licensee's failure to correctly translate the design basis into the design of the Auxiliary Building floor drain system with appropriate margin. The inspectors identified that floor drains in the Residual Heat Removal (RHR) 'A' Pump Room and the Radwaste Pipe Tunnel were interconnected, which resulted in the plant being in an unanalyzed condition that degraded plant safety and could have prevented fulfillment of the safety function of the containment suppression pool. To address the immediate operability concern, the licensee plugged the two floor drains in the Radwaste Pipe Tunnel line to prevent communication with the floor drain system in the RHR 'A' Pump Room. An exposed vertical section of the drain line was then cut and a solid steel plate welded into the pipe per an engineering design change to permanently isolate the floor drains between the two rooms.
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| The finding was of more than minor significance because it was associated with the Design Control attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage).
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| Specifically, the as-found configuration of the interconnecting floor drains resulted in the plant being in an unanalyzed condition that could have prevented fullfillment of the safety function of the containment suppression pool. Although the finding would represent a loss of safety function in the event of a postulated accident, it was determined to be of very low safety significance during a Phase 3 Significance Determination Process review because the delta core damage frequency was determined to be negligible since the initiating event frequency for flooding due to an RHR pump suction pipe failure was sufficiently low. Because this condition had existed since initial plant construction, the performance issue did not necessarily reflect current licensee performance and no cross-cutting aspect was identified.
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| Inspection Report# : 2010002 (pdf)
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| Significance: Dec 31, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation FAILURE TO CORRECTLY INSTALL RELAYS INSIDE OF THE DIVISION 3 DIESEL GENERATOR CONTROL PANEL A finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criteria V, "Instructions, Procedures, and Drawings," was self-revealed on September 23, 2009, when the Division 3 diesel generator (DG) was found to have had two components installed incorrectly. Electrical maintenance technicians had incorrectly replaced time delay relays K-8A and K-32 on September 24, 2007, essentially swapping the locations of the two relays. This rendered the Division 3 DG inoperable for about 2 years and resulted in a loss of safety function for the Division 3 DG and high pressure core spray system under a certain sequence of initiating events. The licensee restored the two time delay relays in the correct configuration and immediately verified that the remaining time delay relays inside the Division 3 DG Control Panel were in their proper locations.
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| The finding was of more than minor significance because, if left uncorrected,it would potentially lead to a more significant safety concern (i.e., the inoperability of risk-significant plant safety systems). In addition, based on review of Example 5c in IMC 0612, "Power Reactor Inspection Reports," Appendix E, "Examples of Minor Issues," the issue would not be considered to be of minor significance because the incorrect relays were installed in the control panel.
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| Although the finding resulted in a loss of safetyfunction for the Division 3 DG and high pressure core spray system, it was determined to be of very low safety significance during a Phase 2 Significance Determination Process Review considering the very limited conditions (i.e., only 45 seconds following shutdown of the engine concurrent with a design basis accident) when the Division 3 DG was incapable of performing its safety function. The resultant exposure time was estimated to be about 27 minutes during the 2-year period. The inspectors concluded that this
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| finding affected the cross-cutting area of human performance because the licensee did not effectively communicate expectations regarding procedural compliance and, as a result, maintenance technicians did not follow their procedures by installing nonconforming components and restoring the safety system to service. (IMC 0305 H.4(b))
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| Inspection Report# : 2009005 (pdf)
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| Significance: Sep 30, 2009 Identified By: Self-Revealing Item Type: FIN Finding INEFFECTIVE CORRECTIVE ACTIONS FOR VIBRATION INDUCED STEM/DISC SEPARATION OF FUEL POOL COOLING SYSTEM TRAIN 'A' FLOW CONTROL VALVE 1FC004A.
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| A finding of very low safety significance was self-revealed on May 27, 2009, when fuel pool cooling system flow control valve 1FC004A failed closed. The licensee failed to implement effective corrective actions in response to the same failure mode for the valve that occurred on November 21, 2005. This resulted in the failure of 1FC0014A once again and the subsequent loss of inventory from the containment upper pool and inoperability of the suppression pool makeup system. The licensee entered this issue into its corrective action program to investigate the cause and to identify appropriate corrective actions. No violation of regulatory requirements was identified.
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| The finding was of more than minor significance because it was associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone and directly affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.
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| Specifically, the May 2009 valve failure resulted in a loss of inventory from the containment upper pool and inoperability of the suppression pool makeup system, therefore impacting its availability for certain initiating events.
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| The finding was of very low safety significance because the issue: (1) was not a design or qualification deficiency; (2) did not represent an actual loss of safety function of a system; (3) did not represent an actual loss of safety function of a single train for greater than its Technical Specification (TS) allowed outage time; (4) did not represent an actual loss of safety function of one or more non-TS trains of equipment designated as risk significant; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors did not identify a cross-cutting area component related to this finding.
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| Inspection Report# : 2009004 (pdf)
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| Significance: Jun 30, 2009 Identified By: NRC Item Type: FIN Finding FAILURE TO EVALUTE SAFETY FUNCTION OF SUPPRESSION POOL MAKEUP SYSTEM The inspectors identified a finding of very low safety significance associated with the licensee's failure to recognize a potential loss of safety function for the suppression pool makeup system following the loss of upper containment pool inventory when spent fuel pool cooling system flow control valve 1FC004A failed closed. No evaluation was performed to ensure that the suppression pool makeup system's safety function would be fulfilled with less than Technical Specification (TS) minimum containment upper pool level. The licensee subsequently performed an evaluation and determined that sufficient margin existed such that the system would have been able to fulfill its safety function with limited margin. Corrective actions to address the inadequate reportability review included training for licensed senior reactor operators and development of a formal operability/reportability review process template. No violation of regulatory requirements was identified.
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| The finding would become a more significant safety concern if left uncorrected and was therefore, more than a minor concern. Specifically, the failure to correctly recognize and evaluate a potential loss of a safety function of systems, structures, and components when performing operability or past operability evaluations could reasonably result in an unrecognized condition of a system failing to fulfill its safety-related function. Because the suppression pool makeup system was primarily associated with long term decay heat removal following certain design basis accidents, the inspectors concluded that this issue was associated with the Mitigating Systems cornerstone. The finding was of very low safety significance because the issue: (1) was not a design or qualification deficiency; (2) did not represent an actual loss of safety function of a system; (3) did not represent an actual loss of safety function of a single train for greater than its TS allowed outage time; (4) did not represent an actual loss of safety function of one or more non-TS trains of equipment designated as risk significant; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors concluded that this finding affected the cross-
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| cutting area of human performance because the licensee did not have a formal process in place with adequate guidance and training to enable licensed senior reactor operators, whose resonsibility it was to evaluate a potential loss of safety function, to correctly do so. As a result, senior reactor operators did not adequately review the TS Bases to understand and evaluate whether the system was able to fullfill its safety function. (IMC 0305 H.1(a))
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| Inspection Report# : 2009003 (pdf)
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| Inspection Report# : 2009005 (pdf)
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| Significance: Jun 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PERFROM SURVEILLANCE TESTING ON THE DIVISION 3 SHUTDOWN SERVICE WATER PUMP WITH ADEQUATE MEASURING AND TEST EQUIPMENT.
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| The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criteria XII, "Control of Measuring and Test Equipment," and 10 CFR 50, Appendix B, Criteria XI, "Test Control." The licensee failed to perform surveillance testing on the Division 3 shutdown service water pump with a lake level gage that was properly controlled and adjusted to ensure that it was readable within the range it was used. The licensee subsequently replaced the unreadable lake level gage section with one that was readable and implemented additional corrective actions to address a lapse in operations standards.
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| The inspectors concluded that this finding would become a more significant safety concern if left uncorrected and it was therefore more than a minor concern. Specifically, the failure to perform surveillance testing with properly controlled and accurate measuring and test equipment could reasonably result in the failure to identify degraded or inoperable safety-related components. Because the shutdown service water system was primarily associated with long term decay heat removal following certain design basis accidents, the inspectors concluded that this issue was associated with the Mitigating Systems Cornerstone. The finding was of very low safety significance because the issue was a design or qualification deficiency confirmed not to result in loss of operability or availability. The inspectors concluded that this finding affected the cross-cutting area of problem identification and resolution because the licensee was not properly maintaining the lake level gage to ensure that it would remain usable and did not correct the degraded level gage in a timely manner after it was identified. As a result, operators accepted the degraded level gage for continued use. (IMC 0305 P.1(d))
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| Inspection Report# : 2009003 (pdf)
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| Barrier Integrity Significance: Mar 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO RECOGNIZE EXAMINATION LIMITATIONS FOR A CONTAINMENT PENETRATION WELD The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," for the licensee's failure to follow procedure instructions and record examination limitations for containment pipe-to-penetration weld 1-MS-B-11. The licensee subsequently documented the failure to record th 1-MS-B-11 limited weld examination in the corrective action program. The licensee planned to submit limited containment pipe-to-penetration weld examinations to the NRC for review and approval.
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| The finding was of more than minor significance because, if left uncorrected, the failure to document limited weld examinations could become a more significant safety concern. Absent NRC identification, the licensee would not have submitted limited weld examinations to the NRC for approval. Further, the inspector could not determine if the NRC would approve the limited weld surface examinations without a licensee evaluation for the extent of additional coverage possible with volumetric weld examinations. This finding was of very low safey-significance based on answering "no" to each of the Phase 1 screening questions identified in the Containment Barrier column of Table 4a in 609.04, "Phase 1 - Initial Screening and Characterization of Findings." Specifically, this finding did not
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| represent an actual open pathway in the physical integrity of reactor containment. This finding has a cross-cutting aspect in the area of Human Performance, Resources because the licensee did not provide complete, accurate and up-to-date design documents (weld construction drawing) to the non-destructive examination staff. Specifically, the lack of a weld construction drawing which included the weld profile appeared to have contributed to the examination staff's failure to recognize that they had not completely examined the required weld surfaces.
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| Inspection Report# : 2010002 (pdf)
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| Significance: Mar 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE TEST CRITERIA IN STANDBY GAS TREATMENT SYSTEM FLOW/HEATER OPERABILITY SURVEILLANCE TEST The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criteria V, "Instructions, Procedures, and Drawings." The licensee failed to include appropriate quantitative or qualitative acceptance criteria in its surveillance test procedure for fulfilling the monthly surveillance requirement to demonstrate operability of the standby gas treatment (SGT) system as described in the Technical Specification Bases. As corrective action, the licensee revised the procedure to include acceptance criteria that system flow is normal and that no blockage, fan or motor failure, or excessive vibration is detected.
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| The finding was of more than minor signifcance because it is associated with the Procedure Quality cornerstone attribute for the Control Room and Auxiliary Building and adversely affected the Barrier Integrity cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, by not providing appropriate acceptance criteria by which the operability of the SGT system trains could be assessed, the ability of the SGT system to collect and treat the design leakage of radionuclides from the primary containment to the secondary containment during an accident could not be assured.
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| The inspectors did not identify a cross-cutting aspect related to this finding.
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| Inspection Report# : 2010002 (pdf)
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| Significance: SL-IV Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO UPDATE THE FINAL SAFETY ANALYSIS REPORT.
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| The inspectors identified a Non-Cited Violation of 10 CFR 50.71, "Maintenance of Records, Making of Reports,"
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| associated with the licensee's failure to correctly update the Updated Fainal Safety Analysis Report (UFSAR) when modifying Technical Specification (TS) requirements for the Control Room ventilation system during implementation of Improved Standard Technical Specifications. Specifically, the licensee failed to change the specified safety function description for the system to maintain positive pressure within the Control Room envelope with respect to adjacent areas during all operating modes except when the system is in the recirculation mode or when the system is in the maximum outside air purge mode. This directly contributed to the licensee's failure to correctly evaluate the operability of Control Room ventilation system Train 'B' when the system was unable to maintain the Control Room envelope at a positive pressure relative to adjacent areas while operating in the normal mode. Subsequent evaluation by the inspectors determined that the safety function description in the UFSAR was inaccurate and the system was operable with the degraded/nonconforming condition. The licensee entered this violation into its corrective action program to investigate the cause and to identify appropriate corrective actions.
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| Because the issue affected the NRC's ability to perform its regulatory function, the violation was reviewed under the traditional enforcement process; however, the underlying technical issue was evaluated using the Significance Determination Process. The finding would become a more significant safety concern if left uncorrected and was therefore more than a minor concern. Specifically, the failure to correctly evaluate a degraded/nonconforming condition potentially affecting the operability of a system, structure, or component (SSC) required to be operable by TS could reasonably result in an unrecognized condition of an SSC failing to fulfill a safety-related function. Because the Control Room ventilation system supports the radiological barrier function to protect operators inside the Control Room following certain design basis accidents, the inspectors concluded that this issue was associated with the Barrier Integrity Cornerstone. The finding was of very low safety significance because it involved only a degradation of the radiological barrier function provided for the Control Room. The inspectors did not identify a cross-cutting aspect related to this finding.
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| Inspection Report# : 2009004 (pdf)
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| Significance: Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ADEQUATELY IMPLEMENT REQUIREMENTS OF THE LEAKAGE REDUCTION AND MONITORING PROGRAM.
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| The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criteria V, "Instructions, Procedures, and Drawings," regarding the licensee's failure to adequately implement periodic visual inspection requirements to monitor and minimize leakage from piping systems connecting to the reactor coolant system. The licensee entered this violation into its corrective action program to investigate the cause and to identify appropriate corrective actions.
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| The finding would become a more significant safety concern if left uncorrected and was therefore more than a minor concern. Specifically, the failure to adequately implement required leakage reduction and monitoring program controls to minimize leakage from reactor coolant sources outside of containment that could contain highly radioactive fluids during a serious transient or accident could reasonably result in higher doses to plant workers and higher potential offsite release levels. Because the leakage reducation and monitoring program is intended to contain highly radioactive fluids within piping systems outside containment, which supports the radiological barrier functions to protect plant workers and the public following serious transients or accidents, the inspectors concluded that this issue was associated with the Barrier Integrity Cornerstone. The finding was of very low safety significance because it invovled only a degradation of the radiological barrier function provided for the Auxiliary Building. The inspectors concluded that this finding affected the cross-cutting area of human performance because the licensee did not provide adequate procedural guidance and training to enable operators to correctly perform and document piping system visual inspections to implement its leakage reduction and monitoring program. As a result, the licensee did not have appropriate objective quality evidence to demonstrate that the program requirements were met. (IMC 0305 H.2(c))
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| Inspection Report# : 2009004 (pdf)
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| Emergency Preparedness Significance: SL-IV Nov 20, 2009 Identified By: NRC Item Type: NCV NonCited Violation Implementation of a Change which Decreased the Effectiveness of the Emergency Plan The inspectors identified a NCV of 10 CFR 50.54(q) associated with 10 CFR 50.47(b)(2) because the licensee failed to obtain prior NRC approval for a change made to its emergency plan that decreased the effectiveness of the plan.
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| Specifically, the licensee removed staffing and capabilities from the minimum on-shift emergency response staffing requirements from the Clinton Power Station Emergency Plan Annex, Section 2, Table B-1. The licensee entered this issue into their corrective action program and replaced staffing back on-shift as required by the 1998 emergency plan annex.
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| This finding was more than minor and of very low safety-significance using IMC 0609, Appendix B, because the finding was associated with the Emergency Preparedness Cornerstone attribute of emergency response organization readiness for minimum on shift emergency response staffing. Because the finding affected the NRC's ability to perform its regulatory function, the inspectors evaluated the significance using the traditional enforcement process.
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| This finding was determined to be a Severity Level IV violation because the licensee failed to meet an emergency planning requirement not directly related to assessment and notification.
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| Inspection Report# : 2009502 (pdf)
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| Inspection Report# : 2010002 (pdf)
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| Occupational Radiation Safety
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| Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.
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| Miscellaneous Significance: N/A Apr 17, 2009 Identified By: NRC Item Type: FIN Finding Biennial PI&R Inspection Summary The inspectors concluded that the implementation of the corrective action program (CAP) at Clinton was generally good. The licensee had a low threshold for identifying station problems and entering them into the CAP. In addition, the station was effective at incorporating operating experience reports into the CAP. The inspectors determined that issues were generally effectively screened and prioritized in a timely manner using established criteria based on plant risk and uncertainty. Casual evaluations sampled were of sufficient depth, considered extent of condition, generic issues, and previous occurrences. Corrective actions program assignments were generally completed in a timely and accurate manner. The team noted that station effectiveness reviews, audits, and self assessment were generally thorough and effective at identifying unrecognized weakness. The inspectors concluded that station employees appeared to be willing to express safety concerns through established processes and a healthy safety conscious work environment (SCWE) existed at the station.
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| Inspection Report# : 2009007 (pdf)
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| Last modified : May 26, 2010
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| Clinton 2Q/2010 Plant Inspection Findings Initiating Events Significance: Mar 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO CONTROL TRANSIENT COMBUSTIBLE MATERIALS IN ACCORDANCE WITH FIRE PROTECTION PROGRAM The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of the Clinton Power Station Unit 1 Operating License (NPF-62, Section 2.F). The licensee failed to implement the Fire Protection Program in accordance with program requirements by failing to follow approved Fire Protection Program procedures for the control of transient combustible materials. The licensee promptly removed the transient combustible materials found by the inspectors.
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| The inspectors concluded that this finding could be reasonably viewed as a precursor to a significant event (i.e., a fire affecting more than one train of safe shutdown equipment). Specifically, the presence of transient combustible materials in a combustible free zone could reasonably result in degradation of the fire protection defense-in-depth elements in place to prevent fires from starting and mitigate the consequences of fires. In addition, based on review of Example 4k in IMC 0612, "Power Reactor Inspection Reports," Appendix E, "Examples of Minor Issues," the issue would not be considered to be of minor significance because the identified transient combustibles were found in a combustible free zone required for separation of redundant trains. The finding was of very low safety significance because the items found in the combustible free zone would not be considered transient combustibles of significance as defined in IMC 0609, Appendix F, "Fire Protection Significance Determination Process," Attachment 2, "Degradation Rating Guidance Specific to Various Fire Protection Program Elements," and therefore the issue was assigned a "low degradation" rating. The inspectors concluded that this finding affected the cross-cutting area of problem identification and resolution. Specifically, the licensee missed an opportunity to identify and remove the transient combustible materials while implementing corrective actions for previous inspector identified findings involving the control of transient combustible materials.
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| Inspection Report# : 2010002 (pdf)
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| Significance: Mar 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO CONTROL COMBUSTIBLE GAS CYLINDERS IN ACCORDANCE WITH FIRE PROTECTION PROGRAM The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of the Clinton Power Station Unit 1 Operating License (NPF-62, Section 2.F). The licensee failed to implement the Fire Protection Program in accordance with program requirements by failing to follow approved Fire Protection Program procedures for the control of combustible gas cylinders in the plant. The licensee promptly removed the combustible gas cylinders found by the inspectors.
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| The inspectors concluded that this finding was associated with the Protection Against External Factors attribute of the Initiating Events Cornerstone and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the fire hazard for the affected area was increased by the uncontrolled presence of the compressed gas cylinders. In addition, based on review of Example 4k in Inspection Manual Chapter 0612, "Power Reactor Inspection Reports,"
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| Appendix E, "Examples of Minor Issues," the issue would not be considered to be of minor significance because a credible fire scenario involving the identified transient combustibles could affect equipment important to safety. The finding was detemined to be of very low safety significance during a Phase 3 Significance Determination Process review since the delta core damage frequency was determined to be negligible. Because a postulated fire in the area
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| where the combustible gas cylinders were found could affect only one train of safe shutdown equipment, the safe shutdown path was not affected by the finding. The inspectors concluded that this finding affected the cross-cutting area of human performance. Specifically, the licensee did not adquately ensure that supervisory and management oversight of work activities involving contractors supported nuclear safety.
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| Inspection Report# : 2010002 (pdf)
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| Significance: Mar 31, 2010 Identified By: Self-Revealing Item Type: FIN Finding FAILURE TO CORRECT INADEQUATE FWLCS RESPONSE RESULTED IN HIGH REACTOR VESSEL WATER LEVEL (LEVEL 8 ) SCRAM A finding of very low safety significance was self-revealed from an event that resulted in a Unit 1 reactor scram. The licensee failed to correct a non-conforming condition with inadequate response from the feedwater level control system (FWLCS) that caused an automatic reactor scram on February 10, 2008, following an unexpected loss of a reactor recirculation pump. This resulted in a second reactor scram for the same cause on October 15, 2009, following the unexpected loss of a reactor recirculation pump. Because the FWLCS is not safety-related, no violation of regulatory requirements was identified. The FWLCS response was corrected in January 2010 and proper system response was verified by the licensee upon start up from the January-February 2010 refueling outage.
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| The finding was of more than minor significance because this issue was associated with the Equipment Performance attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during power operations.
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| Specifically, inadequate FWLCS response resulted in a reactor scram following the unexpected loss of a reactor recirculation pump. The finding was of very low safety significance because the issue: (1) did not contribute to the likelihood of a primary or secondary system loss-of-coolant-accident initiator, (2) did not contribute to both the likelihood of a reactor trip AND the likelihood that mitigation equipment or functions would not be available, and (3) did not increase the likelihood of a fire or internal/external flooding event. The inspectors did not identify a cross cutting aspect related to this finding.
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| Inspection Report# : 2010002 (pdf)
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| Significance: Dec 31, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation FAILURE TO CORRECTLY TORQUE VALVE PACKING GLAND NUTS RESULTED IN VALVE PACKING FAILURE AND UNPLANNED PLANT SHUTDOWN A finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criteria V, "Instructions, Procedures, and Drawings," was self-revealed on September 29, 2009, when a steam leak developed from the reactor core isolation cooling (RCIC) system inboard steam isolation valve (1E51F0063) stem packing. This resulted in a plant shutdown due to a greater than 2 gallon-per-minute increase in unidentified reactor coolant system (RCS) leakage within the previous 24 hours. The licensee failed to correctly tighten the valve packing gland nuts to the as-left torque valve from original packing installation when performing scheduled maintenance to verify the as-found torque value. The licensee replaced the 1E51F0063 valve stem packing during the subsequent forced outage and tightened the gland nuts to the correct torque value.
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| The finding was of more than minor significance because it was associated with the Equipment Performance attribute of the Initiating Events Cornerstone and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the failure to correctly tighten the valve stem packing gland nuts resulted in stem packing failure and a subsequent plant shutdown due to exceeding the Technical Specification (TS) limit for an increase in unidentified RCS leakage.
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| Although the finding resulted in exceeding the TS limit for RCS leakage, it was determined to be of very low safety significance during a Phase 2 Significance Determination Process review because there was no loss of mitigation capability for any safety system and therefore no resultant change in core damage frequency. Because the performance issue was associated with maintenance performed in February 2006, it did not necessarily reflect current licensee performance and no cross-cutting aspect was identified.
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| Inspection Report# : 2009005 (pdf)
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| Mitigating Systems Significance: SL-IV Jun 30, 2010 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO SATISFY 10 CFR 50.72 AND 50.73 REPORTING REQUIREMENTS THUS AFFECTING THE REGULATORY PROCESS..
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| The inspectors identified a Severity Level IV Non- Cited Violation of the NRCs reporting requirements in 10 CFR 50.72(a)(1), Immediate Notification Requirements for Operating Nuclear Power Reactors, and 10 CFR 50.73(a)(1),
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| Licensee Event Report System. The licensee failed make a required 8-hour non-emergency notification call to the NRC Operations Center and failed to submit a required Licensee Event Report within 60 days after discovery of a condition that resulted in the plant being in an unanalyzed condition that significantly degraded plant safety and could have prevented fulfillment of the safety function of the emergency core cooling system. No immediate corrective actions were taken to address this finding; however, the licensee entered this issue into its corrective action program for evaluation.
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| This violation was of more than minor significance because the NRC relies on licensees to identify and report conditions or events meeting the criteria specified in the Technical Specifications and the regulations in order to perform its regulatory function. Because this issue affected the NRC's ability to perform its regulatory function, the inspectors evaluated it using the traditional enforcement process. The underlying technical issue (i.e., interconnecting floor drains between the Residual Heat Removal A Pump Room and the Radwaste Pipe Tunnel) was determined to be a finding of very low safety significance during a Phase 3 SDP evaluation. Consistent with the guidance in Supplement I, Paragraph D.4, of the NRC Enforcement Policy, the violation associated with this finding was determined to be a Severity Level IV Violation.
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| The related performance deficiency is tracked as item 2010-003-06.
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| Inspection Report# : 2010003 (pdf)
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| Significance: Jun 30, 2010 Identified By: NRC Item Type: FIN Finding OPERABILITY ASSESSMENT OF INSERVICE TESTING SURVEILLANCE DISCREPANCIES FOR EXCESS FLOW CHECK VALVES The inspectors identified a finding of very low safety significance associated with the licensee's failure to evaluate the functionality of multiple excess flow check valves that had not been tested in accordance with the American Society of Mechanical Engineers / American National Standards Institute (ASME/ANSI) Code Inservice Testing requirements to establish whether the nonconforming condition warranted starting the Technical Specification (TS) action time for the suppression pool makeup (SPMU) system. In response the the inspectors' questions, the licensee subsequently performed an operability evaluation. No violation of regulatory requirements was identified because subsequent testing by the licnesee determined that the valves were functional.
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| The finding would become a more significant safety concern if left uncorrected and was therefore more than a minor concern. Specifically, the failure to correctly evaluate a degraded/nonconforming condition potentially affecting the operability of structures, systems, or components (SSCs) required to be operable by TS could reasonably result in an unrecognized condition of an SSC failing to fulfill a safety-related function. Because the SPMU system was primarily assoicated with long term decay heat removal following certain design basis accidents, the inspectors concluded that this issue was associated with the Mitigating Systems Cornerstone. The finding was of very low safety significance because the issue: (1) was not a design or qualification deficiency; (2) did not represent an actual loss of safety function of a system; (3) did not represent an actual loss of safety function of a single train for greater than its TS allowed outage time; (4) did not represent an actual loss of safety function of one or more non-TS trains of equipment designated as risk significant; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors concluded that this finding affected the cross-cutting area of human
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| pefformance because the licenesee did not have a formal process in place with adequate guidance and training to enable licensed senior reactor operators to properly and promptly evaluate operability in this instance. As a result, senior reactor operators took it for granted that utilizing the relief allowed by TS Surveillance Requirement 3.0.3 and performing a risk evaluation obviated the need to address the operability of the instrumentation supported by the excess flow check valves for the ASME/ANSI Code noncompliance.
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| Inspection Report# : 2010003 (pdf)
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| Significance: Jun 30, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Satisfy 10 CFR 50.72 and 50.73 Reporting Requirements - performance deficiiency portion.
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| The inspectors identified a finding of very low safety significance of the NRCs reporting requirements in 10 CFR 50.72(a)(1), Immediate Notification Requirements for Operating Nuclear Power Reactors, and 10 CFR 50.73(a)(1),
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| Licensee Event Report System. The licensee failed to make a required 8 hour non emergency notification call to the NRC Operations Center and failed to submit a required Licensee Event Report within 60 days after discovery on October 7, 2009, of a condition that resulted in the plant being in an unanalyzed condition that significantly degraded plant safety and could have prevented fulfillment of the safety function of the emergency core cooling system. No immediate corrective actions were taken to address this finding; however, the licensee entered this issue into its corrective action program for evaluation.
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| This finding was of more than minor significance because the NRC relies on licensees to identify and report conditions or events meeting the criteria specified in the Technical Specifications and the regulations in order to perform its regulatory function. The inspectors assessed the significance of the underlying performance deficiency using the SDP. The underlying technical issue (i.e., interconnecting floor drains between the Residual Heat Removal A Pump Room and the Radwaste Pipe Tunnel) was determined to be a finding of very low safety significance (green) during a Phase 3 Significance Determination Process evaluation. This finding affected the cross cutting area of human performance because the licensee did not use conservative assumptions in decision making while evaluating the reportability of the unanalyzed condition with respect to the reporting requirements in 10 CFR 50.72(a)(1)(ii) and 50.73(a)(1). (IMC 0310 H.1(b)) (Section 1R06.1.b.(1))
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| The related traditional enforcment portion is tracked as item 2010-003-01.
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| Inspection Report# : 2010003 (pdf)
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| Significance: Jun 25, 2010 Identified By: NRC Item Type: NCV NonCited Violation Non Conservative Acceptance Criteria for RHR Pump Performance Testing The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, having very low safety-significance for the licensees failure to ensure adequate acceptance limits were incorporated into test procedures. Specifically, the licensee failed to properly consider instrument loop uncertainties and allowable emergency diesel generator frequency variance when determining the alert and required action values used in the inservice test procedure for testing of the residual heat removal pumps. Consequently, the acceptance criteria for the lower limits on degradation of pump head were non-conservative. This finding was entered into the licensees corrective action program and a preliminary calculation performed by the licensee concluded that the pumps were operable.
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| The finding was more than minor because it was associated with the Mitigating Systems cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the capability of the system to respond to initiating events to prevent undesirable consequences. This finding was of very low safety-significance (Green) because the licensee was able to demonstrate pump operability and therefore there was no loss of safety function. This finding had a cross-cutting aspect in the area of problem identification and resolution because the licensee did not thoroughly evaluate operating experience that included similar issues relating to the failure to appropriately account for instrument uncertainties in design analysis.
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| Inspection Report# : 2010006 (pdf)
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| Significance: Jun 25, 2010 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Test Control of RHR Heat Exchangers The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, having very low safety-significance for the licensees failure to establish test conditions to assure that the 1B residual heat removal heat exchanger would perform satisfactorily in service under accident conditions. Specifically, the inspectors determined that the heat exchanger thermal performance test procedure did not assure adequate temperature differences to provide reliable test results. In addition, the most recent test was performed with lower temperature differences than those identified in plant calculations. This finding was entered into the licensees corrective action program and a preliminary analysis performed by the licensee concluded the test results were acceptable.
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| The finding was more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the residual heat removal heat exchanger performance test procedure did not establish appropriate test conditions to ensure that the component would perform its required function during an accident. Also, the inspectors determined that the finding was similar to Examples 3.j and 3.k of IMC 612, Appendix E, in that there was a reasonable doubt of the operability of the component based on the most recent test conditions. The inspectors determined the finding was of very low safety significance (Green) because it was not a design or qualification deficiency, did not represent a loss of system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding did not have a cross-cutting aspect because it did not represent current performance.
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| Inspection Report# : 2010006 (pdf)
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| Significance: Mar 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation INTERCONNECTING FLOOR DRAINS BETWEEN THE RESIDUAL HEAT REMOVAL 'A' PUMP ROOM AND RADWASTE PIPE TUNNEL The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criteria III, "Design Control," regarding the licensee's failure to correctly translate the design basis into the design of the Auxiliary Building floor drain system with appropriate margin. The inspectors identified that floor drains in the Residual Heat Removal (RHR) 'A' Pump Room and the Radwaste Pipe Tunnel were interconnected, which resulted in the plant being in an unanalyzed condition that degraded plant safety and could have prevented fulfillment of the safety function of the containment suppression pool. To address the immediate operability concern, the licensee plugged the two floor drains in the Radwaste Pipe Tunnel line to prevent communication with the floor drain system in the RHR 'A' Pump Room. An exposed vertical section of the drain line was then cut and a solid steel plate welded into the pipe per an engineering design change to permanently isolate the floor drains between the two rooms.
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| The finding was of more than minor significance because it was associated with the Design Control attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage).
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| Specifically, the as-found configuration of the interconnecting floor drains resulted in the plant being in an unanalyzed condition that could have prevented fullfillment of the safety function of the containment suppression pool. Although the finding would represent a loss of safety function in the event of a postulated accident, it was determined to be of very low safety significance during a Phase 3 Significance Determination Process review because the delta core damage frequency was determined to be negligible since the initiating event frequency for flooding due to an RHR pump suction pipe failure was sufficiently low. Because this condition had existed since initial plant construction, the performance issue did not necessarily reflect current licensee performance and no cross-cutting aspect was identified.
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| Inspection Report# : 2010002 (pdf)
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| Significance: Dec 31, 2009 Identified By: Self-Revealing
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| Item Type: NCV NonCited Violation FAILURE TO CORRECTLY INSTALL RELAYS INSIDE OF THE DIVISION 3 DIESEL GENERATOR CONTROL PANEL A finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criteria V, "Instructions, Procedures, and Drawings," was self-revealed on September 23, 2009, when the Division 3 diesel generator (DG) was found to have had two components installed incorrectly. Electrical maintenance technicians had incorrectly replaced time delay relays K-8A and K-32 on September 24, 2007, essentially swapping the locations of the two relays. This rendered the Division 3 DG inoperable for about 2 years and resulted in a loss of safety function for the Division 3 DG and high pressure core spray system under a certain sequence of initiating events. The licensee restored the two time delay relays in the correct configuration and immediately verified that the remaining time delay relays inside the Division 3 DG Control Panel were in their proper locations.
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| The finding was of more than minor significance because, if left uncorrected,it would potentially lead to a more significant safety concern (i.e., the inoperability of risk-significant plant safety systems). In addition, based on review of Example 5c in IMC 0612, "Power Reactor Inspection Reports," Appendix E, "Examples of Minor Issues," the issue would not be considered to be of minor significance because the incorrect relays were installed in the control panel.
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| Although the finding resulted in a loss of safety function for the Division 3 DG and high pressure core spray system, it was determined to be of very low safety significance during a Phase 2 Significance Determination Process Review considering the very limited conditions (i.e., only 45 seconds following shutdown of the engine concurrent with a design basis accident) when the Division 3 DG was incapable of performing its safety function. The resultant exposure time was estimated to be about 27 minutes during the 2-year period. The inspectors concluded that this finding affected the cross-cutting area of human performance because the licensee did not effectively communicate expectations regarding procedural compliance and, as a result, maintenance technicians did not follow their procedures by installing nonconforming components and restoring the safety system to service.
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| Inspection Report# : 2009005 (pdf)
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| Significance: Sep 30, 2009 Identified By: Self-Revealing Item Type: FIN Finding INEFFECTIVE CORRECTIVE ACTIONS FOR VIBRATION INDUCED STEM/DISC SEPARATION OF FUEL POOL COOLING SYSTEM TRAIN 'A' FLOW CONTROL VALVE 1FC004A.
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| A finding of very low safety significance was self-revealed on May 27, 2009, when fuel pool cooling system flow control valve 1FC004A failed closed. The licensee failed to implement effective corrective actions in response to the same failure mode for the valve that occurred on November 21, 2005. This resulted in the failure of 1FC0014A once again and the subsequent loss of inventory from the containment upper pool and inoperability of the suppression pool makeup system. The licensee entered this issue into its corrective action program to investigate the cause and to identify appropriate corrective actions. No violation of regulatory requirements was identified.
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| The finding was of more than minor significance because it was associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone and directly affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.
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| Specifically, the May 2009 valve failure resulted in a loss of inventory from the containment upper pool and inoperability of the suppression pool makeup system, therefore impacting its availability for certain initiating events.
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| The finding was of very low safety significance because the issue: (1) was not a design or qualification deficiency; (2) did not represent an actual loss of safety function of a system; (3) did not represent an actual loss of safety function of a single train for greater than its Technical Specification (TS) allowed outage time; (4) did not represent an actual loss of safety function of one or more non-TS trains of equipment designated as risk significant; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors did not identify a cross-cutting area component related to this finding.
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| Inspection Report# : 2009004 (pdf)
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| Barrier Integrity Significance: SL-IV Jun 30, 2010
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| Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PERFORM AN ADEQUATE 10 CFR 50.59 EVALUATION FOR CPS PROCEDURE 3711.01 THUS AFFECTING THE NRC'S REGULATORY PROCESS.
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| The inspectors identified a Non-Cited Violation of 10 CFR 50.59, Changes, Tests and Experiments. The licensee failed to perform an adequate 10 CFR 50.59 evaluation and obtain a license amendment prior to implementing CPS 3711.01, CPS [Clinton Power Station] Operations with the Potential to Drain the Reactor Vessel [OPDRV],
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| Revision 0. The procedure established a definition of an OPDRV for use in determining the applicability of several Technical Specification (TS) requirements while in Modes 4 and 5. The licensee failed to recognize that implementing this new procedure, in effect, constituted a change to the TS incorporated into its licensing basis, which would therefore require a license amendment pursuant to 10 CFR 50.59(c)(1)(i) and 10 CFR 50.90. No immediate corrective actions were taken to address this finding; however, the licensee entered this issue into its corrective action program for evaluation.
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| The finding was of more than minor significance because there was a reasonable likelihood that the change requiring a 10 CFR 50.59 evaluation would require NRC review and approval prior to implementation. Because this issue affected the NRC's ability to perform its regulatory function, the inspectors evaluated it using the traditional enforcement process. Based on the results of a modified Phase 2 SDP evaluation, this finding was determined to be of very low safety significance. Consistent with the guidance in Supplement I, Paragraph D.5, of the NRC Enforcement Policy, the violation associated with this finding was determined to be a Severity Level IV Violation.
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| The related performance deficiency is tracked as item 2010-003-07.
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| Inspection Report# : 2010003 (pdf)
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| Significance: Jun 30, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation FAILURE TO FOLLOW PROCEDURE RESULTING IN GATE SEAL LEAKAGE.
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| A finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criteria V, Instructions, Procedures, and Drawings, was self-revealed on January 29, 2010, when the dryer cavity gate seal depressurized during the performance of the containment and reactor vessel isolation functional surveillance procedure. When the seal lost pressure, approximately 46,500 gallons of water leaked from the dryer cavity pool into the reactor cavity. In response to the event, the licensee ensured all personnel were out of the reactor cavity, entered its radioactive spill off-normal procedure, and re-established air pressure to the dryer cavity gate seal. Subsequent investigation revealed that during the gate seal inflation procedure the proper valve operation sequence was not followed. As corrective action, the licensee revised many of its procedures and included a special brief to the refueling outage preparation for Reactor Services personnel.
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| The finding was of more than minor significance because the licensees failure to correctly install the upper containment dryer cavity gate could be reasonably viewed as a precursor to a significant event and, if left uncorrected would potentially lead to a more significant safety concern (i.e., increased dose or personnel contamination). In addition, the finding was similar to Example 4c in Inspection Manual Chapter 0612, Power Reactor Inspection Reports, Appendix E, Examples of Minor Issues, in that data recorded during installation of the dryer cavity gate seal was incorrect and resulted in backup air bottle supply pressure left outside the acceptable range. Because the dryer cavity gate seal is intended to contain highly radioactive fluids within containment, which supports the radiological barrier functions to protect plant workers and the public following serious transients or accidents, the inspectors concluded that this issue was associated with the Barrier Integrity Cornerstone. Although this event resulted in a loss of inventory from the dryer cavity pool and partial flooding of the lower reactor cavity and drywell, it was determined to be of very low safety significance because there was no loss inventory from the reactor vessel and it could not result in the loss of reactor coolant system level instrumentation. The inspectors concluded that this finding affected the cross-cutting area of human performance. The licensee did not effectively communicate expectations regarding procedural compliance in this instance and, as a result, the Reactor Services maintenance craftsman did not correctly follow the procedure by performing steps out of sequence and restoring a system to service that was incorrectly aligned. (IMC 0310 H.4(b))
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| Inspection Report# : 2010003 (pdf)
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| Significance: Jun 30, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an Adequate 10 CFR 50.59 Evaluation for CPS Procedure 3711.01 - performance deficiency portion The inspectors identified a finding of very low safety significance with an associated NCV of 10 CFR 50.59, Changes, Tests and Experiments. The licensee failed to perform an adequate 10 CFR 50.59 evaluation and obtain a license amendment prior to implementing CPS 3711.01, CPS [Clinton Power Station] Operations with the Potential to Drain the Reactor Vessel [OPDRV], Revision 0 on January 11, 2010. The procedure established a definition of an OPDRV for use in determining the applicability of several TS requirements while in Modes 4 and 5. The licensee failed to recognize that implementing this new procedure, in effect, constituted a change to the TS incorporated into its licensing basis, which would, therefore, require a license amendment pursuant to 10 CFR 50.59(c)(1)(i) and 10 CFR 50.90. No immediate corrective actions were taken to address this finding; however, the licensee entered this issue into its corrective action program for evaluation.
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| The finding was of more than minor significance because there was a reasonable likelihood that the change requiring a 10 CFR 50.59 evaluation would require NRC review and approval prior to implementation. The inspectors assessed the significance of the underlying issue using the SDP. Based on the results of a modified Phase 2 SDP evaluation, this finding was determined to be of very low safety significance. The inspectors concluded that this finding affected the cross cutting area of human performance. Specifically, the licensee did not use conservative decision making to demonstrate that the proposed action did not require prior NRC approval. The inspectors noted that the licensee was aware of potential concerns regarding the new procedure prior to completing the initial 10 CFR 50.59 evaluation and again prior to revising the evaluation in response to concerns raised by the inspectors; however, the incorrect conclusion was reached in both revisions of the evaluation that the new procedure was not a change to the TS and that a license amendment was not necessary. (IMC 0310 H.1(b)) (Section 1R13.b.(1))
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| The associated traditional enforcment is tracked as item 2010-003-02.
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| Inspection Report# : 2010003 (pdf)
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| Significance: Mar 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO RECOGNIZE EXAMINATION LIMITATIONS FOR A CONTAINMENT PENETRATION WELD The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," for the licensee's failure to follow procedure instructions and record examination limitations for containment pipe-to-penetration weld 1-MS-B-11. The licensee subsequently documented the failure to record the 1-MS-B-11 limited weld examination in the corrective action program. The licensee planned to submit limited containment pipe-to-penetration weld examinations to the NRC for review and approval.
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| The finding was of more than minor significance because, if left uncorrected, the failure to document limited weld examinations could become a more significant safety concern. Absent NRC identification, the licensee would not have submitted limited weld examinations to the NRC for approval. Further, the inspector could not determine if the NRC would approve the limited weld surface examinations without a licensee evaluation for the extent of additional coverage possible with volumetric weld examinations. This finding was of very low safey-significance based on answering "no" to each of the Phase 1 screening questions identified in the Containment Barrier column of Table 4a in 609.04, "Phase 1 - Initial Screening and Characterization of Findings." Specifically, this finding did not represent an actual open pathway in the physical integrity of reactor containment. This finding has a cross-cutting aspect in the area of Human Performance, Resources because the licensee did not provide complete, accurate and up-to-date design documents (weld construction drawing) to the non-destructive examination staff. Specifically, the lack of a weld construction drawing which included the weld profile appeared to have contributed to the examination staff's failure to recognize that they had not completely examined the required weld surfaces.
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| Inspection Report# : 2010002 (pdf)
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| Significance: Mar 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE TEST CRITERIA IN STANDBY GAS TREATMENT SYSTEM FLOW/HEATER OPERABILITY SURVEILLANCE TEST The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criteria V, "Instructions, Procedures, and Drawings." The licensee failed to include appropriate quantitative or qualitative acceptance criteria in its surveillance test procedure for fulfilling the monthly surveillance requirement to demonstrate operability of the standby gas treatment (SGT) system as described in the Technical Specification Bases. As corrective action, the licensee revised the procedure to include acceptance criteria that system flow is normal and that no blockage, fan or motor failure, or excessive vibration is detected.
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| The finding was of more than minor signifcance because it is associated with the Procedure Quality cornerstone attribute for the Control Room and Auxiliary Building and adversely affected the Barrier Integrity cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, by not providing appropriate acceptance criteria by which the operability of the SGT system trains could be assessed, the ability of the SGT system to collect and treat the design leakage of radionuclides from the primary containment to the secondary containment during an accident could not be assured.
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| The inspectors did not identify a cross-cutting aspect related to this finding.
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| Inspection Report# : 2010002 (pdf)
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| Significance: SL-IV Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO UPDATE THE FINAL SAFETY ANALYSIS REPORT.
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| The inspectors identified a Non-Cited Violation of 10 CFR 50.71, "Maintenance of Records, Making of Reports,"
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| associated with the licensee's failure to correctly update the Updated Fainal Safety Analysis Report (UFSAR) when modifying Technical Specification (TS) requirements for the Control Room ventilation system during implementation of Improved Standard Technical Specifications. Specifically, the licensee failed to change the specified safety function description for the system to maintain positive pressure within the Control Room envelope with respect to adjacent areas during all operating modes except when the system is in the recirculation mode or when the system is in the maximum outside air purge mode. This directly contributed to the licensee's failure to correctly evaluate the operability of Control Room ventilation system Train 'B' when the system was unable to maintain the Control Room envelope at a positive pressure relative to adjacent areas while operating in the normal mode. Subsequent evaluation by the inspectors determined that the safety function description in the UFSAR was inaccurate and the system was operable with the degraded/nonconforming condition. The licensee entered this violation into its corrective action program to investigate the cause and to identify appropriate corrective actions.
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| Because the issue affected the NRC's ability to perform its regulatory function, the violation was reviewed under the traditional enforcement process; however, the underlying technical issue was evaluated using the Significance Determination Process. The finding would become a more significant safety concern if left uncorrected and was therefore more than a minor concern. Specifically, the failure to correctly evaluate a degraded/nonconforming condition potentially affecting the operability of a system, structure, or component (SSC) required to be operable by TS could reasonably result in an unrecognized condition of an SSC failing to fulfill a safety-related function. Because the Control Room ventilation system supports the radiological barrier function to protect operators inside the Control Room following certain design basis accidents, the inspectors concluded that this issue was associated with the Barrier Integrity Cornerstone. The finding was of very low safety significance because it involved only a degradation of the radiological barrier function provided for the Control Room. The inspectors did not identify a cross-cutting aspect related to this finding.
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| Inspection Report# : 2009004 (pdf)
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| Significance: Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ADEQUATELY IMPLEMENT REQUIREMENTS OF THE LEAKAGE REDUCTION AND
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| MONITORING PROGRAM.
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| The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criteria V, Instructions, Procedures, and Drawings, regarding the licensees failure to adequately implement periodic visual inspection requirements per procedure CPS 1019.07, Leakage Reduction and Monitoring Program, to monitor and minimize leakage from piping systems connecting to the reactor coolant system. The inspectors also identified that the procedure itself was inappropriate to the circumstances because it did not provide for adequate and consistent performance of the piping system visual inspections and did not provide for sufficient objective quality evidence to demonstrate that the program requirements were met. The licensee entered this violation into its corrective action program to investigate the cause and to identify appropriate corrective actions.
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| The finding would become a more significant safety concern if left uncorrected and was therefore more than a minor concern. Specifically, the failure to adequately implement required leakage reduction and monitoring program controls to minimize leakage from reactor coolant sources outside of containment that could contain highly radioactive fluids during a serious transient or accident could reasonably result in higher doses to plant workers and higher potential offsite release levels. Because the leakage reduction and monitoring program is intended to contain highly radioactive fluids within piping systems outside containment, which supports the radiological barrier functions to protect plant workers and the public following serious transients or accidents, the inspectors concluded that this issue was associated with the Barrier Integrity Cornerstone. The finding was of very low safety significance because it involved only a degradation of the radiological barrier function provided for the Auxiliary Building. The inspectors concluded that this finding affected the cross-cutting area of human performance because the licensee did not provide adequate procedural guidance and training to enable operators to correctly perform and document piping system visual inspections to implement its leakage reduction and monitoring program. As a result, the licensee did not have appropriate objective quality evidence to demonstrate that the program requirements were met.
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| Inspection Report# : 2009004 (pdf)
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| Emergency Preparedness Significance: Jun 30, 2010 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE EMERGENCY PLAN AUGMENTATION CALL-IN DRILLS The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50.54(q) for the licensees failure to follow and maintain the Emergency Plan, which meets the standards in 10 CFR 50.47(b) and the requirements in Appendix E to 10 CFR 50. Specifically, the licensees Emergency Plan calls for the performance of periodic drills to evaluate the ability to augment its Emergency Response Organization (ERO).
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| However, the Emergency Plan implementing procedure used for the conduct of these augmentation drills exempts certain ERO members from participation in these drills, a situation which prevents the licensee from fully demonstrating its ability to augment all the ERO positions in a timely manner. The licensees approved Emergency Plan does not provide for such an exemption. The licensee entered the finding into the corrective action program.
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| The use of an implementing procedure that causes the conduct of an activity to be inconsistent with the associated requirements in the licensees Emergency Plan results in a failure to follow and maintain the Emergency Plan and is a performance deficiency. As a result of the limitations in the procedure, the licensee failed to conduct call-in drills to demonstrate timely augmentation of ERO positions filled by skilled/technical personnel. The deficiency did not impact the NRCs regulatory process or contribute to actual safety consequences; therefore, the performance deficiency was screened using the Emergency Preparedness Significance Determination Process as a failure to comply. The deficiency was determined to be more than minor because the deficiency adversely affected the Emergency Preparedness Cornerstone objective and had the attribute associated with ERO readiness and in the area of ERO augmentation testing. The inspector evaluated the finding using the Inspection Manual Chapter 0609, Appendix B, Sheet I, Failure to Comply Flowchart. The inspector evaluated the finding as a degraded planning standard function since the licensees conduct of the augmentation exercises did not include all ERO positions. The finding was determined to be of very low safety significance. The inspector determined the finding had a cross cutting aspect in the problem identification and resolution area with a component in self and independent assessments. The licensees
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| augmentation call-in drills were not comprehensive to include all ERO augmentation staffing positions. (IMC 0310 P.3(a))
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| Inspection Report# : 2010003 (pdf)
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| Significance: SL-IV Nov 20, 2009 Identified By: NRC Item Type: NCV NonCited Violation Implementation of a Change which Decreased the Effectiveness of the Emergency Plan The inspectors identified a NCV of 10 CFR 50.54(q) associated with 10 CFR 50.47(b)(2) because the licensee failed to obtain prior NRC approval for a change made to its emergency plan that decreased the effectiveness of the plan.
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| Specifically, the licensee removed staffing and capabilities from the minimum on-shift emergency response staffing requirements from the Clinton Power Station Emergency Plan Annex, Section 2, Table B-1. The licensee entered this issue into their corrective action program and replaced staffing back on-shift as required by the 1998 emergency plan annex.
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| This finding was more than minor and of very low safety-significance using IMC 0609, Appendix B, because the finding was associated with the Emergency Preparedness Cornerstone attribute of emergency response organization readiness for minimum on shift emergency response staffing. Because the finding affected the NRC's ability to perform its regulatory function, the inspectors evaluated the significance using the traditional enforcement process.
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| This finding was determined to be a Severity Level IV violation because the licensee failed to meet an emergency planning requirement not directly related to assessment and notification.
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| Inspection Report# : 2009502 (pdf)
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| Inspection Report# : 2010002 (pdf)
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| Occupational Radiation Safety Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.
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| Miscellaneous Last modified : September 02, 2010
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| Clinton 3Q/2010 Plant Inspection Findings Initiating Events Significance: Mar 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO CONTROL TRANSIENT COMBUSTIBLE MATERIALS IN ACCORDANCE WITH FIRE PROTECTION PROGRAM The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of the Clinton Power Station Unit 1 Operating License (NPF-62, Section 2.F). The licensee failed to implement the Fire Protection Program in accordance with program requirements by failing to follow approved Fire Protection Program procedures for the control of transient combustible materials. The licensee promptly removed the transient combustible materials found by the inspectors.
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| The inspectors concluded that this finding could be reasonably viewed as a precursor to a significant event (i.e., a fire affecting more than one train of safe shutdown equipment). Specifically, the presence of transient combustible materials in a combustible free zone could reasonably result in degradation of the fire protection defense-in-depth elements in place to prevent fires from starting and mitigate the consequences of fires. In addition, based on review of Example 4k in IMC 0612, "Power Reactor Inspection Reports," Appendix E, "Examples of Minor Issues," the issue would not be considered to be of minor significance because the identified transient combustibles were found in a combustible free zone required for separation of redundant trains. The finding was of very low safety significance because the items found in the combustible free zone would not be considered transient combustibles of significance as defined in IMC 0609, Appendix F, "Fire Protection Significance Determination Process," Attachment 2, "Degradation Rating Guidance Specific to Various Fire Protection Program Elements," and therefore the issue was assigned a "low degradation" rating. The inspectors concluded that this finding affected the cross-cutting area of problem identification and resolution. Specifically, the licensee missed an opportunity to identify and remove the transient combustible materials while implementing corrective actions for previous inspector identified findings involving the control of transient combustible materials.
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| Inspection Report# : 2010002 (pdf)
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| Significance: Mar 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO CONTROL COMBUSTIBLE GAS CYLINDERS IN ACCORDANCE WITH FIRE PROTECTION PROGRAM The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of the Clinton Power Station Unit 1 Operating License (NPF-62, Section 2.F). The licensee failed to implement the Fire Protection Program in accordance with program requirements by failing to follow approved Fire Protection Program procedures for the control of combustible gas cylinders in the plant. The licensee promptly removed the combustible gas cylinders found by the inspectors.
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| The inspectors concluded that this finding was associated with the Protection Against External Factors attribute of the Initiating Events Cornerstone and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the fire hazard for the affected area was increased by the uncontrolled presence of the compressed gas cylinders. In addition, based on review of Example 4k in Inspection Manual Chapter 0612, "Power Reactor Inspection Reports,"
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| Appendix E, "Examples of Minor Issues," the issue would not be considered to be of minor significance because a credible fire scenario involving the identified transient combustibles could affect equipment important to safety. The finding was detemined to be of very low safety significance during a Phase 3 Significance Determination Process review since the delta core damage frequency was determined to be negligible. Because a postulated fire in the area
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| where the combustible gas cylinders were found could affect only one train of safe shutdown equipment, the safe shutdown path was not affected by the finding. The inspectors concluded that this finding affected the cross-cutting area of human performance. Specifically, the licensee did not adquately ensure that supervisory and management oversight of work activities involving contractors supported nuclear safety.
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| Inspection Report# : 2010002 (pdf)
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| Significance: Mar 31, 2010 Identified By: Self-Revealing Item Type: FIN Finding FAILURE TO CORRECT INADEQUATE FWLCS RESPONSE RESULTED IN HIGH REACTOR VESSEL WATER LEVEL (LEVEL 8 ) SCRAM A finding of very low safety significance was self-revealed from an event that resulted in a Unit 1 reactor scram. The licensee failed to correct a non-conforming condition with inadequate response from the feedwater level control system (FWLCS) that caused an automatic reactor scram on February 10, 2008, following an unexpected loss of a reactor recirculation pump. This resulted in a second reactor scram for the same cause on October 15, 2009, following the unexpected loss of a reactor recirculation pump. Because the FWLCS is not safety-related, no violation of regulatory requirements was identified. The FWLCS response was corrected in January 2010 and proper system response was verified by the licensee upon start up from the January-February 2010 refueling outage.
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| The finding was of more than minor significance because this issue was associated with the Equipment Performance attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during power operations.
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| Specifically, inadequate FWLCS response resulted in a reactor scram following the unexpected loss of a reactor recirculation pump. The finding was of very low safety significance because the issue: (1) did not contribute to the likelihood of a primary or secondary system loss-of-coolant-accident initiator, (2) did not contribute to both the likelihood of a reactor trip AND the likelihood that mitigation equipment or functions would not be available, and (3) did not increase the likelihood of a fire or internal/external flooding event. The inspectors did not identify a cross cutting aspect related to this finding.
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| Inspection Report# : 2010002 (pdf)
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| Significance: Dec 31, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation FAILURE TO CORRECTLY TORQUE VALVE PACKING GLAND NUTS RESULTED IN VALVE PACKING FAILURE AND UNPLANNED PLANT SHUTDOWN A finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criteria V, "Instructions, Procedures, and Drawings," was self-revealed on September 29, 2009, when a steam leak developed from the reactor core isolation cooling (RCIC) system inboard steam isolation valve (1E51F0063) stem packing. This resulted in a plant shutdown due to a greater than 2 gallon-per-minute increase in unidentified reactor coolant system (RCS) leakage within the previous 24 hours. The licensee failed to correctly tighten the valve packing gland nuts to the as-left torque valve from original packing installation when performing scheduled maintenance to verify the as-found torque value. The licensee replaced the 1E51F0063 valve stem packing during the subsequent forced outage and tightened the gland nuts to the correct torque value.
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| The finding was of more than minor significance because it was associated with the Equipment Performance attribute of the Initiating Events Cornerstone and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the failure to correctly tighten the valve stem packing gland nuts resulted in stem packing failure and a subsequent plant shutdown due to exceeding the Technical Specification (TS) limit for an increase in unidentified RCS leakage.
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| Although the finding resulted in exceeding the TS limit for RCS leakage, it was determined to be of very low safety significance during a Phase 2 Significance Determination Process review because there was no loss of mitigation capability for any safety system and therefore no resultant change in core damage frequency. Because the performance issue was associated with maintenance performed in February 2006, it did not necessarily reflect current licensee performance and no cross-cutting aspect was identified.
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| Inspection Report# : 2009005 (pdf)
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| Mitigating Systems Significance: SL-IV Jun 30, 2010 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO SATISFY 10 CFR 50.72 AND 50.73 REPORTING REQUIREMENTS THUS AFFECTING THE REGULATORY PROCESS..
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| The inspectors identified a Severity Level IV Non- Cited Violation of the NRCs reporting requirements in 10 CFR 50.72(a)(1), Immediate Notification Requirements for Operating Nuclear Power Reactors, and 10 CFR 50.73(a)(1),
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| Licensee Event Report System. The licensee failed make a required 8-hour non-emergency notification call to the NRC Operations Center and failed to submit a required Licensee Event Report within 60 days after discovery of a condition that resulted in the plant being in an unanalyzed condition that significantly degraded plant safety and could have prevented fulfillment of the safety function of the emergency core cooling system. No immediate corrective actions were taken to address this finding; however, the licensee entered this issue into its corrective action program for evaluation.
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| This violation was of more than minor significance because the NRC relies on licensees to identify and report conditions or events meeting the criteria specified in the Technical Specifications and the regulations in order to perform its regulatory function. Because this issue affected the NRC's ability to perform its regulatory function, the inspectors evaluated it using the traditional enforcement process. The underlying technical issue (i.e., interconnecting floor drains between the Residual Heat Removal A Pump Room and the Radwaste Pipe Tunnel) was determined to be a finding of very low safety significance during a Phase 3 SDP evaluation. Consistent with the guidance in Supplement I, Paragraph D.4, of the NRC Enforcement Policy, the violation associated with this finding was determined to be a Severity Level IV Violation.
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| The related performance deficiency is tracked as item 2010-003-06.
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| Inspection Report# : 2010003 (pdf)
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| Significance: Jun 30, 2010 Identified By: NRC Item Type: FIN Finding OPERABILITY ASSESSMENT OF INSERVICE TESTING SURVEILLANCE DISCREPANCIES FOR EXCESS FLOW CHECK VALVES The inspectors identified a finding of very low safety significance associated with the licensee's failure to evaluate the functionality of multiple excess flow check valves that had not been tested in accordance with the American Society of Mechanical Engineers / American National Standards Institute (ASME/ANSI) Code Inservice Testing requirements to establish whether the nonconforming condition warranted starting the Technical Specification (TS) action time for the suppression pool makeup (SPMU) system. In response the the inspectors' questions, the licensee subsequently performed an operability evaluation. No violation of regulatory requirements was identified because subsequent testing by the licnesee determined that the valves were functional.
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| The finding would become a more significant safety concern if left uncorrected and was therefore more than a minor concern. Specifically, the failure to correctly evaluate a degraded/nonconforming condition potentially affecting the operability of structures, systems, or components (SSCs) required to be operable by TS could reasonably result in an unrecognized condition of an SSC failing to fulfill a safety-related function. Because the SPMU system was primarily assoicated with long term decay heat removal following certain design basis accidents, the inspectors concluded that this issue was associated with the Mitigating Systems Cornerstone. The finding was of very low safety significance because the issue: (1) was not a design or qualification deficiency; (2) did not represent an actual loss of safety function of a system; (3) did not represent an actual loss of safety function of a single train for greater than its TS allowed outage time; (4) did not represent an actual loss of safety function of one or more non-TS trains of equipment designated as risk significant; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors concluded that this finding affected the cross-cutting area of human
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| pefformance because the licenesee did not have a formal process in place with adequate guidance and training to enable licensed senior reactor operators to properly and promptly evaluate operability in this instance. As a result, senior reactor operators took it for granted that utilizing the relief allowed by TS Surveillance Requirement 3.0.3 and performing a risk evaluation obviated the need to address the operability of the instrumentation supported by the excess flow check valves for the ASME/ANSI Code noncompliance.
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| Inspection Report# : 2010003 (pdf)
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| Significance: Jun 30, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Satisfy 10 CFR 50.72 and 50.73 Reporting Requirements - performance deficiiency portion.
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| The inspectors identified a finding of very low safety significance of the NRCs reporting requirements in 10 CFR 50.72(a)(1), Immediate Notification Requirements for Operating Nuclear Power Reactors, and 10 CFR 50.73(a)(1),
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| Licensee Event Report System. The licensee failed to make a required 8 hour non emergency notification call to the NRC Operations Center and failed to submit a required Licensee Event Report within 60 days after discovery on October 7, 2009, of a condition that resulted in the plant being in an unanalyzed condition that significantly degraded plant safety and could have prevented fulfillment of the safety function of the emergency core cooling system. No immediate corrective actions were taken to address this finding; however, the licensee entered this issue into its corrective action program for evaluation.
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| This finding was of more than minor significance because the NRC relies on licensees to identify and report conditions or events meeting the criteria specified in the Technical Specifications and the regulations in order to perform its regulatory function. The inspectors assessed the significance of the underlying performance deficiency using the SDP. The underlying technical issue (i.e., interconnecting floor drains between the Residual Heat Removal A Pump Room and the Radwaste Pipe Tunnel) was determined to be a finding of very low safety significance (green) during a Phase 3 Significance Determination Process evaluation. This finding affected the cross cutting area of human performance because the licensee did not use conservative assumptions in decision making while evaluating the reportability of the unanalyzed condition with respect to the reporting requirements in 10 CFR 50.72(a)(1)(ii) and 50.73(a)(1). (IMC 0310 H.1(b)) (Section 1R06.1.b.(1))
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| The related traditional enforcment portion is tracked as item 2010-003-01.
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| Inspection Report# : 2010003 (pdf)
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| Significance: Jun 25, 2010 Identified By: NRC Item Type: NCV NonCited Violation Non Conservative Acceptance Criteria for RHR Pump Performance Testing The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, having very low safety-significance for the licensees failure to ensure adequate acceptance limits were incorporated into test procedures. Specifically, the licensee failed to properly consider instrument loop uncertainties and allowable emergency diesel generator frequency variance when determining the alert and required action values used in the inservice test procedure for testing of the residual heat removal pumps. Consequently, the acceptance criteria for the lower limits on degradation of pump head were non-conservative. This finding was entered into the licensees corrective action program and a preliminary calculation performed by the licensee concluded that the pumps were operable.
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| The finding was more than minor because it was associated with the Mitigating Systems cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the capability of the system to respond to initiating events to prevent undesirable consequences. This finding was of very low safety-significance (Green) because the licensee was able to demonstrate pump operability and therefore there was no loss of safety function. This finding had a cross-cutting aspect in the area of problem identification and resolution because the licensee did not thoroughly evaluate operating experience that included similar issues relating to the failure to appropriately account for instrument uncertainties in design analysis.
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| Inspection Report# : 2010006 (pdf)
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| Significance: Jun 25, 2010 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Test Control of RHR Heat Exchangers The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, having very low safety-significance for the licensees failure to establish test conditions to assure that the 1B residual heat removal heat exchanger would perform satisfactorily in service under accident conditions. Specifically, the inspectors determined that the heat exchanger thermal performance test procedure did not assure adequate temperature differences to provide reliable test results. In addition, the most recent test was performed with lower temperature differences than those identified in plant calculations. This finding was entered into the licensees corrective action program and a preliminary analysis performed by the licensee concluded the test results were acceptable.
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| The finding was more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the residual heat removal heat exchanger performance test procedure did not establish appropriate test conditions to ensure that the component would perform its required function during an accident. Also, the inspectors determined that the finding was similar to Examples 3.j and 3.k of IMC 612, Appendix E, in that there was a reasonable doubt of the operability of the component based on the most recent test conditions. The inspectors determined the finding was of very low safety significance (Green) because it was not a design or qualification deficiency, did not represent a loss of system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding did not have a cross-cutting aspect because it did not represent current performance.
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| Inspection Report# : 2010006 (pdf)
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| Significance: Mar 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation INTERCONNECTING FLOOR DRAINS BETWEEN THE RESIDUAL HEAT REMOVAL 'A' PUMP ROOM AND RADWASTE PIPE TUNNEL The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criteria III, "Design Control," regarding the licensee's failure to correctly translate the design basis into the design of the Auxiliary Building floor drain system with appropriate margin. The inspectors identified that floor drains in the Residual Heat Removal (RHR) 'A' Pump Room and the Radwaste Pipe Tunnel were interconnected, which resulted in the plant being in an unanalyzed condition that degraded plant safety and could have prevented fulfillment of the safety function of the containment suppression pool. To address the immediate operability concern, the licensee plugged the two floor drains in the Radwaste Pipe Tunnel line to prevent communication with the floor drain system in the RHR 'A' Pump Room. An exposed vertical section of the drain line was then cut and a solid steel plate welded into the pipe per an engineering design change to permanently isolate the floor drains between the two rooms.
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| The finding was of more than minor significance because it was associated with the Design Control attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage).
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| Specifically, the as-found configuration of the interconnecting floor drains resulted in the plant being in an unanalyzed condition that could have prevented fullfillment of the safety function of the containment suppression pool. Although the finding would represent a loss of safety function in the event of a postulated accident, it was determined to be of very low safety significance during a Phase 3 Significance Determination Process review because the delta core damage frequency was determined to be negligible since the initiating event frequency for flooding due to an RHR pump suction pipe failure was sufficiently low. Because this condition had existed since initial plant construction, the performance issue did not necessarily reflect current licensee performance and no cross-cutting aspect was identified.
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| Inspection Report# : 2010002 (pdf)
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| Significance: Dec 31, 2009 Identified By: Self-Revealing
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| Item Type: NCV NonCited Violation FAILURE TO CORRECTLY INSTALL RELAYS INSIDE OF THE DIVISION 3 DIESEL GENERATOR CONTROL PANEL A finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criteria V, "Instructions, Procedures, and Drawings," was self-revealed on September 23, 2009, when the Division 3 diesel generator (DG) was found to have had two components installed incorrectly. Electrical maintenance technicians had incorrectly replaced time delay relays K-8A and K-32 on September 24, 2007, essentially swapping the locations of the two relays. This rendered the Division 3 DG inoperable for about 2 years and resulted in a loss of safety function for the Division 3 DG and high pressure core spray system under a certain sequence of initiating events. The licensee restored the two time delay relays in the correct configuration and immediately verified that the remaining time delay relays inside the Division 3 DG Control Panel were in their proper locations.
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| The finding was of more than minor significance because, if left uncorrected,it would potentially lead to a more significant safety concern (i.e., the inoperability of risk-significant plant safety systems). In addition, based on review of Example 5c in IMC 0612, "Power Reactor Inspection Reports," Appendix E, "Examples of Minor Issues," the issue would not be considered to be of minor significance because the incorrect relays were installed in the control panel.
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| Although the finding resulted in a loss of safety function for the Division 3 DG and high pressure core spray system, it was determined to be of very low safety significance during a Phase 2 Significance Determination Process Review considering the very limited conditions (i.e., only 45 seconds following shutdown of the engine concurrent with a design basis accident) when the Division 3 DG was incapable of performing its safety function. The resultant exposure time was estimated to be about 27 minutes during the 2-year period. The inspectors concluded that this finding affected the cross-cutting area of human performance because the licensee did not effectively communicate expectations regarding procedural compliance and, as a result, maintenance technicians did not follow their procedures by installing nonconforming components and restoring the safety system to service.
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| Inspection Report# : 2009005 (pdf)
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| Barrier Integrity Significance: SL-IV Jun 30, 2010 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PERFORM AN ADEQUATE 10 CFR 50.59 EVALUATION FOR CPS PROCEDURE 3711.01 THUS AFFECTING THE NRC'S REGULATORY PROCESS.
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| The inspectors identified a Non-Cited Violation of 10 CFR 50.59, Changes, Tests and Experiments. The licensee failed to perform an adequate 10 CFR 50.59 evaluation and obtain a license amendment prior to implementing CPS 3711.01, CPS [Clinton Power Station] Operations with the Potential to Drain the Reactor Vessel [OPDRV],
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| Revision 0. The procedure established a definition of an OPDRV for use in determining the applicability of several Technical Specification (TS) requirements while in Modes 4 and 5. The licensee failed to recognize that implementing this new procedure, in effect, constituted a change to the TS incorporated into its licensing basis, which would therefore require a license amendment pursuant to 10 CFR 50.59(c)(1)(i) and 10 CFR 50.90. No immediate corrective actions were taken to address this finding; however, the licensee entered this issue into its corrective action program for evaluation.
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| The finding was of more than minor significance because there was a reasonable likelihood that the change requiring a 10 CFR 50.59 evaluation would require NRC review and approval prior to implementation. Because this issue affected the NRC's ability to perform its regulatory function, the inspectors evaluated it using the traditional enforcement process. Based on the results of a modified Phase 2 SDP evaluation, this finding was determined to be of very low safety significance. Consistent with the guidance in Supplement I, Paragraph D.5, of the NRC Enforcement Policy, the violation associated with this finding was determined to be a Severity Level IV Violation.
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| The related performance deficiency is tracked as item 2010-003-07.
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| Inspection Report# : 2010003 (pdf)
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| Significance: Jun 30, 2010
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| Identified By: Self-Revealing Item Type: NCV NonCited Violation FAILURE TO FOLLOW PROCEDURE RESULTING IN GATE SEAL LEAKAGE.
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| A finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criteria V, Instructions, Procedures, and Drawings, was self-revealed on January 29, 2010, when the dryer cavity gate seal depressurized during the performance of the containment and reactor vessel isolation functional surveillance procedure. When the seal lost pressure, approximately 46,500 gallons of water leaked from the dryer cavity pool into the reactor cavity. In response to the event, the licensee ensured all personnel were out of the reactor cavity, entered its radioactive spill off-normal procedure, and re-established air pressure to the dryer cavity gate seal. Subsequent investigation revealed that during the gate seal inflation procedure the proper valve operation sequence was not followed. As corrective action, the licensee revised many of its procedures and included a special brief to the refueling outage preparation for Reactor Services personnel.
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| The finding was of more than minor significance because the licensees failure to correctly install the upper containment dryer cavity gate could be reasonably viewed as a precursor to a significant event and, if left uncorrected would potentially lead to a more significant safety concern (i.e., increased dose or personnel contamination). In addition, the finding was similar to Example 4c in Inspection Manual Chapter 0612, Power Reactor Inspection Reports, Appendix E, Examples of Minor Issues, in that data recorded during installation of the dryer cavity gate seal was incorrect and resulted in backup air bottle supply pressure left outside the acceptable range. Because the dryer cavity gate seal is intended to contain highly radioactive fluids within containment, which supports the radiological barrier functions to protect plant workers and the public following serious transients or accidents, the inspectors concluded that this issue was associated with the Barrier Integrity Cornerstone. Although this event resulted in a loss of inventory from the dryer cavity pool and partial flooding of the lower reactor cavity and drywell, it was determined to be of very low safety significance because there was no loss inventory from the reactor vessel and it could not result in the loss of reactor coolant system level instrumentation. The inspectors concluded that this finding affected the cross-cutting area of human performance. The licensee did not effectively communicate expectations regarding procedural compliance in this instance and, as a result, the Reactor Services maintenance craftsman did not correctly follow the procedure by performing steps out of sequence and restoring a system to service that was incorrectly aligned. (IMC 0310 H.4(b))
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| Inspection Report# : 2010003 (pdf)
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| Significance: Jun 30, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an Adequate 10 CFR 50.59 Evaluation for CPS Procedure 3711.01 - performance deficiency portion The inspectors identified a finding of very low safety significance with an associated NCV of 10 CFR 50.59, Changes, Tests and Experiments. The licensee failed to perform an adequate 10 CFR 50.59 evaluation and obtain a license amendment prior to implementing CPS 3711.01, CPS [Clinton Power Station] Operations with the Potential to Drain the Reactor Vessel [OPDRV], Revision 0 on January 11, 2010. The procedure established a definition of an OPDRV for use in determining the applicability of several TS requirements while in Modes 4 and 5. The licensee failed to recognize that implementing this new procedure, in effect, constituted a change to the TS incorporated into its licensing basis, which would, therefore, require a license amendment pursuant to 10 CFR 50.59(c)(1)(i) and 10 CFR 50.90. No immediate corrective actions were taken to address this finding; however, the licensee entered this issue into its corrective action program for evaluation.
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| The finding was of more than minor significance because there was a reasonable likelihood that the change requiring a 10 CFR 50.59 evaluation would require NRC review and approval prior to implementation. The inspectors assessed the significance of the underlying issue using the SDP. Based on the results of a modified Phase 2 SDP evaluation, this finding was determined to be of very low safety significance. The inspectors concluded that this finding affected the cross cutting area of human performance. Specifically, the licensee did not use conservative decision making to demonstrate that the proposed action did not require prior NRC approval. The inspectors noted that the licensee was aware of potential concerns regarding the new procedure prior to completing the initial 10 CFR 50.59 evaluation and again prior to revising the evaluation in response to concerns raised by the inspectors; however, the incorrect conclusion was reached in both revisions of the evaluation that the new procedure was not a change to the TS and that
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| a license amendment was not necessary. (IMC 0310 H.1(b)) (Section 1R13.b.(1))
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| The associated traditional enforcment is tracked as item 2010-003-02.
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| Inspection Report# : 2010003 (pdf)
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| Significance: Mar 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO RECOGNIZE EXAMINATION LIMITATIONS FOR A CONTAINMENT PENETRATION WELD The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," for the licensee's failure to follow procedure instructions and record examination limitations for containment pipe-to-penetration weld 1-MS-B-11. The licensee subsequently documented the failure to record the 1-MS-B-11 limited weld examination in the corrective action program. The licensee planned to submit limited containment pipe-to-penetration weld examinations to the NRC for review and approval.
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| The finding was of more than minor significance because, if left uncorrected, the failure to document limited weld examinations could become a more significant safety concern. Absent NRC identification, the licensee would not have submitted limited weld examinations to the NRC for approval. Further, the inspector could not determine if the NRC would approve the limited weld surface examinations without a licensee evaluation for the extent of additional coverage possible with volumetric weld examinations. This finding was of very low safey-significance based on answering "no" to each of the Phase 1 screening questions identified in the Containment Barrier column of Table 4a in 609.04, "Phase 1 - Initial Screening and Characterization of Findings." Specifically, this finding did not represent an actual open pathway in the physical integrity of reactor containment. This finding has a cross-cutting aspect in the area of Human Performance, Resources because the licensee did not provide complete, accurate and up-to-date design documents (weld construction drawing) to the non-destructive examination staff. Specifically, the lack of a weld construction drawing which included the weld profile appeared to have contributed to the examination staff's failure to recognize that they had not completely examined the required weld surfaces.
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| Inspection Report# : 2010002 (pdf)
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| Significance: Mar 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE TEST CRITERIA IN STANDBY GAS TREATMENT SYSTEM FLOW/HEATER OPERABILITY SURVEILLANCE TEST The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criteria V, "Instructions, Procedures, and Drawings." The licensee failed to include appropriate quantitative or qualitative acceptance criteria in its surveillance test procedure for fulfilling the monthly surveillance requirement to demonstrate operability of the standby gas treatment (SGT) system as described in the Technical Specification Bases. As corrective action, the licensee revised the procedure to include acceptance criteria that system flow is normal and that no blockage, fan or motor failure, or excessive vibration is detected.
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| The finding was of more than minor signifcance because it is associated with the Procedure Quality cornerstone attribute for the Control Room and Auxiliary Building and adversely affected the Barrier Integrity cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, by not providing appropriate acceptance criteria by which the operability of the SGT system trains could be assessed, the ability of the SGT system to collect and treat the design leakage of radionuclides from the primary containment to the secondary containment during an accident could not be assured.
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| The inspectors did not identify a cross-cutting aspect related to this finding.
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| Inspection Report# : 2010002 (pdf)
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| Emergency Preparedness
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| Significance: Jun 30, 2010 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE EMERGENCY PLAN AUGMENTATION CALL-IN DRILLS The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50.54(q) for the licensees failure to follow and maintain the Emergency Plan, which meets the standards in 10 CFR 50.47(b) and the requirements in Appendix E to 10 CFR 50. Specifically, the licensees Emergency Plan calls for the performance of periodic drills to evaluate the ability to augment its Emergency Response Organization (ERO).
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| However, the Emergency Plan implementing procedure used for the conduct of these augmentation drills exempts certain ERO members from participation in these drills, a situation which prevents the licensee from fully demonstrating its ability to augment all the ERO positions in a timely manner. The licensees approved Emergency Plan does not provide for such an exemption. The licensee entered the finding into the corrective action program.
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| The use of an implementing procedure that causes the conduct of an activity to be inconsistent with the associated requirements in the licensees Emergency Plan results in a failure to follow and maintain the Emergency Plan and is a performance deficiency. As a result of the limitations in the procedure, the licensee failed to conduct call-in drills to demonstrate timely augmentation of ERO positions filled by skilled/technical personnel. The deficiency did not impact the NRCs regulatory process or contribute to actual safety consequences; therefore, the performance deficiency was screened using the Emergency Preparedness Significance Determination Process as a failure to comply. The deficiency was determined to be more than minor because the deficiency adversely affected the Emergency Preparedness Cornerstone objective and had the attribute associated with ERO readiness and in the area of ERO augmentation testing. The inspector evaluated the finding using the Inspection Manual Chapter 0609, Appendix B, Sheet I, Failure to Comply Flowchart. The inspector evaluated the finding as a degraded planning standard function since the licensees conduct of the augmentation exercises did not include all ERO positions. The finding was determined to be of very low safety significance. The inspector determined the finding had a cross cutting aspect in the problem identification and resolution area with a component in self and independent assessments. The licensees augmentation call-in drills were not comprehensive to include all ERO augmentation staffing positions. (IMC 0310 P.3(a))
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| Inspection Report# : 2010003 (pdf)
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| Significance: SL-IV Nov 20, 2009 Identified By: NRC Item Type: NCV NonCited Violation Implementation of a Change which Decreased the Effectiveness of the Emergency Plan The inspectors identified a NCV of 10 CFR 50.54(q) associated with 10 CFR 50.47(b)(2) because the licensee failed to obtain prior NRC approval for a change made to its emergency plan that decreased the effectiveness of the plan.
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| Specifically, the licensee removed staffing and capabilities from the minimum on-shift emergency response staffing requirements from the Clinton Power Station Emergency Plan Annex, Section 2, Table B-1. The licensee entered this issue into their corrective action program and replaced staffing back on-shift as required by the 1998 emergency plan annex.
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| This finding was more than minor and of very low safety-significance using IMC 0609, Appendix B, because the finding was associated with the Emergency Preparedness Cornerstone attribute of emergency response organization readiness for minimum on shift emergency response staffing. Because the finding affected the NRC's ability to perform its regulatory function, the inspectors evaluated the significance using the traditional enforcement process.
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| This finding was determined to be a Severity Level IV violation because the licensee failed to meet an emergency planning requirement not directly related to assessment and notification.
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| Inspection Report# : 2010002 (pdf)
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| Inspection Report# : 2009502 (pdf)
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| Occupational Radiation Safety
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| Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.
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| Miscellaneous Last modified : November 29, 2010
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| Clinton 4Q/2010 Plant Inspection Findings Initiating Events Significance: Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO CONTROL TRANSIENT COMBUSTIBLE MATERIALS IN ACCORDANCE WITH FIRE PROTECTION PROGRAM.
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| The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of the Clinton Power Station Unit 1 Operating License (NPF-62, Section 2.F). The licensee failed to implement the Fire Protection Program in accordance with program requirements by failing to follow approved Fire Protection Program procedures for the control of transient combustible materials. The licensee promptly removed the transient combustible materials found by the inspectors.
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| The inspectors concluded that this finding could be reasonably viewed as a precursor to a significant event (i.e., a fire affecting more than one train of safe shutdown equipment). Specifically, the presence of transient combustible materials in a combustible free zone could reasonably result in degradation of the fire protection defense-in-depth elements in place to prevent fires from starting and mitigate the consequences of fires. In addition, based on review of Example 4k in IMC 0612, "Power Reactor Inspection Reports," Appendix E, "Examples of Minor Issues," the issue would not be considered to be of minor significance because the identified transient combustibles were found in a combustible free zone required for separation of redundant trains. The finding was of very low safety significance because the items found in the combustible free zone would not be considered transient combustibles of significance as defined in IMC 0609, Appendix F, "Fire Protection Significance Determination Process," Attachment 2, "Degradation Rating Guidance Specific to Various Fire Protection Program Elements," and, therefore, the issue was assigned a "low degradation" rating. The inspectors condluded that this finding affected the cross-cutting area of human performance. Specifically, the licensee failed to recognize that moving a bullet-resistant container (BRC) was an infrequent activity and, as such, a pre-job briefing should have been performed and was not. In addition, a questioning attitude was not cultivated by the licensee once the correct location of the BRC wash challenged such that security staff proceeded in the face of uncertainty. Therefore, the inspectors concluded that the licensee's work practices that support human performance were less than effective. (IMC 0305 H.4(a))
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| Inspection Report# : 2010005 (pdf)
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| Significance: Mar 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO CONTROL TRANSIENT COMBUSTIBLE MATERIALS IN ACCORDANCE WITH FIRE PROTECTION PROGRAM The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of the Clinton Power Station Unit 1 Operating License (NPF-62, Section 2.F). The licensee failed to implement the Fire Protection Program in accordance with program requirements by failing to follow approved Fire Protection Program procedures for the control of transient combustible materials. The licensee promptly removed the transient combustible materials found by the inspectors.
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| The inspectors concluded that this finding could be reasonably viewed as a precursor to a significant event (i.e., a fire affecting more than one train of safe shutdown equipment). Specifically, the presence of transient combustible materials in a combustible free zone could reasonably result in degradation of the fire protection defense-in-depth elements in place to prevent fires from starting and mitigate the consequences of fires. In addition, based on review of Example 4k in IMC 0612, "Power Reactor Inspection Reports," Appendix E, "Examples of Minor Issues," the issue would not be considered to be of minor significance because the identified transient combustibles were found in a combustible free zone required for separation of redundant trains. The finding was of very low safety significance
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| because the items found in the combustible free zone would not be considered transient combustibles of significance as defined in IMC 0609, Appendix F, "Fire Protection Significance Determination Process," Attachment 2, "Degradation Rating Guidance Specific to Various Fire Protection Program Elements," and therefore the issue was assigned a "low degradation" rating. The inspectors concluded that this finding affected the cross-cutting area of problem identification and resolution. Specifically, the licensee missed an opportunity to identify and remove the transient combustible materials while implementing corrective actions for previous inspector identified findings involving the control of transient combustible materials.
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| Inspection Report# : 2010002 (pdf)
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| Significance: Mar 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO CONTROL COMBUSTIBLE GAS CYLINDERS IN ACCORDANCE WITH FIRE PROTECTION PROGRAM The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of the Clinton Power Station Unit 1 Operating License (NPF-62, Section 2.F). The licensee failed to implement the Fire Protection Program in accordance with program requirements by failing to follow approved Fire Protection Program procedures for the control of combustible gas cylinders in the plant. The licensee promptly removed the combustible gas cylinders found by the inspectors.
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| The inspectors concluded that this finding was associated with the Protection Against External Factors attribute of the Initiating Events Cornerstone and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the fire hazard for the affected area was increased by the uncontrolled presence of the compressed gas cylinders. In addition, based on review of Example 4k in Inspection Manual Chapter 0612, "Power Reactor Inspection Reports,"
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| Appendix E, "Examples of Minor Issues," the issue would not be considered to be of minor significance because a credible fire scenario involving the identified transient combustibles could affect equipment important to safety. The finding was detemined to be of very low safety significance during a Phase 3 Significance Determination Process review since the delta core damage frequency was determined to be negligible. Because a postulated fire in the area where the combustible gas cylinders were found could affect only one train of safe shutdown equipment, the safe shutdown path was not affected by the finding. The inspectors concluded that this finding affected the cross-cutting area of human performance. Specifically, the licensee did not adquately ensure that supervisory and management oversight of work activities involving contractors supported nuclear safety.
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| Inspection Report# : 2010002 (pdf)
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| Significance: Mar 31, 2010 Identified By: Self-Revealing Item Type: FIN Finding FAILURE TO CORRECT INADEQUATE FWLCS RESPONSE RESULTED IN HIGH REACTOR VESSEL WATER LEVEL (LEVEL 8 ) SCRAM A finding of very low safety significance was self-revealed from an event that resulted in a Unit 1 reactor scram. The licensee failed to correct a non-conforming condition with inadequate response from the feedwater level control system (FWLCS) that caused an automatic reactor scram on February 10, 2008, following an unexpected loss of a reactor recirculation pump. This resulted in a second reactor scram for the same cause on October 15, 2009, following the unexpected loss of a reactor recirculation pump. Because the FWLCS is not safety-related, no violation of regulatory requirements was identified. The FWLCS response was corrected in January 2010 and proper system response was verified by the licensee upon start up from the January-February 2010 refueling outage.
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| The finding was of more than minor significance because this issue was associated with the Equipment Performance attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during power operations.
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| Specifically, inadequate FWLCS response resulted in a reactor scram following the unexpected loss of a reactor recirculation pump. The finding was of very low safety significance because the issue: (1) did not contribute to the likelihood of a primary or secondary system loss-of-coolant-accident initiator, (2) did not contribute to both the likelihood of a reactor trip AND the likelihood that mitigation equipment or functions would not be available, and (3)
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| did not increase the likelihood of a fire or internal/external flooding event. The inspectors did not identify a cross cutting aspect related to this finding.
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| Inspection Report# : 2010002 (pdf)
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| Mitigating Systems Significance: SL-IV Jun 30, 2010 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO SATISFY 10 CFR 50.72 AND 50.73 REPORTING REQUIREMENTS.
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| The inspectors identified a Severity Level IV Non- Cited Violation of the NRCs reporting requirements in 10 CFR 50.72(a)(1), Immediate Notification Requirements for Operating Nuclear Power Reactors, and 10 CFR 50.73(a)(1),
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| Licensee Event Report System. The licensee failed make a required 8-hour non-emergency notification call to the NRC Operations Center and failed to submit a required Licensee Event Report within 60 days after discovery of a condition that resulted in the plant being in an unanalyzed condition that significantly degraded plant safety and could have prevented fulfillment of the safety function of the emergency core cooling system. No immediate corrective actions were taken to address this finding; however, the licensee entered this issue into its corrective action program for evaluation.
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| This violation was of more than minor significance because the NRC relies on licensees to identify and report conditions or events meeting the criteria specified in the Technical Specifications and the regulations in order to perform its regulatory function. Because this issue affected the NRC's ability to perform its regulatory function, the inspectors evaluated it using the traditional enforcement process. The underlying technical issue (i.e., interconnecting floor drains between the Residual Heat Removal A Pump Room and the Radwaste Pipe Tunnel) was determined to be a finding of very low safety significance during a Phase 3 SDP evaluation. Consistent with the guidance in Supplement I, Paragraph D.4, of the NRC Enforcement Policy, the violation associated with this finding was determined to be a Severity Level IV Violation.
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| The related performance deficiency is tracked as item 2010-003-06.
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| Inspection Report# : 2010003 (pdf)
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| Significance: Jun 30, 2010 Identified By: NRC Item Type: FIN Finding OPERABILITY ASSESSMENT OF INSERVICE TESTING SURVEILLANCE DISCREPANCIES FOR EXCESS FLOW CHECK VALVES The inspectors identified a finding of very low safety significance associated with the licensee's failure to evaluate the functionality of multiple excess flow check valves that had not been tested in accordance with the American Society of Mechanical Engineers / American National Standards Institute (ASME/ANSI) Code Inservice Testing requirements to establish whether the nonconforming condition warranted starting the Technical Specification (TS) action time for the suppression pool makeup (SPMU) system. In response the the inspectors' questions, the licensee subsequently performed an operability evaluation. No violation of regulatory requirements was identified because subsequent testing by the licnesee determined that the valves were functional.
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| The finding would become a more significant safety concern if left uncorrected and was therefore more than a minor concern. Specifically, the failure to correctly evaluate a degraded/nonconforming condition potentially affecting the operability of structures, systems, or components (SSCs) required to be operable by TS could reasonably result in an unrecognized condition of an SSC failing to fulfill a safety-related function. Because the SPMU system was primarily assoicated with long term decay heat removal following certain design basis accidents, the inspectors concluded that this issue was associated with the Mitigating Systems Cornerstone. The finding was of very low safety significance because the issue: (1) was not a design or qualification deficiency; (2) did not represent an actual loss of safety function of a system; (3) did not represent an actual loss of safety function of a single train for greater than its TS allowed outage time; (4) did not represent an actual loss of safety function of one or more non-TS trains of equipment
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| designated as risk significant; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors concluded that this finding affected the cross-cutting area of human pefformance because the licenesee did not have a formal process in place with adequate guidance and training to enable licensed senior reactor operators to properly and promptly evaluate operability in this instance. As a result, senior reactor operators took it for granted that utilizing the relief allowed by TS Surveillance Requirement 3.0.3 and performing a risk evaluation obviated the need to address the operability of the instrumentation supported by the excess flow check valves for the ASME/ANSI Code noncompliance.
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| Inspection Report# : 2010003 (pdf)
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| Significance: Jun 30, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Satisfy 10 CFR 50.72 and 50.73 Reporting Requirements - performance deficiiency portion.
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| The inspectors identified a finding of very low safety significance of the NRCs reporting requirements in 10 CFR 50.72(a)(1), Immediate Notification Requirements for Operating Nuclear Power Reactors, and 10 CFR 50.73(a)(1),
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| Licensee Event Report System. The licensee failed to make a required 8-hour non-emergency notification call to the NRC Operations Center and failed to submit a required Licensee Event Report within 60 days after discovery on October 7, 2009, of a condition that resulted in the plant being in an unanalyzed condition that significantly degraded plant safety and could have prevented fulfillment of the safety function of the emergency core cooling system. No immediate corrective actions were taken to address this finding; however, the licensee entered this issue into its corrective action program for evaluation.
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| This finding was of more than minor significance because the NRC relies on licensees to identify and report conditions or events meeting the criteria specified in the Technical Specifications and the regulations in order to perform its regulatory function. The inspectors assessed the significance of the underlying performance deficiency using the SDP. The underlying technical issue (i.e., interconnecting floor drains between the Residual Heat Removal A Pump Room and the Radwaste Pipe Tunnel) was determined to be a finding of very low safety significance (green) during a Phase 3 Significance Determination Process evaluation. This finding affected the cross cutting area of human performance because the licensee did not use conservative assumptions in decision making while evaluating the reportability of the unanalyzed condition with respect to the reporting requirements in 10 CFR 50.72(a)(1)(ii) and 50.73(a)(1).
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| The related traditional enforcment portion is tracked as item 2010-003-01.
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| Inspection Report# : 2010003 (pdf)
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| Significance: Jun 25, 2010 Identified By: NRC Item Type: NCV NonCited Violation NON CONSERVATIVE ACCEPTANCE CRITERIA FOR RHR PUMP PERFORMANCE TESTING The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, having very low safety-significance for the licensees failure to ensure adequate acceptance limits were incorporated into test procedures. Specifically, the licensee failed to properly consider instrument loop uncertainties and allowable emergency diesel generator frequency variance when determining the alert and required action values used in the inservice test procedure for testing of the residual heat removal pumps. Consequently, the acceptance criteria for the lower limits on degradation of pump head were non-conservative. This finding was entered into the licensees corrective action program and a preliminary calculation performed by the licensee concluded that the pumps were operable.
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| The finding was more than minor because it was associated with the Mitigating Systems cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the capability of the system to respond to initiating events to prevent undesirable consequences. This finding was of very low safety-significance (Green) because the licensee was able to demonstrate pump operability and therefore there was no loss of safety function. This finding had a cross-cutting aspect in the area of problem identification and resolution because the licensee did not thoroughly evaluate operating experience that included similar issues relating to the failure to appropriately account for instrument uncertainties in design analysis.
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| Inspection Report# : 2010006 (pdf)
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| Significance: Jun 25, 2010 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE TEST CONTROL OF RHR HEAT EXCHANGERS The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, having very low safety-significance for the licensees failure to establish test conditions to assure that the 1B residual heat removal heat exchanger would perform satisfactorily in service under accident conditions. Specifically, the inspectors determined that the heat exchanger thermal performance test procedure did not assure adequate temperature differences to provide reliable test results. In addition, the most recent test was performed with lower temperature differences than those identified in plant calculations. This finding was entered into the licensees corrective action program and a preliminary analysis performed by the licensee concluded the test results were acceptable.
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| The finding was more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the residual heat removal heat exchanger performance test procedure did not establish appropriate test conditions to ensure that the component would perform its required function during an accident. Also, the inspectors determined that the finding was similar to Examples 3.j and 3.k of IMC 612, Appendix E, in that there was a reasonable doubt of the operability of the component based on the most recent test conditions. The inspectors determined the finding was of very low safety significance (Green) because it was not a design or qualification deficiency, did not represent a loss of system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding did not have a cross-cutting aspect because it did not represent current performance.
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| Inspection Report# : 2010006 (pdf)
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| Significance: Mar 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation INTERCONNECTING FLOOR DRAINS BETWEEN THE RESIDUAL HEAT REMOVAL 'A' PUMP ROOM AND RADWASTE PIPE TUNNEL The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criteria III, "Design Control," regarding the licensee's failure to correctly translate the design basis into the design of the Auxiliary Building floor drain system with appropriate margin. The inspectors identified that floor drains in the Residual Heat Removal (RHR) 'A' Pump Room and the Radwaste Pipe Tunnel were interconnected, which resulted in the plant being in an unanalyzed condition that degraded plant safety and could have prevented fulfillment of the safety function of the containment suppression pool. To address the immediate operability concern, the licensee plugged the two floor drains in the Radwaste Pipe Tunnel line to prevent communication with the floor drain system in the RHR 'A' Pump Room. An exposed vertical section of the drain line was then cut and a solid steel plate welded into the pipe per an engineering design change to permanently isolate the floor drains between the two rooms.
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| The finding was of more than minor significance because it was associated with the Design Control attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage).
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| Specifically, the as-found configuration of the interconnecting floor drains resulted in the plant being in an unanalyzed condition that could have prevented fullfillment of the safety function of the containment suppression pool. Although the finding would represent a loss of safety function in the event of a postulated accident, it was determined to be of very low safety significance during a Phase 3 Significance Determination Process review because the delta core damage frequency was determined to be negligible since the initiating event frequency for flooding due to an RHR pump suction pipe failure was sufficiently low. Because this condition had existed since initial plant construction, the performance issue did not necessarily reflect current licensee performance and no cross-cutting aspect was identified.
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| Inspection Report# : 2010002 (pdf)
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| Barrier Integrity Significance: Dec 31, 2010 Identified By: Self-Revealing Item Type: FIN Finding FAILURE TO PERFORM PREVENTATIVE MAINTENANCE OF DIVISION 1 SELF TEST SYSTEM (STS) POWER SUPPLY RESULTS IN SPURIOUS REPOSITIONING OF SAFETY RELATED VALVES.
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| A finding of very low safety significance was self-revealed on August 24, 2010, when the Reactor Water Cleanup (RT) System return line outboard primary containment isolation valve went closed. Many other unintended valve repositioning events occurred from August 25 through August 26, 2010. The licensee failed to perform preventative maintenance on the Division 1 Self Test System (STS) safety-related 5 Volt (V) power supply. As a result, a degraded voltage condition existed in the test circuit, which was identified as the cause for the above valve repositioning events.
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| As a corrective action, the licensee has since installed a temporary plant modification of dual 5 V power supplies for all four divisions of the STS. No violation of regulatory requirements was identified.
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| The finding was of more than minor significance because the failure to perform preventative maintanance on critical components, if left uncorrected, would potentially lead to a more significant safety concern. This finding was of very low safety significance based on answering "no" to each of the Phase 1 screening questions identified in the Containment Barrier column of Table 4a in Attachment 0609.04, "Phase 1 - Initial Screening and Characterization of Findings." The inspectors concluded that this finding affected the cross-cutting area of human performance.
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| Specifically, in the area of resources the licensee did not adequately maintain long term plant safety by the maintenance of design margins, minimizing preventative maintenance deferrals, and ensuring maintenance and engineering backlogs which are low enough to support safety. (IMC 0310 H.2(a))
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| Inspection Report# : 2010005 (pdf)
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| Significance: SL-IV Jun 30, 2010 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PERFORM AN ADEQUATE 10 CFR 50.59 EVALUATION FOR CPS PROCEDURE 3711.01.
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| The inspectors identified a Non-Cited Violation of 10 CFR 50.59, Changes, Tests and Experiments. The licensee failed to perform an adequate 10 CFR 50.59 evaluation and obtain a license amendment prior to implementing CPS 3711.01, CPS [Clinton Power Station] Operations with the Potential to Drain the Reactor Vessel [OPDRV],
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| Revision 0. The procedure established a definition of an OPDRV for use in determining the applicability of several Technical Specification (TS) requirements while in Modes 4 and 5. The licensee failed to recognize that implementing this new procedure, in effect, constituted a change to the TS incorporated into its licensing basis, which would therefore require a license amendment pursuant to 10 CFR 50.59(c)(1)(i) and 10 CFR 50.90. No immediate corrective actions were taken to address this violation; however, the licensee entered this issue into its corrective action program for evaluation.
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| The violation was of more than minor significance because there was a reasonable likelihood that the change requiring a 10 CFR 50.59 evaluation would require NRC review and approval prior to implementation. Because this issue affected the NRC's ability to perform its regulatory function, the inspectors evaluated it using the traditional enforcement process. Based on the results of a modified Phase 2 SDP evaluation, the underlying technical issue was determined to be of very low safety significance. Consistent with the guidance in Supplement I, Paragraph D.5, of the NRC Enforcement Policy, the violation associated with this finding was determined to be a Severity Level IV Violation.
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| The related performance deficiency is tracked as item 2010-003-07.
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| Inspection Report# : 2010003 (pdf)
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| Significance: Jun 30, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation
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| FAILURE TO FOLLOW PROCEDURE RESULTING IN GATE SEAL LEAKAGE.
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| A finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criteria V, Instructions, Procedures, and Drawings, was self-revealed on January 29, 2010, when the dryer cavity gate seal depressurized during the performance of the containment and reactor vessel isolation functional surveillance procedure. When the seal lost pressure, approximately 46,500 gallons of water leaked from the dryer cavity pool into the reactor cavity. In response to the event, the licensee ensured all personnel were out of the reactor cavity, entered its radioactive spill off-normal procedure, and re-established air pressure to the dryer cavity gate seal. Subsequent investigation revealed that during the gate seal inflation procedure the proper valve operation sequence was not followed. As corrective action, the licensee revised many of its procedures and included a special brief to the refueling outage preparation for Reactor Services personnel.
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| The finding was of more than minor significance because the licensees failure to correctly install the upper containment dryer cavity gate could be reasonably viewed as a precursor to a significant event and, if left uncorrected would potentially lead to a more significant safety concern (i.e., increased dose or personnel contamination). In addition, the finding was similar to Example 4c in Inspection Manual Chapter 0612, Power Reactor Inspection Reports, Appendix E, Examples of Minor Issues, in that data recorded during installation of the dryer cavity gate seal was incorrect and resulted in backup air bottle supply pressure left outside the acceptable range. Because the dryer cavity gate seal is intended to contain highly radioactive fluids within containment, which supports the radiological barrier functions to protect plant workers and the public following serious transients or accidents, the inspectors concluded that this issue was associated with the Barrier Integrity Cornerstone. Although this event resulted in a loss of inventory from the dryer cavity pool and partial flooding of the lower reactor cavity and drywell, it was determined to be of very low safety significance because there was no loss inventory from the reactor vessel and it could not result in the loss of reactor coolant system level instrumentation. The inspectors concluded that this finding affected the cross-cutting area of human performance. The licensee did not effectively communicate expectations regarding procedural compliance in this instance and, as a result, the Reactor Services maintenance craftsman did not correctly follow the procedure by performing steps out of sequence and restoring a system to service that was incorrectly aligned.
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| Inspection Report# : 2010003 (pdf)
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| Significance: Jun 30, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an Adequate 10 CFR 50.59 Evaluation for CPS Procedure 3711.01 - performance deficiency portion The inspectors identified a finding of very low safety significance with an associated NCV of 10 CFR 50.59, Changes, Tests and Experiments. The licensee failed to perform an adequate 10 CFR 50.59 evaluation and obtain a license amendment prior to implementing CPS 3711.01, CPS [Clinton Power Station] Operations with the Potential to Drain the Reactor Vessel [OPDRV], Revision 0 on January 11, 2010. The procedure established a definition of an OPDRV for use in determining the applicability of several TS requirements while in Modes 4 and 5. The licensee failed to recognize that implementing this new procedure, in effect, constituted a change to the TS incorporated into its licensing basis, which would, therefore, require a license amendment pursuant to 10 CFR 50.59(c)(1)(i) and 10 CFR 50.90. No immediate corrective actions were taken to address this finding; however, the licensee entered this issue into its corrective action program for evaluation.
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| The finding was of more than minor significance because there was a reasonable likelihood that the change requiring a 10 CFR 50.59 evaluation would require NRC review and approval prior to implementation. The inspectors assessed the significance of the underlying issue using the SDP. Based on the results of a modified Phase 2 SDP evaluation, this finding was determined to be of very low safety significance. The inspectors concluded that this finding affected the cross cutting area of human performance. Specifically, the licensee did not use conservative decision making to demonstrate that the proposed action did not require prior NRC approval. The inspectors noted that the licensee was aware of potential concerns regarding the new procedure prior to completing the initial 10 CFR 50.59 evaluation and again prior to revising the evaluation in response to concerns raised by the inspectors; however, the incorrect conclusion was reached in both revisions of the evaluation that the new procedure was not a change to the TS and that a license amendment was not necessary.
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| The associated traditional enforcment is tracked as item 2010-003-02.
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| Inspection Report# : 2010003 (pdf)
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| Significance: Mar 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO RECOGNIZE EXAMINATION LIMITATIONS FOR A CONTAINMENT PENETRATION WELD The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," for the licensee's failure to follow procedure instructions and record examination limitations for containment pipe-to-penetration weld 1-MS-B-11. The licensee subsequently documented the failure to record the 1-MS-B-11 limited weld examination in the corrective action program. The licensee planned to submit limited containment pipe-to-penetration weld examinations to the NRC for review and approval.
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| The finding was of more than minor significance because, if left uncorrected, the failure to document limited weld examinations could become a more significant safety concern. Absent NRC identification, the licensee would not have submitted limited weld examinations to the NRC for approval. Further, the inspector could not determine if the NRC would approve the limited weld surface examinations without a licensee evaluation for the extent of additional coverage possible with volumetric weld examinations. This finding was of very low safey-significance based on answering "no" to each of the Phase 1 screening questions identified in the Containment Barrier column of Table 4a in 609.04, "Phase 1 - Initial Screening and Characterization of Findings." Specifically, this finding did not represent an actual open pathway in the physical integrity of reactor containment. This finding has a cross-cutting aspect in the area of Human Performance, Resources because the licensee did not provide complete, accurate and up-to-date design documents (weld construction drawing) to the non-destructive examination staff. Specifically, the lack of a weld construction drawing which included the weld profile appeared to have contributed to the examination staff's failure to recognize that they had not completely examined the required weld surfaces.
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| Inspection Report# : 2010002 (pdf)
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| Significance: Mar 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE TEST CRITERIA IN STANDBY GAS TREATMENT SYSTEM FLOW/HEATER OPERABILITY SURVEILLANCE TEST The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criteria V, "Instructions, Procedures, and Drawings." The licensee failed to include appropriate quantitative or qualitative acceptance criteria in its surveillance test procedure for fulfilling the monthly surveillance requirement to demonstrate operability of the standby gas treatment (SGT) system as described in the Technical Specification Bases. As corrective action, the licensee revised the procedure to include acceptance criteria that system flow is normal and that no blockage, fan or motor failure, or excessive vibration is detected.
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| The finding was of more than minor signifcance because it is associated with the Procedure Quality cornerstone attribute for the Control Room and Auxiliary Building and adversely affected the Barrier Integrity cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, by not providing appropriate acceptance criteria by which the operability of the SGT system trains could be assessed, the ability of the SGT system to collect and treat the design leakage of radionuclides from the primary containment to the secondary containment during an accident could not be assured.
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| The inspectors did not identify a cross-cutting aspect related to this finding.
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| Inspection Report# : 2010002 (pdf)
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| Emergency Preparedness
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| Significance: Jun 30, 2010 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE EMERGENCY PLAN AUGMENTATION CALL-IN DRILLS The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50.54(q) for the licensees failure to follow and maintain the Emergency Plan, which meets the standards in 10 CFR 50.47(b) and the requirements in Appendix E to 10 CFR 50. Specifically, the licensees Emergency Plan calls for the performance of periodic drills to evaluate the ability to augment its Emergency Response Organization (ERO).
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| However, the Emergency Plan implementing procedure used for the conduct of these augmentation drills exempts certain ERO members from participation in these drills, a situation which prevents the licensee from fully demonstrating its ability to augment all the ERO positions in a timely manner. The licensees approved Emergency Plan does not provide for such an exemption. The licensee entered the finding into the corrective action program.
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| The use of an implementing procedure that causes the conduct of an activity to be inconsistent with the associated requirements in the licensees Emergency Plan results in a failure to follow and maintain the Emergency Plan and is a performance deficiency. As a result of the limitations in the procedure, the licensee failed to conduct call-in drills to demonstrate timely augmentation of ERO positions filled by skilled/technical personnel. The deficiency did not impact the NRCs regulatory process or contribute to actual safety consequences; therefore, the performance deficiency was screened using the Emergency Preparedness Significance Determination Process as a failure to comply. The deficiency was determined to be more than minor because the deficiency adversely affected the Emergency Preparedness Cornerstone objective and had the attribute associated with ERO readiness and in the area of ERO augmentation testing. The inspector evaluated the finding using the Inspection Manual Chapter 0609, Appendix B, Sheet I, Failure to Comply Flowchart. The inspector evaluated the finding as a degraded planning standard function since the licensees conduct of the augmentation exercises did not include all ERO positions. The finding was determined to be of very low safety significance. Because the finding did not reflect current licensee performance, no cross-cutting aspect was identified.
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| Inspection Report# : 2010003 (pdf)
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| Occupational Radiation Safety Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.
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| Miscellaneous Last modified : March 03, 2011
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| Clinton 1Q/2011 Plant Inspection Findings Initiating Events Significance: Mar 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO CONTROL TRANSIENT COMBUSTIBLE MATERIALS IN ACCORDANCE WITH FIRE PROTECTION PROGRAM.
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| The inspectors identified a finding of very low safety significance with an associated non-cited violation of the Clinton Power Station Unit 1 Operating License (NPF 62, Section 2.F). The licensee failed to implement the Fire Protection Program in accordance with program requirements by failing to follow approved Fire Protection Program procedures for the control of transient combustible materials. The licensee promptly removed the transient combustible materials found by the inspectors and initiated compensatory measures.
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| The inspectors concluded that this finding could be reasonably viewed as a precursor to a significant event (i.e., a fire affecting more than one train of safe shutdown equipment). Specifically, the presence of transient combustible materials in a combustible free zone could reasonably result in degradation of the fire protection defense in depth elements in place to prevent fires from starting and mitigate the consequences of fires. In addition, based on review of Example 4k in IMC 0612, Power Reactor Inspection Reports, Appendix E, Examples of Minor Issues, the issue would not be considered to be of minor significance because the identified transient combustibles were found in a combustible free zone required for separation of redundant trains. The finding was of very low safety significance because the items found in the combustible free zone would not be considered transient combustibles of significance as defined in IMC 0609, Appendix F, Fire Protection Significance Determination Process, Attachment 2, Degradation Rating Guidance Specific to Various Fire Protection Program Elements, and, therefore, the issue was assigned a low degradation rating. The inspectors concluded that this finding affected the cross cutting area of human performance. Although a pre-job briefing was not required by the licensees procedure for the work activity, job site conditions and a discussion that the work was within a Transient Combustible Free Zone (TCFZ) was not included in the briefing. In addition, the workers 2 Minute Drill performed at the job site did not identify that work activities were within a TCFZ. Therefore, the inspectors concluded that the licensees work practices which support human performance were less than effective (H.4(a)).
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| Inspection Report# : 2011002 (pdf)
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| Significance: Mar 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO MEET SURVEILLANCE TESTING REQUIREMENT FOR HYDROGEN IGNITERS IN ACCESSIBLE AREAS OF THE PRIMARY CONTAINMENT AND DRYWELL.
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| The inspectors identified a finding of very low safety significance with an associated non-cited violation of Technical Specification Surveillance Requirement (TSSR) 3.6.3.2.4. The licensee failed to verify that each required hydrogen igniter in accessible areas of the Primary Containment and Drywell develops a surface temperature of = 1700 degrees Fahrenheit (°F) every 24 months. The licensee performed a risk assessment of the missed surveillance in accordance with TSSR 3.0.3, which determined that completion of the surveillance could be delayed up to the 24 month surveillance interval without a significant increase in plant risk. The licensee also completed an operability evaluation for the TS nonconformance and concluded that there was reasonable assurance that the affected hydrogen igniters were operable based on the results of surveillance testing to measure voltage/current draw.
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| The finding was of more than minor significance because it was associated with the Human Performance attribute for the Containment and adversely affected the Barrier Integrity Cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the licensee did not correctly evaluate a change to perform the surveillance test with the unit at power beginning in March 2002. It was not recognized that TSSR 3.6.3.2.4 would not be met for accessible hydrogen igniters in the
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| Drywell and 755 Elevation Steam Tunnel when performing the test with the unit at power and the licensee incorrectly believed that performance of the current/voltage surveillance test procedure for inaccessible igniters was an appropriate substitute, contrary to existing procedural guidance. The finding was a licensee performance deficiency of very low safety significance because it did not involve an actual reduction in the function of hydrogen igniters in the Primary Containment and Drywell. The inspectors concluded that because the scheduling change to perform the surveillance with the unit at power took place prior to surveillance testing beginning in March 2002, it did not necessarily reflect current licensee performance and no cross-cutting aspect was identified.
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| Inspection Report# : 2011002 (pdf)
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| Significance: Mar 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE TESTING CONTROLS TO PERFORM SURVEILLANCE TESTING OF HYDROGEN IGNITERS IN THE PRIMARY CONTAINMENT AND DRYWELL.
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| The inspectors identified a finding of very low safety significance with an associated non-cited violation of 10 CFR 50, Appendix B, Criterion XI, Test Control. The licensee failed to establish a test program adequate to assure testing of hydrogen igniters in accessible areas of the Primary Containment and Drywell pursuant to TSSR 3.6.3.2.4. The licensee entered this violation into its corrective action program to investigate the cause and to identify appropriate corrective actions.
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| The finding was of more than minor significance because it was associated with the Procedure Quality attribute for the Containment and adversely affected the Barrier Integrity Cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The finding was a licensee performance deficiency of very low safety significance because it did not involve an actual reduction in the function of hydrogen igniters in the Primary Containment and Drywell. The inspectors concluded that this finding affected the cross-cutting aspect of human performance. Specifically, adequate licensee resources involving personnel and procedures did not support successful human performance. CPS 9867.05 was not appropriate to the circumstances because it contained errors and did not provide adequate testing controls for the performance of the surveillance test (H.2(c)).
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| Inspection Report# : 2011002 (pdf)
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| Significance: Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO CONTROL TRANSIENT COMBUSTIBLE MATERIALS IN ACCORDANCE WITH FIRE PROTECTION PROGRAM.
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| The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of the Clinton Power Station Unit 1 Operating License (NPF-62, Section 2.F). The licensee failed to implement the Fire Protection Program in accordance with program requirements by failing to follow approved Fire Protection Program procedures for the control of transient combustible materials. The licensee promptly removed the transient combustible materials found by the inspectors.
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| The inspectors concluded that this finding could be reasonably viewed as a precursor to a significant event (i.e., a fire affecting more than one train of safe shutdown equipment). Specifically, the presence of transient combustible materials in a combustible free zone could reasonably result in degradation of the fire protection defense-in-depth elements in place to prevent fires from starting and mitigate the consequences of fires. In addition, based on review of Example 4k in IMC 0612, "Power Reactor Inspection Reports," Appendix E, "Examples of Minor Issues," the issue would not be considered to be of minor significance because the identified transient combustibles were found in a combustible free zone required for separation of redundant trains. The finding was of very low safety significance because the items found in the combustible free zone would not be considered transient combustibles of significance as defined in IMC 0609, Appendix F, "Fire Protection Significance Determination Process," Attachment 2, "Degradation Rating Guidance Specific to Various Fire Protection Program Elements," and, therefore, the issue was assigned a "low degradation" rating. The inspectors condluded that this finding affected the cross-cutting area of human performance. Specifically, the licensee failed to recognize that moving a bullet-resistant container (BRC) was
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| an infrequent activity and, as such, a pre-job briefing should have been performed and was not. In addition, a questioning attitude was not cultivated by the licensee once the correct location of the BRC wash challenged such that security staff proceeded in the face of uncertainty. Therefore, the inspectors concluded that the licensee's work practices that support human performance were less than effective. (IMC 0305 H.4(a))
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| Inspection Report# : 2010005 (pdf)
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| Mitigating Systems Significance: Mar 18, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure to Ensure Fire Door was Closed and Latched.
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| A finding of very low safety significance and associated NCV of Clinton Power Station Unit 1 Operating License NPF-62, Section 2.F was identified by the inspectors for the licensee's failure to ensure fire doors were closed and latched. Specifically, during a walkdown of fire area CB-1e 737 General Access Area, fire door 1DR1-432 located between fire area CB-1e and D-6 Emergency Diesel 2 Room, was found unlatched/not fully closed. The door was a 3-hour fire rated door credited for fire barrier between the two fire areas. Site personnel closed the door when it was found open and the door remained fully closed when challenged. The issue was entered into the licensee corrective action program as AR 01187906.
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| The inspectors determined that this finding was more than minor because the finding affected the Mitigating Systems cornerstone attributes of protection against external factors (Fire) and affected the cornerstone objective of ensuring the capability of the system to respond to events to prevent undesirable consequences. This finding was of very low safety significance (Green) based on answering Yes to Question 7 of Task 1.3.2. of Appendix F of IMC 0609. The inspectors did not identify a cross-cutting aspect associated with this finding because the underlining cause of unlatched door was indeterminate during the inspection.
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| Inspection Report# : 2011009 (pdf)
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| Significance: SL-IV Jun 30, 2010 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO SATISFY 10 CFR 50.72 AND 50.73 REPORTING REQUIREMENTS.
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| The inspectors identified a Severity Level IV Non- Cited Violation of the NRCs reporting requirements in 10 CFR 50.72(a)(1), Immediate Notification Requirements for Operating Nuclear Power Reactors, and 10 CFR 50.73(a)(1),
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| Licensee Event Report System. The licensee failed make a required 8-hour non-emergency notification call to the NRC Operations Center and failed to submit a required Licensee Event Report within 60 days after discovery of a condition that resulted in the plant being in an unanalyzed condition that significantly degraded plant safety and could have prevented fulfillment of the safety function of the emergency core cooling system. No immediate corrective actions were taken to address this finding; however, the licensee entered this issue into its corrective action program for evaluation.
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| This violation was of more than minor significance because the NRC relies on licensees to identify and report conditions or events meeting the criteria specified in the Technical Specifications and the regulations in order to perform its regulatory function. Because this issue affected the NRC's ability to perform its regulatory function, the inspectors evaluated it using the traditional enforcement process. The underlying technical issue (i.e., interconnecting floor drains between the Residual Heat Removal A Pump Room and the Radwaste Pipe Tunnel) was determined to be a finding of very low safety significance during a Phase 3 SDP evaluation. Consistent with the guidance in Supplement I, Paragraph D.4, of the NRC Enforcement Policy, the violation associated with this finding was determined to be a Severity Level IV Violation.
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| The related performance deficiency is tracked as item 2010-003-06.
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| Inspection Report# : 2011010 (pdf)
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| Inspection Report# : 2010003 (pdf)
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| Significance: Jun 30, 2010 Identified By: NRC Item Type: FIN Finding OPERABILITY ASSESSMENT OF INSERVICE TESTING SURVEILLANCE DISCREPANCIES FOR EXCESS FLOW CHECK VALVES The inspectors identified a finding of very low safety significance associated with the licensee's failure to evaluate the functionality of multiple excess flow check valves that had not been tested in accordance with the American Society of Mechanical Engineers / American National Standards Institute (ASME/ANSI) Code Inservice Testing requirements to establish whether the nonconforming condition warranted starting the Technical Specification (TS) action time for the suppression pool makeup (SPMU) system. In response the the inspectors' questions, the licensee subsequently performed an operability evaluation. No violation of regulatory requirements was identified because subsequent testing by the licnesee determined that the valves were functional.
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| The finding would become a more significant safety concern if left uncorrected and was therefore more than a minor concern. Specifically, the failure to correctly evaluate a degraded/nonconforming condition potentially affecting the operability of structures, systems, or components (SSCs) required to be operable by TS could reasonably result in an unrecognized condition of an SSC failing to fulfill a safety-related function. Because the SPMU system was primarily assoicated with long term decay heat removal following certain design basis accidents, the inspectors concluded that this issue was associated with the Mitigating Systems Cornerstone. The finding was of very low safety significance because the issue: (1) was not a design or qualification deficiency; (2) did not represent an actual loss of safety function of a system; (3) did not represent an actual loss of safety function of a single train for greater than its TS allowed outage time; (4) did not represent an actual loss of safety function of one or more non-TS trains of equipment designated as risk significant; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors concluded that this finding affected the cross-cutting area of human pefformance because the licenesee did not have a formal process in place with adequate guidance and training to enable licensed senior reactor operators to properly and promptly evaluate operability in this instance. As a result, senior reactor operators took it for granted that utilizing the relief allowed by TS Surveillance Requirement 3.0.3 and performing a risk evaluation obviated the need to address the operability of the instrumentation supported by the excess flow check valves for the ASME/ANSI Code noncompliance.
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| Inspection Report# : 2010003 (pdf)
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| Significance: Jun 30, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Satisfy 10 CFR 50.72 and 50.73 Reporting Requirements - performance deficiiency portion.
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| The inspectors identified a finding of very low safety significance of the NRCs reporting requirements in 10 CFR 50.72(a)(1), Immediate Notification Requirements for Operating Nuclear Power Reactors, and 10 CFR 50.73(a)(1),
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| Licensee Event Report System. The licensee failed to make a required 8-hour non-emergency notification call to the NRC Operations Center and failed to submit a required Licensee Event Report within 60 days after discovery on October 7, 2009, of a condition that resulted in the plant being in an unanalyzed condition that significantly degraded plant safety and could have prevented fulfillment of the safety function of the emergency core cooling system. No immediate corrective actions were taken to address this finding; however, the licensee entered this issue into its corrective action program for evaluation.
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| This finding was of more than minor significance because the NRC relies on licensees to identify and report conditions or events meeting the criteria specified in the Technical Specifications and the regulations in order to perform its regulatory function. The inspectors assessed the significance of the underlying performance deficiency using the SDP. The underlying technical issue (i.e., interconnecting floor drains between the Residual Heat Removal A Pump Room and the Radwaste Pipe Tunnel) was determined to be a finding of very low safety significance (green) during a Phase 3 Significance Determination Process evaluation. This finding affected the cross cutting area of human performance because the licensee did not use conservative assumptions in decision making while evaluating the reportability of the unanalyzed condition with respect to the reporting requirements in 10 CFR 50.72(a)(1)(ii) and 50.73(a)(1).
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| The related traditional enforcment portion is tracked as item 2010-003-01.
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| Inspection Report# : 2010003 (pdf)
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| Significance: Jun 25, 2010 Identified By: NRC Item Type: NCV NonCited Violation NON CONSERVATIVE ACCEPTANCE CRITERIA FOR RHR PUMP PERFORMANCE TESTING The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, having very low safety-significance for the licensees failure to ensure adequate acceptance limits were incorporated into test procedures. Specifically, the licensee failed to properly consider instrument loop uncertainties and allowable emergency diesel generator frequency variance when determining the alert and required action values used in the inservice test procedure for testing of the residual heat removal pumps. Consequently, the acceptance criteria for the lower limits on degradation of pump head were non-conservative. This finding was entered into the licensees corrective action program and a preliminary calculation performed by the licensee concluded that the pumps were operable.
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| The finding was more than minor because it was associated with the Mitigating Systems cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the capability of the system to respond to initiating events to prevent undesirable consequences. This finding was of very low safety-significance (Green) because the licensee was able to demonstrate pump operability and therefore there was no loss of safety function. This finding had a cross-cutting aspect in the area of problem identification and resolution because the licensee did not thoroughly evaluate operating experience that included similar issues relating to the failure to appropriately account for instrument uncertainties in design analysis.
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| Inspection Report# : 2010006 (pdf)
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| Significance: Jun 25, 2010 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE TEST CONTROL OF RHR HEAT EXCHANGERS The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, having very low safety-significance for the licensees failure to establish test conditions to assure that the 1B residual heat removal heat exchanger would perform satisfactorily in service under accident conditions. Specifically, the inspectors determined that the heat exchanger thermal performance test procedure did not assure adequate temperature differences to provide reliable test results. In addition, the most recent test was performed with lower temperature differences than those identified in plant calculations. This finding was entered into the licensees corrective action program and a preliminary analysis performed by the licensee concluded the test results were acceptable.
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| The finding was more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the residual heat removal heat exchanger performance test procedure did not establish appropriate test conditions to ensure that the component would perform its required function during an accident. Also, the inspectors determined that the finding was similar to Examples 3.j and 3.k of IMC 612, Appendix E, in that there was a reasonable doubt of the operability of the component based on the most recent test conditions. The inspectors determined the finding was of very low safety significance (Green) because it was not a design or qualification deficiency, did not represent a loss of system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding did not have a cross-cutting aspect because it did not represent current performance.
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| Inspection Report# : 2010006 (pdf)
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| Barrier Integrity
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| Significance: Dec 31, 2010 Identified By: Self-Revealing Item Type: FIN Finding FAILURE TO PERFORM PREVENTATIVE MAINTENANCE OF DIVISION 1 SELF TEST SYSTEM (STS) POWER SUPPLY RESULTS IN SPURIOUS REPOSITIONING OF SAFETY RELATED VALVES.
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| A finding of very low safety significance was self-revealed on August 24, 2010, when the Reactor Water Cleanup (RT) System return line outboard primary containment isolation valve went closed. Many other unintended valve repositioning events occurred from August 25 through August 26, 2010. The licensee failed to perform preventative maintenance on the Division 1 Self Test System (STS) safety-related 5 Volt (V) power supply. As a result, a degraded voltage condition existed in the test circuit, which was identified as the cause for the above valve repositioning events.
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| As a corrective action, the licensee has since installed a temporary plant modification of dual 5 V power supplies for all four divisions of the STS. No violation of regulatory requirements was identified.
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| The finding was of more than minor significance because the failure to perform preventative maintanance on critical components, if left uncorrected, would potentially lead to a more significant safety concern. This finding was of very low safety significance based on answering "no" to each of the Phase 1 screening questions identified in the Containment Barrier column of Table 4a in Attachment 0609.04, "Phase 1 - Initial Screening and Characterization of Findings." The inspectors concluded that this finding affected the cross-cutting area of human performance.
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| Specifically, in the area of resources the licensee did not adequately maintain long term plant safety by the maintenance of design margins, minimizing preventative maintenance deferrals, and ensuring maintenance and engineering backlogs which are low enough to support safety. (IMC 0310 H.2(a))
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| Inspection Report# : 2010005 (pdf)
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| Significance: SL-IV Jun 30, 2010 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PERFORM AN ADEQUATE 10 CFR 50.59 EVALUATION FOR CPS PROCEDURE 3711.01.
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| The inspectors identified a Non-Cited Violation of 10 CFR 50.59, Changes, Tests and Experiments. The licensee failed to perform an adequate 10 CFR 50.59 evaluation and obtain a license amendment prior to implementing CPS 3711.01, CPS [Clinton Power Station] Operations with the Potential to Drain the Reactor Vessel [OPDRV],
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| Revision 0. The procedure established a definition of an OPDRV for use in determining the applicability of several Technical Specification (TS) requirements while in Modes 4 and 5. The licensee failed to recognize that implementing this new procedure, in effect, constituted a change to the TS incorporated into its licensing basis, which would therefore require a license amendment pursuant to 10 CFR 50.59(c)(1)(i) and 10 CFR 50.90. No immediate corrective actions were taken to address this violation; however, the licensee entered this issue into its corrective action program for evaluation.
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| The violation was of more than minor significance because there was a reasonable likelihood that the change requiring a 10 CFR 50.59 evaluation would require NRC review and approval prior to implementation. Because this issue affected the NRC's ability to perform its regulatory function, the inspectors evaluated it using the traditional enforcement process. Based on the results of a modified Phase 2 SDP evaluation, the underlying technical issue was determined to be of very low safety significance. Consistent with the guidance in Supplement I, Paragraph D.5, of the NRC Enforcement Policy, the violation associated with this finding was determined to be a Severity Level IV Violation.
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| The related performance deficiency is tracked as item 2010-003-07.
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| Inspection Report# : 2011010 (pdf)
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| Inspection Report# : 2010003 (pdf)
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| Significance: Jun 30, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation FAILURE TO FOLLOW PROCEDURE RESULTING IN GATE SEAL LEAKAGE.
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| A finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criteria V, Instructions, Procedures, and Drawings, was self-revealed on January 29, 2010, when the dryer cavity gate seal
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| depressurized during the performance of the containment and reactor vessel isolation functional surveillance procedure. When the seal lost pressure, approximately 46,500 gallons of water leaked from the dryer cavity pool into the reactor cavity. In response to the event, the licensee ensured all personnel were out of the reactor cavity, entered its radioactive spill off-normal procedure, and re-established air pressure to the dryer cavity gate seal. Subsequent investigation revealed that during the gate seal inflation procedure the proper valve operation sequence was not followed. As corrective action, the licensee revised many of its procedures and included a special brief to the refueling outage preparation for Reactor Services personnel.
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| The finding was of more than minor significance because the licensees failure to correctly install the upper containment dryer cavity gate could be reasonably viewed as a precursor to a significant event and, if left uncorrected would potentially lead to a more significant safety concern (i.e., increased dose or personnel contamination). In addition, the finding was similar to Example 4c in Inspection Manual Chapter 0612, Power Reactor Inspection Reports, Appendix E, Examples of Minor Issues, in that data recorded during installation of the dryer cavity gate seal was incorrect and resulted in backup air bottle supply pressure left outside the acceptable range. Because the dryer cavity gate seal is intended to contain highly radioactive fluids within containment, which supports the radiological barrier functions to protect plant workers and the public following serious transients or accidents, the inspectors concluded that this issue was associated with the Barrier Integrity Cornerstone. Although this event resulted in a loss of inventory from the dryer cavity pool and partial flooding of the lower reactor cavity and drywell, it was determined to be of very low safety significance because there was no loss inventory from the reactor vessel and it could not result in the loss of reactor coolant system level instrumentation. The inspectors concluded that this finding affected the cross-cutting area of human performance. The licensee did not effectively communicate expectations regarding procedural compliance in this instance and, as a result, the Reactor Services maintenance craftsman did not correctly follow the procedure by performing steps out of sequence and restoring a system to service that was incorrectly aligned.
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| Inspection Report# : 2010003 (pdf)
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| Significance: Jun 30, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform an Adequate 10 CFR 50.59 Evaluation for CPS Procedure 3711.01 - performance deficiency portion The inspectors identified a finding of very low safety significance with an associated NCV of 10 CFR 50.59, Changes, Tests and Experiments. The licensee failed to perform an adequate 10 CFR 50.59 evaluation and obtain a license amendment prior to implementing CPS 3711.01, CPS [Clinton Power Station] Operations with the Potential to Drain the Reactor Vessel [OPDRV], Revision 0 on January 11, 2010. The procedure established a definition of an OPDRV for use in determining the applicability of several TS requirements while in Modes 4 and 5. The licensee failed to recognize that implementing this new procedure, in effect, constituted a change to the TS incorporated into its licensing basis, which would, therefore, require a license amendment pursuant to 10 CFR 50.59(c)(1)(i) and 10 CFR 50.90. No immediate corrective actions were taken to address this finding; however, the licensee entered this issue into its corrective action program for evaluation.
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| The finding was of more than minor significance because there was a reasonable likelihood that the change requiring a 10 CFR 50.59 evaluation would require NRC review and approval prior to implementation. The inspectors assessed the significance of the underlying issue using the SDP. Based on the results of a modified Phase 2 SDP evaluation, this finding was determined to be of very low safety significance. The inspectors concluded that this finding affected the cross cutting area of human performance. Specifically, the licensee did not use conservative decision making to demonstrate that the proposed action did not require prior NRC approval. The inspectors noted that the licensee was aware of potential concerns regarding the new procedure prior to completing the initial 10 CFR 50.59 evaluation and again prior to revising the evaluation in response to concerns raised by the inspectors; however, the incorrect conclusion was reached in both revisions of the evaluation that the new procedure was not a change to the TS and that a license amendment was not necessary.
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| The associated traditional enforcment is tracked as item 2010-003-02.
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| Inspection Report# : 2010003 (pdf)
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| Emergency Preparedness Significance: Jun 30, 2010 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE EMERGENCY PLAN AUGMENTATION CALL-IN DRILLS The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50.54(q) for the licensees failure to follow and maintain the Emergency Plan, which meets the standards in 10 CFR 50.47(b) and the requirements in Appendix E to 10 CFR 50. Specifically, the licensees Emergency Plan calls for the performance of periodic drills to evaluate the ability to augment its Emergency Response Organization (ERO).
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| However, the Emergency Plan implementing procedure used for the conduct of these augmentation drills exempts certain ERO members from participation in these drills, a situation which prevents the licensee from fully demonstrating its ability to augment all the ERO positions in a timely manner. The licensees approved Emergency Plan does not provide for such an exemption. The licensee entered the finding into the corrective action program.
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| The use of an implementing procedure that causes the conduct of an activity to be inconsistent with the associated requirements in the licensees Emergency Plan results in a failure to follow and maintain the Emergency Plan and is a performance deficiency. As a result of the limitations in the procedure, the licensee failed to conduct call-in drills to demonstrate timely augmentation of ERO positions filled by skilled/technical personnel. The deficiency did not impact the NRCs regulatory process or contribute to actual safety consequences; therefore, the performance deficiency was screened using the Emergency Preparedness Significance Determination Process as a failure to comply. The deficiency was determined to be more than minor because the deficiency adversely affected the Emergency Preparedness Cornerstone objective and had the attribute associated with ERO readiness and in the area of ERO augmentation testing. The inspector evaluated the finding using the Inspection Manual Chapter 0609, Appendix B, Sheet I, Failure to Comply Flowchart. The inspector evaluated the finding as a degraded planning standard function since the licensees conduct of the augmentation exercises did not include all ERO positions. The finding was determined to be of very low safety significance. Because the finding did not reflect current licensee performance, no cross-cutting aspect was identified.
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| Inspection Report# : 2010003 (pdf)
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| Occupational Radiation Safety Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.
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| Miscellaneous Last modified : June 07, 2011
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| Clinton 2Q/2011 Plant Inspection Findings Initiating Events Significance: Jun 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO MEET SURVEILLANCE TESTING REQUIREMENT FOR REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES The inspectors identified a finding of very low safety significance (Green) with an associated Non Cited Violation of Technical Specification Surveillance Requirement (TSSR) 3.4.6.1. The licensee failed to correctly incorporate the required test pressure limits of the TSSR into the surveillance test procedure and subsequently tested multiple reactor coolant system (RCS) pressure isolation valves (PIVs) at pressures greater than the maximum test pressure of 1025 pounds-per-square-inch gage, invalidating the testing. The licensee performed a risk assessment of the missed surveillance in accordance with TSSR 3.0.3, which determined that completion of the surveillance could be delayed up to the 24 month surveillance interval without a significant increase in plant risk. The licensee also completed an operability evaluation for the TS nonconformance and concluded that there was reasonable assurance that the affected RCS PIVs were operable based on engineering judgment.
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| The finding was of more than minor significance because it affected the Initiating Events Cornerstone and was associated with the Procedure Quality attribute. Specifically, the licensee did not correctly incorporate the required test pressure limits of TSSR 3.4.6.1 into the surveillance test procedure. This resulted in testing multiple RCS PIVs at pressures greater than the maximum test pressure of 1025 psig. The finding was determined to be a licensee performance deficiency of very low safety significance because the finding would not result in exceeding the TS limit for RCS leakage and would not have likely affected mitigation systems resulting in a loss of safety function. The inspectors concluded that because the licensees missed opportunity to correct the test pressure discrepancy in its surveillance test procedure occurred in January 2005 and no other more recent opportunities reasonably existed to identify and correct the problem, this issue would not be reflective of current licensee performance and no cross-cutting aspect was identified.
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| Inspection Report# : 2011003 (pdf)
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| Significance: Jun 03, 2011 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PERFORM EFFECTIVENESS REVIEW.
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| The inspectors identified a finding of very low safety significance with an associated NCV of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings." The licensee failed to perform an effectiveness review (EFR) to ensure that corrective actions (CAs) taken to prevent recurrence of a significant condition adverse to quality were actually effective to preclude repetition. The licensee entered this violation into its corrective action program as ARs 1221616, 1221661, and 1223806 to investigate the cause and to identify appropriate CAs.
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| The finding was of more than minor significance because it was similar to Example 4a in IMC 0612, "Power Inspection Reports," Appendix E, "Examples of Minor Issues," in that, the licensee routinely failed to perform EFR evaluations on similar CAs related to significant conditions adverse to quality. The finding was a licensee performance deficiency of very low safety significance due to answering 'no' to all questions under the Initiating Events Cornerstone column of IMC 0609 Attachment 4, "Phase 1 - Initial Screening and Characterization of Findings," The inspectors concluded that this finding affected the cross-cutting aspect of problem identification and resolution. Specifically, the licensee failed to thoroughly evaluate problems to include conducting EFRs of CAs to ensure that problems were resolved. [IMC 0310 P.1(c)]
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| Inspection Report# : 2011008 (pdf)
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| Significance: Mar 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO CONTROL TRANSIENT COMBUSTIBLE MATERIALS IN ACCORDANCE WITH FIRE PROTECTION PROGRAM.
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| The inspectors identified a finding of very low safety significance with an associated non-cited violation of the Clinton Power Station Unit 1 Operating License (NPF 62, Section 2.F). The licensee failed to implement the Fire Protection Program in accordance with program requirements by failing to follow approved Fire Protection Program procedures for the control of transient combustible materials. The licensee promptly removed the transient combustible materials found by the inspectors and initiated compensatory measures.
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| The inspectors concluded that this finding could be reasonably viewed as a precursor to a significant event (i.e., a fire affecting more than one train of safe shutdown equipment). Specifically, the presence of transient combustible materials in a combustible free zone could reasonably result in degradation of the fire protection defense in depth elements in place to prevent fires from starting and mitigate the consequences of fires. In addition, based on review of Example 4k in IMC 0612, Power Reactor Inspection Reports, Appendix E, Examples of Minor Issues, the issue would not be considered to be of minor significance because the identified transient combustibles were found in a combustible free zone required for separation of redundant trains. The finding was of very low safety significance because the items found in the combustible free zone would not be considered transient combustibles of significance as defined in IMC 0609, Appendix F, Fire Protection Significance Determination Process, Attachment 2, Degradation Rating Guidance Specific to Various Fire Protection Program Elements, and, therefore, the issue was assigned a low degradation rating. The inspectors concluded that this finding affected the cross cutting area of human performance. Although a pre-job briefing was not required by the licensees procedure for the work activity, job site conditions and a discussion that the work was within a Transient Combustible Free Zone (TCFZ) was not included in the briefing. In addition, the workers 2 Minute Drill performed at the job site did not identify that work activities were within a TCFZ. Therefore, the inspectors concluded that the licensees work practices which support human performance were less than effective (IMC 0301 H.4(a)).
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| Inspection Report# : 2011002 (pdf)
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| Significance: Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO CONTROL TRANSIENT COMBUSTIBLE MATERIALS IN ACCORDANCE WITH FIRE PROTECTION PROGRAM.
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| The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of the Clinton Power Station Unit 1 Operating License (NPF-62, Section 2.F). The licensee failed to implement the Fire Protection Program in accordance with program requirements by failing to follow approved Fire Protection Program procedures for the control of transient combustible materials. The licensee promptly removed the transient combustible materials found by the inspectors.
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| The inspectors concluded that this finding could be reasonably viewed as a precursor to a significant event (i.e., a fire affecting more than one train of safe shutdown equipment). Specifically, the presence of transient combustible materials in a combustible free zone could reasonably result in degradation of the fire protection defense-in-depth elements in place to prevent fires from starting and mitigate the consequences of fires. In addition, based on review of Example 4k in IMC 0612, "Power Reactor Inspection Reports," Appendix E, "Examples of Minor Issues," the issue would not be considered to be of minor significance because the identified transient combustibles were found in a combustible free zone required for separation of redundant trains. The finding was of very low safety significance because the items found in the combustible free zone would not be considered transient combustibles of significance as defined in IMC 0609, Appendix F, "Fire Protection Significance Determination Process," Attachment 2, "Degradation Rating Guidance Specific to Various Fire Protection Program Elements," and, therefore, the issue was assigned a "low degradation" rating. The inspectors condluded that this finding affected the cross-cutting area of human performance. Specifically, the licensee failed to recognize that moving a bullet-resistant container (BRC) was an infrequent activity and, as such, a pre-job briefing should have been performed and was not. In addition, a questioning attitude was not cultivated by the licensee once the correct location of the BRC wash challenged such that
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| security staff proceeded in the face of uncertainty. Therefore, the inspectors concluded that the licensee's work practices that support human performance were less than effective. (IMC 0305 H.4(a))
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| Inspection Report# : 2010005 (pdf)
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| Mitigating Systems Significance: Jun 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation DEFICIENCIES WITH RCIC ROOM HEAT-UP ANALYSES The inspectors identified a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, having very low safety significance for the failure to include all of the applicable heat loads in the reactor core isolation cooling (RCIC) room heat up calculation and not having a calculation of record for the RCIC room heat up under a station blackout (SBO) scenario. The licensee entered this issue into the corrective action program and performed preliminary calculations to verify that the issues did not exceed any design limits.
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| The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of Equipment Performance, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding screened as very low safety significance because the licensee determined the RCIC room cooler was capable of removing the additional heat load; and RCIC room temperature remained within the design limits without the room cooler during a SBO scenario. The inspectors determined that this finding did not represent current licensee performance and no cross-cutting aspect was assigned.
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| Inspection Report# : 2011003 (pdf)
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| Significance: Jun 03, 2011 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO MAINTAIN A QUALITY RECORD AS EVIDENCE OF AN ACTIVITY AFFECTING QUALITY OF SAFETY-RELATED EQUIPMENT DUE TO INAPPROPRIATE CORRECTIVE ACTIONS The inspectors identified a finding of very low safety significance with an associated NCV of 10 CFR Part 50, Appendix B, Criterion XVII, Quality Assurance Records. Specifically, the licensee failed to maintain a quality record documenting a nondestructive examination (NDE) of a safety-related spreader beam lifting device. After losing the original NDE report, the licensees corrective action (CA) was to recreate the report from memory and maintain the recreated report as the quality record. Upon review and questioning from the NRC, the licensee was able to locate the missing NDE report in the records archive. This issue was entered into the licensees CAP as AR1223723.
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| The inspectors determined the finding was more than minor because, if left uncorrected, failure to maintain a quality record as evidence of an activity affecting quality of safety related equipment due to inappropriate disposition of CAs pertaining to missing/lost quality records could become a more significant safety concern. This finding was of very low safety significance because this finding did not represent an actual loss of any safety function of the Mitigation Systems. The inspectors concluded that this finding affected the cross-cutting aspect of Problem Identification and Resolution. Specifically, the licensee did not take appropriate corrective actions to address a lost quality record.
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| Inspection Report# : 2011008 (pdf)
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| Significance: Jun 03, 2011 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ACCOUNT FOR CABLE RESISTANCE IN OPERABILITY DETERMINATIONS.
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| The inspectors identified a finding of very low safety significance with an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," related to calculational errors found in the licensee's operability
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| determination. Specifically, on four separate operability determinations, the licensee failed to account for the cable resistance when determining the maximum allowable contact resistance associated with the second level undervoltage (UV) relays for the 4.16 kV Buses. The licensee entered this violation into its corrective action program as Action Requests 1226340 and 1224313 and performed a preliminary calculation which determined that the error reduced the available margin in the circuit resistance but did not change the overall conclusions for the past operability calls made for the four different occasions.
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| The inspectors determined that this finding was more than minor because it was associated with the Mitigating Systems Cornerstone attribute of design control and adversely affected the cornerstone objective of ensuring availability and reliability of systems that respond to initiating events to prevent undesirable consquences. This finding was of very low safety significance (Green) because the licensee was able to demonstrate that the operability calls that were previously made relating to the second level UV relays were still valid and acceptable. The inspectors condcluded that this finding affected the cross-cutting aspect of human performance. Specifically, the licensee failed to use conservative assumptions in decision making related to immediate operability determinations of conditions adverse to quality. [IMC 0310 H.1(b)]
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| Inspection Report# : 2011008 (pdf)
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| Significance: Mar 18, 2011 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ENSURE FIRE DOOR WAS CLOSED AND LATCHED A finding of very low safety significance and associated NCV of Clinton Power Station Unit 1 Operating License NPF-62, Section 2.F was identified by the inspectors for the licensee's failure to ensure fire doors were closed and latched. Specifically, during a walkdown of fire area CB-1e 737 General Access Area, fire door 1DR1-432 located between fire area CB-1e and D-6 Emergency Diesel 2 Room, was found unlatched/not fully closed. The door was a 3-hour fire rated door credited for fire barrier between the two fire areas. Site personnel closed the door when it was found open and the door remained fully closed when challenged. The issue was entered into the licensee corrective action program as AR 01187906.
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| The inspectors determined that this finding was more than minor because the finding affected the Mitigating Systems cornerstone attributes of protection against external factors (Fire) and affected the cornerstone objective of ensuring the capability of the system to respond to events to prevent undesirable consequences. This finding was of very low safety significance (Green) based on answering Yes to Question 7 of Task 1.3.2. of Appendix F of IMC 0609. The inspectors did not identify a cross-cutting aspect associated with this finding because the underlining cause of unlatched door was indeterminate during the inspection.
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| Inspection Report# : 2011009 (pdf)
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| Barrier Integrity Significance: Mar 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO MEET SURVEILLANCE TESTING REQUIREMENT FOR HYDROGEN IGNITERS IN ACCESSIBLE AREAS OF THE PRIMARY CONTAINMENT AND DRYWELL.
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| The inspectors identified a finding of very low safety significance with an associated non-cited violation of Technical Specification Surveillance Requirement (TSSR) 3.6.3.2.4. The licensee failed to verify that each required hydrogen igniter in accessible areas of the Primary Containment and Drywell develops a surface temperature of greater than or equal to 1700 degrees Fahrenheit (°F) every 24 months. The licensee performed a risk assessment of the missed surveillance in accordance with TSSR 3.0.3, which determined that completion of the surveillance could be delayed up to the 24 month surveillance interval without a significant increase in plant risk. The licensee also completed an operability evaluation for the TS nonconformance and concluded that there was reasonable assurance that the affected hydrogen igniters were operable based on the results of surveillance testing to measure voltage/current draw.
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| The finding was of more than minor significance because it was associated with the Human Performance attribute for the Containment and adversely affected the Barrier Integrity Cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the licensee did not correctly evaluate a change to perform the surveillance test with the unit at power beginning in March 2002. It was not recognized that TSSR 3.6.3.2.4 would not be met for accessible hydrogen igniters in the Drywell and 755 Elevation Steam Tunnel when performing the test with the unit at power and the licensee incorrectly believed that performance of the current/voltage surveillance test procedure for inaccessible igniters was an appropriate substitute, contrary to existing procedural guidance. The finding was a licensee performance deficiency of very low safety significance because it did not involve an actual reduction in the function of hydrogen igniters in the Primary Containment and Drywell. The inspectors concluded that because the scheduling change to perform the surveillance with the unit at power took place prior to surveillance testing beginning in March 2002, it did not necessarily reflect current licensee performance and no cross-cutting aspect was identified.
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| Inspection Report# : 2011002 (pdf)
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| Significance: Mar 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE TESTING CONTROLS TO PERFORM SURVEILLANCE TESTING OF HYDROGEN IGNITERS IN THE PRIMARY CONTAINMENT AND DRYWELL.
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| The inspectors identified a finding of very low safety significance with an associated non-cited violation of 10 CFR 50, Appendix B, Criterion XI, Test Control. The licensee failed to establish a test program adequate to assure testing of hydrogen igniters in accessible areas of the Primary Containment and Drywell pursuant to TSSR 3.6.3.2.4. The licensee entered this violation into its corrective action program to investigate the cause and to identify appropriate corrective actions.
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| The finding was of more than minor significance because it was associated with the Procedure Quality attribute for the Containment and adversely affected the Barrier Integrity Cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The finding was a licensee performance deficiency of very low safety significance because it did not involve an actual reduction in the function of hydrogen igniters in the Primary Containment and Drywell. The inspectors concluded that this finding affected the cross-cutting aspect of human performance. Specifically, adequate licensee resources involving personnel and procedures did not support successful human performance. CPS 9867.05 was not appropriate to the circumstances because it contained errors and did not provide adequate testing controls for the performance of the surveillance test (IMC 0310 H.2(c)).
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| Inspection Report# : 2011002 (pdf)
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| Significance: Dec 31, 2010 Identified By: Self-Revealing Item Type: FIN Finding FAILURE TO PERFORM PREVENTATIVE MAINTENANCE OF DIVISION 1 SELF TEST SYSTEM (STS) POWER SUPPLY RESULTS IN SPURIOUS REPOSITIONING OF SAFETY RELATED VALVES.
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| A finding of very low safety significance was self-revealed on August 24, 2010, when the Reactor Water Cleanup (RT) System return line outboard primary containment isolation valve went closed. Many other unintended valve repositioning events occurred from August 25 through August 26, 2010. The licensee failed to perform preventative maintenance on the Division 1 Self Test System (STS) safety-related 5 Volt (V) power supply. As a result, a degraded voltage condition existed in the test circuit, which was identified as the cause for the above valve repositioning events.
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| As a corrective action, the licensee has since installed a temporary plant modification of dual 5 V power supplies for all four divisions of the STS. No violation of regulatory requirements was identified.
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| The finding was of more than minor significance because the failure to perform preventative maintanance on critical components, if left uncorrected, would potentially lead to a more significant safety concern. This finding was of very low safety significance based on answering "no" to each of the Phase 1 screening questions identified in the
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| Containment Barrier column of Table 4a in Attachment 0609.04, "Phase 1 - Initial Screening and Characterization of Findings." The inspectors concluded that this finding affected the cross-cutting area of human performance.
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| Specifically, in the area of resources the licensee did not adequately maintain long term plant safety by the maintenance of design margins, minimizing preventative maintenance deferrals, and ensuring maintenance and engineering backlogs which are low enough to support safety. (IMC 0310 H.2(a))
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| Inspection Report# : 2010005 (pdf)
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| Emergency Preparedness Significance: SL-IV Jun 22, 2011 Identified By: NRC Item Type: NCV NonCited Violation CHANGES TO EAL BASIS DECREASED THE EFFECTIVENESS OF THE PLAN WITHOUT PRIOR NRC APPROVAL (TRADITIONAL ENFORCEMENT PORTION)
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| The inspector identified a finding of very low safety significance involving a Severity Level IV NCV of 10 CFR 50.54 (q) for failing to obtain prior approval for an emergency plan change which decreased the effectiveness of the plan.
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| Specifically, the licensee modified the Emergency Action Level (EAL) Basis in EAL HU6, Revision 12, which indefinitely extended the start of the 15 minute emergency classification clock beyond a credible notification that a fire is occurring or indication of a valid fire detection system alarm. This change decreased the effectiveness of the emergency plan by reducing the capability to perform a risk significant planning function in a timely manner.
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| The violation affected the NRCs ability to perform its regulatory function because it involved implementing a change that decreased the effectiveness of the emergency plan without NRC approval. Therefore, this issue was evaluated using Traditional Enforcement. The NRC determined that a Severity Level IV violation was appropriate due to the reduction of the capability to perform a risk significant planning standard function in a timely manner. The licensee entered this issue into its corrective action program and revised the EAL basis to restore compliance.
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| The related performance deficiency is tracked as item 2010-502-02.
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| Inspection Report# : 2010502 (pdf)
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| Significance: Jun 22, 2011 Identified By: NRC Item Type: NCV NonCited Violation CHANGES TO EAL BASIS DECREASED THE EFFECTIVENESS OF THE PLAN WITHOUT PRIOR NRC APPROVAL (PERFORMANCE DEFICIENCY PORTION)
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| The inspector identified a finding of very low safety significance involving a Severity Level IV NCV of 10 CFR 50.54 (q) for failing to obtain prior approval for an emergency plan change which decreased the effectiveness of the plan.
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| Specifically, the licensee modified the Emergency Action Level (EAL) Basis in EAL HU6, Revision 12, which indefinitely extended the start of the 15 minute emergency classification clock beyond a credible notification that a fire is occurring or indication of a valid fire detection system alarm. This change decreased the effectiveness of the emergency plan by reducing the capability to perform a risk significant planning function in a timely manner.
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| The finding was more than minor using IMC 0612, because it is associated with the emergency preparedness cornerstone attribute of procedure quality for EAL and emergency plan changes, and it adversely affected the cornerstone objective of ensuring that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Therefore, the performance deficiency was a finding.
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| Using IMC 0609, Appendix B, the inspector determined that the finding had a very low safety significance because the finding is a failure to comply with 10 CFR 50.54(q) involving the risk significant planning standard 50.47(b)(4),
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| which, in this case, met the example of a Green finding because it involved one Unusual Event classification (EAL HU6).
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| The associated traditional enforcement is tracked as item 2010-502-01.
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| Inspection Report# : 2010502 (pdf)
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| Occupational Radiation Safety Public Radiation Safety Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.
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| Miscellaneous Last modified : October 14, 2011
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| Clinton 3Q/2011 Plant Inspection Findings Initiating Events Significance: Jun 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO MEET SURVEILLANCE TESTING REQUIREMENT FOR REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES The inspectors identified a finding of very low safety significance (Green) with an associated Non Cited Violation of Technical Specification Surveillance Requirement (TSSR) 3.4.6.1. The licensee failed to correctly incorporate the required test pressure limits of the TSSR into the surveillance test procedure and subsequently tested multiple reactor coolant system (RCS) pressure isolation valves (PIVs) at pressures greater than the maximum test pressure of 1025 pounds-per-square-inch gage, invalidating the testing. The licensee performed a risk assessment of the missed surveillance in accordance with TSSR 3.0.3, which determined that completion of the surveillance could be delayed up to the 24 month surveillance interval without a significant increase in plant risk. The licensee also completed an operability evaluation for the TS nonconformance and concluded that there was reasonable assurance that the affected RCS PIVs were operable based on engineering judgment.
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| The finding was of more than minor significance because it affected the Initiating Events Cornerstone and was associated with the Procedure Quality attribute. Specifically, the licensee did not correctly incorporate the required test pressure limits of TSSR 3.4.6.1 into the surveillance test procedure. This resulted in testing multiple RCS PIVs at pressures greater than the maximum test pressure of 1025 psig. The finding was determined to be a licensee performance deficiency of very low safety significance because the finding would not result in exceeding the TS limit for RCS leakage and would not have likely affected mitigation systems resulting in a loss of safety function. The inspectors concluded that because the licensees missed opportunity to correct the test pressure discrepancy in its surveillance test procedure occurred in January 2005 and no other more recent opportunities reasonably existed to identify and correct the problem, this issue would not be reflective of current licensee performance and no cross-cutting aspect was identified.
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| Inspection Report# : 2011003 (pdf)
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| Significance: Jun 03, 2011 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PERFORM EFFECTIVENESS REVIEW.
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| The inspectors identified a finding of very low safety significance with an associated NCV of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings." The licensee failed to perform an effectiveness review (EFR) to ensure that corrective actions (CAs) taken to prevent recurrence of a significant condition adverse to quality were actually effective to preclude repetition. The licensee entered this violation into its corrective action program as ARs 1221616, 1221661, and 1223806 to investigate the cause and to identify appropriate CAs.
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| The finding was of more than minor significance because it was similar to Example 4a in IMC 0612, "Power Inspection Reports," Appendix E, "Examples of Minor Issues," in that, the licensee routinely failed to perform EFR evaluations on similar CAs related to significant conditions adverse to quality. The finding was a licensee performance deficiency of very low safety significance due to answering 'no' to all questions under the Initiating Events Cornerstone column of IMC 0609 Attachment 4, "Phase 1 - Initial Screening and Characterization of Findings," The inspectors concluded that this finding affected the cross-cutting aspect of problem identification and resolution. Specifically, the licensee failed to thoroughly evaluate problems to include conducting EFRs of CAs to ensure that problems were resolved. [IMC 0310 P.1(c)]
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| Inspection Report# : 2011008 (pdf)
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| Significance: Mar 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO CONTROL TRANSIENT COMBUSTIBLE MATERIALS IN ACCORDANCE WITH FIRE PROTECTION PROGRAM.
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| The inspectors identified a finding of very low safety significance with an associated non-cited violation of the Clinton Power Station Unit 1 Operating License (NPF 62, Section 2.F). The licensee failed to implement the Fire Protection Program in accordance with program requirements by failing to follow approved Fire Protection Program procedures for the control of transient combustible materials. The licensee promptly removed the transient combustible materials found by the inspectors and initiated compensatory measures.
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| The inspectors concluded that this finding could be reasonably viewed as a precursor to a significant event (i.e., a fire affecting more than one train of safe shutdown equipment). Specifically, the presence of transient combustible materials in a combustible free zone could reasonably result in degradation of the fire protection defense in depth elements in place to prevent fires from starting and mitigate the consequences of fires. In addition, based on review of Example 4k in IMC 0612, Power Reactor Inspection Reports, Appendix E, Examples of Minor Issues, the issue would not be considered to be of minor significance because the identified transient combustibles were found in a combustible free zone required for separation of redundant trains. The finding was of very low safety significance because the items found in the combustible free zone would not be considered transient combustibles of significance as defined in IMC 0609, Appendix F, Fire Protection Significance Determination Process, Attachment 2, Degradation Rating Guidance Specific to Various Fire Protection Program Elements, and, therefore, the issue was assigned a low degradation rating. The inspectors concluded that this finding affected the cross cutting area of human performance. Although a pre-job briefing was not required by the licensees procedure for the work activity, job site conditions and a discussion that the work was within a Transient Combustible Free Zone (TCFZ) was not included in the briefing. In addition, the workers 2 Minute Drill performed at the job site did not identify that work activities were within a TCFZ. Therefore, the inspectors concluded that the licensees work practices which support human performance were less than effective (IMC 0310 H.4(a)).
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| Inspection Report# : 2011002 (pdf)
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| Significance: Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO CONTROL TRANSIENT COMBUSTIBLE MATERIALS IN ACCORDANCE WITH FIRE PROTECTION PROGRAM.
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| The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of the Clinton Power Station Unit 1 Operating License (NPF-62, Section 2.F). The licensee failed to implement the Fire Protection Program in accordance with program requirements by failing to follow approved Fire Protection Program procedures for the control of transient combustible materials. The licensee promptly removed the transient combustible materials found by the inspectors.
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| The inspectors concluded that this finding could be reasonably viewed as a precursor to a significant event (i.e., a fire affecting more than one train of safe shutdown equipment). Specifically, the presence of transient combustible materials in a combustible free zone could reasonably result in degradation of the fire protection defense-in-depth elements in place to prevent fires from starting and mitigate the consequences of fires. In addition, based on review of Example 4k in IMC 0612, "Power Reactor Inspection Reports," Appendix E, "Examples of Minor Issues," the issue would not be considered to be of minor significance because the identified transient combustibles were found in a combustible free zone required for separation of redundant trains. The finding was of very low safety significance because the items found in the combustible free zone would not be considered transient combustibles of significance as defined in IMC 0609, Appendix F, "Fire Protection Significance Determination Process," Attachment 2, "Degradation Rating Guidance Specific to Various Fire Protection Program Elements," and, therefore, the issue was assigned a "low degradation" rating. The inspectors condluded that this finding affected the cross-cutting area of human performance. Specifically, the licensee failed to recognize that moving a bullet-resistant container (BRC) was an infrequent activity and, as such, a pre-job briefing should have been performed and was not. In addition, a questioning attitude was not cultivated by the licensee once the correct location of the BRC wash challenged such that
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| security staff proceeded in the face of uncertainty. Therefore, the inspectors concluded that the licensee's work practices that support human performance were less than effective. (IMC 0305 H.4(a))
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| Inspection Report# : 2010005 (pdf)
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| Mitigating Systems Significance: Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PERFORM CODE REQUIRED CAUSE AND EFFECT FAILURE EVALUATIONS FOR DIESEL STARTING AIR AND FUEL OIL SYSTEM RELIEF VALVES.
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| The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50.55a. The licensee failed to perform American Society of Mechanical Engineers (ASME) Code required cause and effect failure evaluations for set pressure test failures of diesel generator (DG) starting air and fuel oil system relief valves. The licensee entered this issue into its corrective action program for evaluation and subsequently completed an engineering evaluation to address past operability of the associated DG starting air and fuel oil systems due to the relief valve test failures. The licensee also moved up its schedule to test the remaining relief valves.
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| The finding was of more than minor significance because it could lead to a more significant safety concern if left uncorrected. Specifically, the failure to perform Code required cause and effect evaluations for relief valve set pressure test failures could lead to a generic problem with valves in the same or other valve groups remaining uncorrected with a potential impact on operability of safety significant mitigating systems. Because the DG starting air and fuel oil systems are relied upon to support DG operability, the inspectors concluded that this issue was associated with the Mitigating Systems Cornerstone. The finding was determined to be a licensee performance deficiency of very low safety significance because the finding: (1) was not a design or qualification deficiency; (2) did not represent an actual loss of safety function of a system; (3) did not represent an actual loss of safety function of a single train or greater than its Technical Specification (TS) allowed outage time; (4) did not represent an actual loss of safety function of one or more non-TS trains of equipment designated as risk significant; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors concluded that the finding affected the cross-cutting area of human performance in that the licensee's work practices did not ensure adequate supervisory and management oversight of work activities, such that nuclear safety was supported.
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| Specifically, the relief valve test failures were left unresolved and were not evaluated as required by the Code for an extended period of time with several failed tests. (IMC 0310, H.4(C))
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| Inspection Report# : 2011004 (pdf)
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| Significance: Sep 30, 2011 Identified By: NRC Item Type: FIN Finding FAILURE TO CORRECT A CONDITION ADVERSE TO QUALITY FOR IMPROPERLY IMPLEMENTED ENGINEERING CORRECTIVE ACTIONS The inspectors identified a finding of very low safety significance due to the licensee's failure to effectively implement corrective actions for a condition adverse to quality described in Apparent Cause Evaluation 1095413, "NOS [Nuclear Oversight] Identified Improperly Implemented Engineering Corrective Actions Cause Repeat Operational Challenges." No violation of regulatory requirements was identified. The licensee entered this issue into its corrective action program to investigate the cause and to identify appropriate corrective actions.
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| The finding was of more than minor significance because it was associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone and directly affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.
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| Specifically, improperly implemented engineering corrective actions could result in additional repeat operational equipment challenges. The finding was of very low safety significance because the issue: (1) was not a design or qualification deficiency; (2) did not represent an actual loss of safety function of a system; (3) did not represent an actual loss of safety function of a single train for greater than its Technical Specification (TS) allowed outage time; (4)
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| did not represent an actual loss of safety function of one or more non-TS trains of equipment designated as risk significant; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors concluded that this finding affected the cross-cutting area of problem identification and resolution. Specifically, the licensee failed to take appropriate corrective actions to address known deficiencies in its process for tracking and closing work orders that implement corrective actions. The actions taken were neither lasting nor effective. (IMC 0310, P.1(d))
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| Inspection Report# : 2011004 (pdf)
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| Significance: Jun 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation DEFICIENCIES WITH RCIC ROOM HEAT-UP ANALYSES The inspectors identified a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, having very low safety significance for the failure to include all of the applicable heat loads in the reactor core isolation cooling (RCIC) room heat up calculation and not having a calculation of record for the RCIC room heat up under a station blackout (SBO) scenario. The licensee entered this issue into the corrective action program and performed preliminary calculations to verify that the issues did not exceed any design limits.
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| The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of Equipment Performance, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding screened as very low safety significance because the licensee determined the RCIC room cooler was capable of removing the additional heat load; and RCIC room temperature remained within the design limits without the room cooler during a SBO scenario. The inspectors determined that this finding did not represent current licensee performance and no cross-cutting aspect was assigned.
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| Inspection Report# : 2011003 (pdf)
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| Significance: Jun 03, 2011 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO MAINTAIN A QUALITY RECORD AS EVIDENCE OF AN ACTIVITY AFFECTING QUALITY OF SAFETY-RELATED EQUIPMENT DUE TO INAPPROPRIATE CORRECTIVE ACTIONS The inspectors identified a finding of very low safety significance with an associated NCV of 10 CFR Part 50, Appendix B, Criterion XVII, Quality Assurance Records. Specifically, the licensee failed to maintain a quality record documenting a nondestructive examination (NDE) of a safety-related spreader beam lifting device. After losing the original NDE report, the licensees corrective action (CA) was to recreate the report from memory and maintain the recreated report as the quality record. Upon review and questioning from the NRC, the licensee was able to locate the missing NDE report in the records archive. This issue was entered into the licensees CAP as AR1223723.
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| The inspectors determined the finding was more than minor because, if left uncorrected, failure to maintain a quality record as evidence of an activity affecting quality of safety related equipment due to inappropriate disposition of CAs pertaining to missing/lost quality records could become a more significant safety concern. This finding was of very low safety significance because this finding did not represent an actual loss of any safety function of the Mitigation Systems. The inspectors concluded that this finding affected the cross-cutting aspect of Problem Identification and Resolution. Specifically, the licensee did not take appropriate corrective actions to address a lost quality record. [IMC 0310 P.1(d)]
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| Inspection Report# : 2011008 (pdf)
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| Significance: Jun 03, 2011 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ACCOUNT FOR CABLE RESISTANCE IN OPERABILITY DETERMINATIONS.
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| The inspectors identified a finding of very low safety significance with an associated NCV of 10 CFR Part 50,
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| Appendix B, Criterion III, "Design Control," related to calculational errors found in the licensee's operability determination. Specifically, on four separate operability determinations, the licensee failed to account for the cable resistance when determining the maximum allowable contact resistance associated with the second level undervoltage (UV) relays for the 4.16 kV Buses. The licensee entered this violation into its corrective action program as Action Requests 1226340 and 1224313 and performed a preliminary calculation which determined that the error reduced the available margin in the circuit resistance but did not change the overall conclusions for the past operability calls made for the four different occasions.
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| The inspectors determined that this finding was more than minor because it was associated with the Mitigating Systems Cornerstone attribute of design control and adversely affected the cornerstone objective of ensuring availability and reliability of systems that respond to initiating events to prevent undesirable consquences. This finding was of very low safety significance (Green) because the licensee was able to demonstrate that the operability calls that were previously made relating to the second level UV relays were still valid and acceptable. The inspectors condcluded that this finding affected the cross-cutting aspect of human performance. Specifically, the licensee failed to use conservative assumptions in decision making related to immediate operability determinations of conditions adverse to quality. [IMC 0310 H.1(b)]
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| Inspection Report# : 2011008 (pdf)
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| Significance: Mar 18, 2011 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ENSURE FIRE DOOR WAS CLOSED AND LATCHED A finding of very low safety significance and associated NCV of Clinton Power Station Unit 1 Operating License NPF-62, Section 2.F was identified by the inspectors for the licensee's failure to ensure fire doors were closed and latched. Specifically, during a walkdown of fire area CB-1e 737 General Access Area, fire door 1DR1-432 located between fire area CB-1e and D-6 Emergency Diesel 2 Room, was found unlatched/not fully closed. The door was a 3-hour fire rated door credited for fire barrier between the two fire areas. Site personnel closed the door when it was found open and the door remained fully closed when challenged. The issue was entered into the licensee corrective action program as AR 01187906.
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| The inspectors determined that this finding was more than minor because the finding affected the Mitigating Systems cornerstone attributes of protection against external factors (Fire) and affected the cornerstone objective of ensuring the capability of the system to respond to events to prevent undesirable consequences. This finding was of very low safety significance (Green) based on answering Yes to Question 7 of Task 1.3.2. of Appendix F of IMC 0609. The inspectors did not identify a cross-cutting aspect associated with this finding because the underlining cause of unlatched door was indeterminate during the inspection.
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| Inspection Report# : 2011009 (pdf)
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| Barrier Integrity Significance: Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO MEET TECHNICAL SPECIFICATION 3.7.3 FOR OPERABILITY OF CONTROL ROOM VENTILATION SYSTEM The inspectors identified a finding of very low safety significance with an associated non-cited violation of Technical Specification (TS) 3.7.3, "Control Room Ventilation System," following the discovery of a crack on the Train B Control Room ventilation (VC) system return fan hub during investigation of the cause for high noise and vibration levels observed on May 23, 2011. The licensee failed to correctly evaluate the operability of the Train B VC system return fan in a timely manner to prevent exceeding the TS allowed outage time for entry into Mode 3. The licensee replaced the fan and returned it to an operable status.
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| The failure to correctly evaluate a degraded/nonconforming condition potentially affecting the operability of structures, systems, and components (SSCs) required to be operable by TS would become a more significant safety concern if left uncorrected because it could reasonably result in an unrecognized condition of an SSC failing to fulfill a safety-related function. The finding was, therefore, of more than minor significance. Because the Control Room ventilation system supports the radiological barrier function to protect operators inside the Control Room following certain design basis accidents, the inspectors concluded that this issue was associated with the Barrier Integrity Cornerstone. The finding was a licensee performance deficiency of very low safety significance because it involved only a degradation of the ridiological barrier function provided for the Control Room. The inspectors concluded that this finding affected the cross-cutting area of human performance. Specifically, the licensee decision making to delay inspection of the fan hub and blades until after a new fan was delivered on site to confirm the initial operability determination was not conservative and not consistent with demonstrating that nuclear safety is an overriding priority.
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| (IMC 0310, H.1(b))
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| Inspection Report# : 2011004 (pdf)
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| Significance: Sep 30, 2011 Identified By: NRC Item Type: FIN Finding FAILURE TO EVALUATE OPERABILITY OF CONTROL ROOM VENTILATION SYSTEM FOR DEGRADED FLOW CONDITION The inspectors identified a finding of very low safety significance. The licensee failed to appropriately evaluate the operability of Control Room Ventilation Train A after identifying a degraded/nonconforming system flow condition while performing surveillance testing on April 1, 2011, that could have affected the ability of the system to perform its safety function. No violation of regulatory requirements was identified. The licensee initiated corrective actions to provide "read & sign" training for licensed operators and a procedure change to add an acceptance criterion for filtered flow rate in the surveillance test procedure.
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| The failure to correctly evaluate a degraded/nonconforming condition potentially affecting the operability of structures, systems, and components (SSCs) required to be operable by Technical Specifications (TS) would become a more significant safety concern if left uncorrected because it could reasonably result in an unrecognized condition of an SSC failing to fulfill a safety-related function. The finding was therefore of more than minor significance. Because the Control Room ventilation system supports the radiological barrier function to protect operators inside the Control Room following certain design basis accidents, the inspectors concluded that this issue was associated with the Barrier Integrity Cornerstone. The finding was a licensee performance deficiency of very low safety significance because it involved only a degradation of the radiological barrier function provided for the Control Room. The inspectors concluded that this finding affected the cross-cutting area of human performance. Specifically, licensee decision making using a systematic process to evaluate the operability of an SSC required to be operable by TS when a degraded/nonconforming condition was identified was not appropriately implemented as designed by licensed senior reactor operators. (IMC 0310 H.1(a))
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| Inspection Report# : 2011004 (pdf)
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| Significance: Mar 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO MEET SURVEILLANCE TESTING REQUIREMENT FOR HYDROGEN IGNITERS IN ACCESSIBLE AREAS OF THE PRIMARY CONTAINMENT AND DRYWELL.
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| The inspectors identified a finding of very low safety significance with an associated non-cited violation of Technical Specification Surveillance Requirement (TSSR) 3.6.3.2.4. The licensee failed to verify that each required hydrogen igniter in accessible areas of the Primary Containment and Drywell develops a surface temperature of greater than or equal to 1700 degrees Fahrenheit (°F) every 24 months. The licensee performed a risk assessment of the missed surveillance in accordance with TSSR 3.0.3, which determined that completion of the surveillance could be delayed up to the 24 month surveillance interval without a significant increase in plant risk. The licensee also completed an operability evaluation for the TS nonconformance and concluded that there was reasonable assurance that the affected hydrogen igniters were operable based on the results of surveillance testing to measure voltage/current draw.
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| The finding was of more than minor significance because it was associated with the Human Performance attribute for
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| the Containment and adversely affected the Barrier Integrity Cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the licensee did not correctly evaluate a change to perform the surveillance test with the unit at power beginning in March 2002. It was not recognized that TSSR 3.6.3.2.4 would not be met for accessible hydrogen igniters in the Drywell and 755 Elevation Steam Tunnel when performing the test with the unit at power and the licensee incorrectly believed that performance of the current/voltage surveillance test procedure for inaccessible igniters was an appropriate substitute, contrary to existing procedural guidance. The finding was a licensee performance deficiency of very low safety significance because it did not involve an actual reduction in the function of hydrogen igniters in the Primary Containment and Drywell. The inspectors concluded that because the scheduling change to perform the surveillance with the unit at power took place prior to surveillance testing beginning in March 2002, it did not necessarily reflect current licensee performance and no cross-cutting aspect was identified.
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| Inspection Report# : 2011002 (pdf)
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| Significance: Mar 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE TESTING CONTROLS TO PERFORM SURVEILLANCE TESTING OF HYDROGEN IGNITERS IN THE PRIMARY CONTAINMENT AND DRYWELL.
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| The inspectors identified a finding of very low safety significance with an associated non-cited violation of 10 CFR 50, Appendix B, Criterion XI, Test Control. The licensee failed to establish a test program adequate to assure testing of hydrogen igniters in accessible areas of the Primary Containment and Drywell pursuant to TSSR 3.6.3.2.4. The licensee entered this violation into its corrective action program to investigate the cause and to identify appropriate corrective actions.
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| The finding was of more than minor significance because it was associated with the Procedure Quality attribute for the Containment and adversely affected the Barrier Integrity Cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The finding was a licensee performance deficiency of very low safety significance because it did not involve an actual reduction in the function of hydrogen igniters in the Primary Containment and Drywell. The inspectors concluded that this finding affected the cross-cutting aspect of human performance. Specifically, adequate licensee resources involving personnel and procedures did not support successful human performance. CPS 9867.05 was not appropriate to the circumstances because it contained errors and did not provide adequate testing controls for the performance of the surveillance test (IMC 0310 H.2(c)).
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| Inspection Report# : 2011002 (pdf)
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| Significance: Dec 31, 2010 Identified By: Self-Revealing Item Type: FIN Finding FAILURE TO PERFORM PREVENTATIVE MAINTENANCE OF DIVISION 1 SELF TEST SYSTEM (STS) POWER SUPPLY RESULTS IN SPURIOUS REPOSITIONING OF SAFETY RELATED VALVES.
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| A finding of very low safety significance was self-revealed on August 24, 2010, when the Reactor Water Cleanup (RT) System return line outboard primary containment isolation valve went closed. Many other unintended valve repositioning events occurred from August 25 through August 26, 2010. The licensee failed to perform preventative maintenance on the Division 1 Self Test System (STS) safety-related 5 Volt (V) power supply. As a result, a degraded voltage condition existed in the test circuit, which was identified as the cause for the above valve repositioning events.
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| As a corrective action, the licensee has since installed a temporary plant modification of dual 5 V power supplies for all four divisions of the STS. No violation of regulatory requirements was identified.
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| The finding was of more than minor significance because the failure to perform preventative maintanance on critical components, if left uncorrected, would potentially lead to a more significant safety concern. This finding was of very low safety significance based on answering "no" to each of the Phase 1 screening questions identified in the Containment Barrier column of Table 4a in Attachment 0609.04, "Phase 1 - Initial Screening and Characterization of Findings." The inspectors concluded that this finding affected the cross-cutting area of human performance.
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| Specifically, in the area of resources the licensee did not adequately maintain long term plant safety by the maintenance of design margins, minimizing preventative maintenance deferrals, and ensuring maintenance and engineering backlogs which are low enough to support safety. (IMC 0310 H.2(a))
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| Inspection Report# : 2010005 (pdf)
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| Emergency Preparedness Significance: Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation MISSING RESPIRATOR SPECTACLE KITS The inspectors identified a finding of very low safety significance and associated NCV of 10 CFR 50.54(q) for the failure to provide spectacle adapter kits for all eyeglass wearers (i.e, non-soft contact wearers) who were key emergency response organization (ERO) personnel that were potentially required to wear a self-contained breathing apparatus (SCBA) in order to fulfill emergency response functions. The licensee's corrective actions included revising procedures that govern the training and qualification of licensed operators to include steps that ensure licensed operators and other ERO members who require corrective lenses are provided SCBA lens inserts.
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| The finding was more than minor because this is associated with the Emergency Planning cornerstone, and if left uncorrected, the performance deficiency has the potential to lead to a more significant safety concern, in that, emergency responders having inadequate vision could challenge the licensee's state of operational readiness and emergency response capabilities. The finding was assessed using IMC 0609, Attachment B Emergency Planning Significance Determination Process (SDP) and determined to be of very low safety significance because this failure to comply represented a planning standard issue, however it did not result in a risk significant planning standard nor was it indicative of a planning standard functional failure. The failure to make provisions for respirator vision corrective lenses to licensed operators that required corrective lenses as a condition of their license was caused by a program weakness. Consequently, the cause of this finding has a cross-cutting aspect in the area of human performance.
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| Specifically, the licensee did not ensure that equipment was available for key emergency response personnel. (IMC 0310,H.2(d))
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| Inspection Report# : 2011004 (pdf)
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| Significance: SL-IV Jun 22, 2011 Identified By: NRC Item Type: NCV NonCited Violation CHANGES TO EAL BASIS DECREASED THE EFFECTIVENESS OF THE PLAN WITHOUT PRIOR NRC APPROVAL (TRADITIONAL ENFORCEMENT PORTION)
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| The inspector identified a finding of very low safety significance involving a Severity Level IV NCV of 10 CFR 50.54 (q) for failing to obtain prior approval for an emergency plan change which decreased the effectiveness of the plan.
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| Specifically, the licensee modified the Emergency Action Level (EAL) Basis in EAL HU6, Revision 12, which indefinitely extended the start of the 15 minute emergency classification clock beyond a credible notification that a fire is occurring or indication of a valid fire detection system alarm. This change decreased the effectiveness of the emergency plan by reducing the capability to perform a risk significant planning function in a timely manner.
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| The violation affected the NRCs ability to perform its regulatory function because it involved implementing a change that decreased the effectiveness of the emergency plan without NRC approval. Therefore, this issue was evaluated using Traditional Enforcement. The NRC determined that a Severity Level IV violation was appropriate due to the reduction of the capability to perform a risk significant planning standard function in a timely manner. The licensee entered this issue into its corrective action program and revised the EAL basis to restore compliance.
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| The related performance deficiency is tracked as item 2010-502-02.
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| Inspection Report# : 2010502 (pdf)
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| Significance: Jun 22, 2011
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| Identified By: NRC Item Type: NCV NonCited Violation CHANGES TO EAL BASIS DECREASED THE EFFECTIVENESS OF THE PLAN WITHOUT PRIOR NRC APPROVAL (PERFORMANCE DEFICIENCY PORTION)
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| The inspector identified a finding of very low safety significance involving a Severity Level IV NCV of 10 CFR 50.54 (q) for failing to obtain prior approval for an emergency plan change which decreased the effectiveness of the plan.
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| Specifically, the licensee modified the Emergency Action Level (EAL) Basis in EAL HU6, Revision 12, which indefinitely extended the start of the 15 minute emergency classification clock beyond a credible notification that a fire is occurring or indication of a valid fire detection system alarm. This change decreased the effectiveness of the emergency plan by reducing the capability to perform a risk significant planning function in a timely manner.
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| The finding was more than minor using IMC 0612, because it is associated with the emergency preparedness cornerstone attribute of procedure quality for EAL and emergency plan changes, and it adversely affected the cornerstone objective of ensuring that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Therefore, the performance deficiency was a finding.
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| Using IMC 0609, Appendix B, the inspector determined that the finding had a very low safety significance because the finding is a failure to comply with 10 CFR 50.54(q) involving the risk significant planning standard 50.47(b)(4),
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| which, in this case, met the example of a Green finding because it involved one Unusual Event classification (EAL HU6).
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| The associated traditional enforcement is tracked as item 2010-502-01.
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| Inspection Report# : 2010502 (pdf)
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| Occupational Radiation Safety Public Radiation Safety Significance: Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IMPLEMENT PACKAGE DESIGN SPECIFICATIONS The inspectors identified a finding of very low-safety-significance and an associated non-cited violation of 10 CFR 71.5 for the failure to implement package design specifications. Specifically, the licensee failed to ensure the proper closure of a DOT 7A Type A package as required by Department of Transportation (DOT) regulations for packaging contained within 49 CFR 173. As a part of their corrective actions, the licensee completed a detailed review of all radioactive material shipments for the past 36 months to ensure Package Certification documents for other packages used as a Type A container satisfied requirements.
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| The finding was more than minor because it affected the Public Radiation Safety Cornerstone objective to ensure adequate protection of public health and safety from exposure to radioactive materials during transit. Specifically, the failure to correctly close a DOT Type A package could lead to a more significant safety concern by increasing the potential for a package breach occuring during transit. Using IMC 0609, Attachment D for the Public Radiation Safety Significance Determination Process (SDP) the inspectors determined the finding to be of very low safety significance.
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| This deficiency has a cross-cutting aspect in Human Performance (Resources). (IMC 0310 H.2(b))
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| Inspection Report# : 2011004 (pdf)
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| Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings
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| pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.
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| Miscellaneous Last modified : January 04, 2012
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| Clinton 4Q/2011 Plant Inspection Findings Initiating Events Significance: Jun 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO MEET SURVEILLANCE TESTING REQUIREMENT FOR REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES The inspectors identified a finding of very low safety significance (Green) with an associated Non Cited Violation of Technical Specification Surveillance Requirement (TSSR) 3.4.6.1. The licensee failed to correctly incorporate the required test pressure limits of the TSSR into the surveillance test procedure and subsequently tested multiple reactor coolant system (RCS) pressure isolation valves (PIVs) at pressures greater than the maximum test pressure of 1025 pounds-per-square-inch gage, invalidating the testing. The licensee performed a risk assessment of the missed surveillance in accordance with TSSR 3.0.3, which determined that completion of the surveillance could be delayed up to the 24 month surveillance interval without a significant increase in plant risk. The licensee also completed an operability evaluation for the TS nonconformance and concluded that there was reasonable assurance that the affected RCS PIVs were operable based on engineering judgment.
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| The finding was of more than minor significance because it affected the Initiating Events Cornerstone and was associated with the Procedure Quality attribute. Specifically, the licensee did not correctly incorporate the required test pressure limits of TSSR 3.4.6.1 into the surveillance test procedure. This resulted in testing multiple RCS PIVs at pressures greater than the maximum test pressure of 1025 psig. The finding was determined to be a licensee performance deficiency of very low safety significance because the finding would not result in exceeding the TS limit for RCS leakage and would not have likely affected mitigation systems resulting in a loss of safety function. The inspectors concluded that because the licensees missed opportunity to correct the test pressure discrepancy in its surveillance test procedure occurred in January 2005 and no other more recent opportunities reasonably existed to identify and correct the problem, this issue would not be reflective of current licensee performance and no cross-cutting aspect was identified.
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| Inspection Report# : 2011003 (pdf)
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| Significance: Jun 03, 2011 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PERFORM EFFECTIVENESS REVIEW.
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| The inspectors identified a finding of very low safety significance with an associated NCV of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings." The licensee failed to perform an effectiveness review (EFR) to ensure that corrective actions (CAs) taken to prevent recurrence of a significant condition adverse to quality were actually effective to preclude repetition. The licensee entered this violation into its corrective action program as ARs 1221616, 1221661, and 1223806 to investigate the cause and to identify appropriate CAs.
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| The finding was of more than minor significance because it was similar to Example 4a in IMC 0612, "Power Inspection Reports," Appendix E, "Examples of Minor Issues," in that, the licensee routinely failed to perform EFR evaluations on similar CAs related to significant conditions adverse to quality. The finding was a licensee performance deficiency of very low safety significance due to answering 'no' to all questions under the Initiating Events Cornerstone column of IMC 0609 Attachment 4, "Phase 1 - Initial Screening and Characterization of Findings," The inspectors concluded that this finding affected the cross-cutting aspect of problem identification and resolution. Specifically, the licensee failed to thoroughly evaluate problems to include conducting EFRs of CAs to ensure that problems were resolved. [IMC 0310 P.1(c)]
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| Inspection Report# : 2011008 (pdf)
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| Significance: Mar 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO CONTROL TRANSIENT COMBUSTIBLE MATERIALS IN ACCORDANCE WITH FIRE PROTECTION PROGRAM.
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| The inspectors identified a finding of very low safety significance with an associated non-cited violation of the Clinton Power Station Unit 1 Operating License (NPF 62, Section 2.F). The licensee failed to implement the Fire Protection Program in accordance with program requirements by failing to follow approved Fire Protection Program procedures for the control of transient combustible materials. The licensee promptly removed the transient combustible materials found by the inspectors and initiated compensatory measures.
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| The inspectors concluded that this finding could be reasonably viewed as a precursor to a significant event (i.e., a fire affecting more than one train of safe shutdown equipment). Specifically, the presence of transient combustible materials in a combustible free zone could reasonably result in degradation of the fire protection defense in depth elements in place to prevent fires from starting and mitigate the consequences of fires. In addition, based on review of Example 4k in IMC 0612, Power Reactor Inspection Reports, Appendix E, Examples of Minor Issues, the issue would not be considered to be of minor significance because the identified transient combustibles were found in a combustible free zone required for separation of redundant trains. The finding was of very low safety significance because the items found in the combustible free zone would not be considered transient combustibles of significance as defined in IMC 0609, Appendix F, Fire Protection Significance Determination Process, Attachment 2, Degradation Rating Guidance Specific to Various Fire Protection Program Elements, and, therefore, the issue was assigned a low degradation rating. The inspectors concluded that this finding affected the cross cutting area of human performance. Although a pre-job briefing was not required by the licensees procedure for the work activity, job site conditions and a discussion that the work was within a Transient Combustible Free Zone (TCFZ) was not included in the briefing. In addition, the workers 2 Minute Drill performed at the job site did not identify that work activities were within a TCFZ. Therefore, the inspectors concluded that the licensees work practices which support human performance were less than effective (IMC 0310 H.4(a)).
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| Inspection Report# : 2011002 (pdf)
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| Mitigating Systems Significance: Dec 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PROPERLY APPLY AN APPROVED ASME CODE CASE The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50.55a due to the licensee's failure to adequately apply American Society of Mechanical Engineers Section XI Code Case N-513-3 when it evaluated a degraded section of safety related shutdown service water system piping for operability. Specifically, the licensee failed to perform required daily walkdowns to confirm its analysis of conditions used in its operability evaluation remained valid. After this issue was identified by the inspectors, the licensee promptly resumed the daily compensatory action to verify the leak rate until the piping system was repaired.
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| The finding was of more than minor significance because it was associated with the Protection Against External Factors attribute of the Mitigating Systems Cornerstone, and it directly affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Additionally, if left uncorrected, improper application of an approved code case would become a more significant safety concern in that it could result in the failure to identify inoperable safety related piping. The finding was a licensee performance deficiency of very low safety significance because it did not result in an actual loss of safety function of a single train for greater than its Technical Specification allowed outage time. The inspectors concluded that there was no specific performance characteristic that was a significant cause to the performance deficiency in this instance; therefore no cross-cutting aspect was identified.
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| Inspection Report# : 2011005 (pdf)
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| Significance: Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PERFORM CODE REQUIRED CAUSE AND EFFECT FAILURE EVALUATIONS FOR DIESEL STARTING AIR AND FUEL OIL SYSTEM RELIEF VALVES.
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| The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50.55a. The licensee failed to perform American Society of Mechanical Engineers (ASME) Code required cause and effect failure evaluations for set pressure test failures of diesel generator (DG) starting air and fuel oil system relief valves. The licensee entered this issue into its corrective action program for evaluation and subsequently completed an engineering evaluation to address past operability of the associated DG starting air and fuel oil systems due to the relief valve test failures. The licensee also moved up its schedule to test the remaining relief valves.
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| The finding was of more than minor significance because it could lead to a more significant safety concern if left uncorrected. Specifically, the failure to perform Code required cause and effect evaluations for relief valve set pressure test failures could lead to a generic problem with valves in the same or other valve groups remaining uncorrected with a potential impact on operability of safety significant mitigating systems. Because the DG starting air and fuel oil systems are relied upon to support DG operability, the inspectors concluded that this issue was associated with the Mitigating Systems Cornerstone. The finding was determined to be a licensee performance deficiency of very low safety significance because the finding: (1) was not a design or qualification deficiency; (2) did not represent an actual loss of safety function of a system; (3) did not represent an actual loss of safety function of a single train or greater than its Technical Specification (TS) allowed outage time; (4) did not represent an actual loss of safety function of one or more non-TS trains of equipment designated as risk significant; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors concluded that the finding affected the cross-cutting area of human performance in that the licensee's work practices did not ensure adequate supervisory and management oversight of work activities, such that nuclear safety was supported.
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| Specifically, the relief valve test failures were left unresolved and were not evaluated as required by the Code for an extended period of time with several failed tests. (IMC 0310, H.4(C))
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| Inspection Report# : 2011004 (pdf)
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| Significance: Sep 30, 2011 Identified By: NRC Item Type: FIN Finding FAILURE TO CORRECT A CONDITION ADVERSE TO QUALITY FOR IMPROPERLY IMPLEMENTED ENGINEERING CORRECTIVE ACTIONS The inspectors identified a finding of very low safety significance due to the licensee's failure to effectively implement corrective actions for a condition adverse to quality described in Apparent Cause Evaluation 1095413, "NOS [Nuclear Oversight] Identified Improperly Implemented Engineering Corrective Actions Cause Repeat Operational Challenges." No violation of regulatory requirements was identified. The licensee entered this issue into its corrective action program to investigate the cause and to identify appropriate corrective actions.
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| The finding was of more than minor significance because it was associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone and directly affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.
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| Specifically, improperly implemented engineering corrective actions could result in additional repeat operational equipment challenges. The finding was of very low safety significance because the issue: (1) was not a design or qualification deficiency; (2) did not represent an actual loss of safety function of a system; (3) did not represent an actual loss of safety function of a single train for greater than its Technical Specification (TS) allowed outage time; (4) did not represent an actual loss of safety function of one or more non-TS trains of equipment designated as risk significant; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors concluded that this finding affected the cross-cutting area of problem identification and resolution. Specifically, the licensee failed to take appropriate corrective actions to address known deficiencies in its process for tracking and closing work orders that implement corrective actions. The actions taken were neither lasting nor effective. (IMC 0310, P.1(d))
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| Inspection Report# : 2011004 (pdf)
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| Significance: Jun 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation DEFICIENCIES WITH RCIC ROOM HEAT-UP ANALYSES The inspectors identified a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, having very low safety significance for the failure to include all of the applicable heat loads in the reactor core isolation cooling (RCIC) room heat up calculation and not having a calculation of record for the RCIC room heat up under a station blackout (SBO) scenario. The licensee entered this issue into the corrective action program and performed preliminary calculations to verify that the issues did not exceed any design limits.
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| The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of Equipment Performance, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding screened as very low safety significance because the licensee determined the RCIC room cooler was capable of removing the additional heat load; and RCIC room temperature remained within the design limits without the room cooler during a SBO scenario. The inspectors determined that this finding did not represent current licensee performance and no cross-cutting aspect was assigned.
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| Inspection Report# : 2011003 (pdf)
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| Significance: Jun 03, 2011 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO MAINTAIN A QUALITY RECORD AS EVIDENCE OF AN ACTIVITY AFFECTING QUALITY OF SAFETY-RELATED EQUIPMENT DUE TO INAPPROPRIATE CORRECTIVE ACTIONS The inspectors identified a finding of very low safety significance with an associated NCV of 10 CFR Part 50, Appendix B, Criterion XVII, Quality Assurance Records. Specifically, the licensee failed to maintain a quality record documenting a nondestructive examination (NDE) of a safety-related spreader beam lifting device. After losing the original NDE report, the licensees corrective action (CA) was to recreate the report from memory and maintain the recreated report as the quality record. Upon review and questioning from the NRC, the licensee was able to locate the missing NDE report in the records archive. This issue was entered into the licensees CAP as AR1223723.
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| The inspectors determined the finding was more than minor because, if left uncorrected, failure to maintain a quality record as evidence of an activity affecting quality of safety related equipment due to inappropriate disposition of CAs pertaining to missing/lost quality records could become a more significant safety concern. This finding was of very low safety significance because this finding did not represent an actual loss of any safety function of the Mitigation Systems. The inspectors concluded that this finding affected the cross-cutting aspect of Problem Identification and Resolution. Specifically, the licensee did not take appropriate corrective actions to address a lost quality record. [IMC 0310 P.1(d)]
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| Inspection Report# : 2011008 (pdf)
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| Significance: Jun 03, 2011 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ACCOUNT FOR CABLE RESISTANCE IN OPERABILITY DETERMINATIONS.
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| The inspectors identified a finding of very low safety significance with an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," related to calculational errors found in the licensee's operability determination. Specifically, on four separate operability determinations, the licensee failed to account for the cable resistance when determining the maximum allowable contact resistance associated with the second level undervoltage (UV) relays for the 4.16 kV Buses. The licensee entered this violation into its corrective action program as Action Requests 1226340 and 1224313 and performed a preliminary calculation which determined that the error reduced the available margin in the circuit resistance but did not change the overall conclusions for the past operability calls made for the four different occasions.
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| The inspectors determined that this finding was more than minor because it was associated with the Mitigating Systems Cornerstone attribute of design control and adversely affected the cornerstone objective of ensuring availability and reliability of systems that respond to initiating events to prevent undesirable consquences. This finding was of very low safety significance (Green) because the licensee was able to demonstrate that the operability calls that were previously made relating to the second level UV relays were still valid and acceptable. The inspectors condcluded that this finding affected the cross-cutting aspect of human performance. Specifically, the licensee failed to use conservative assumptions in decision making related to immediate operability determinations of conditions adverse to quality. [IMC 0310 H.1(b)]
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| Inspection Report# : 2011008 (pdf)
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| Significance: Mar 18, 2011 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ENSURE FIRE DOOR WAS CLOSED AND LATCHED A finding of very low safety significance and associated NCV of Clinton Power Station Unit 1 Operating License NPF-62, Section 2.F was identified by the inspectors for the licensee's failure to ensure fire doors were closed and latched. Specifically, during a walkdown of fire area CB-1e 737 General Access Area, fire door 1DR1-432 located between fire area CB-1e and D-6 Emergency Diesel 2 Room, was found unlatched/not fully closed. The door was a 3-hour fire rated door credited for fire barrier between the two fire areas. Site personnel closed the door when it was found open and the door remained fully closed when challenged. The issue was entered into the licensee corrective action program as AR 01187906.
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| The inspectors determined that this finding was more than minor because the finding affected the Mitigating Systems cornerstone attributes of protection against external factors (Fire) and affected the cornerstone objective of ensuring the capability of the system to respond to events to prevent undesirable consequences. This finding was of very low safety significance (Green) based on answering Yes to Question 7 of Task 1.3.2. of Appendix F of IMC 0609. The inspectors did not identify a cross-cutting aspect associated with this finding because the underlining cause of unlatched door was indeterminate during the inspection.
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| Inspection Report# : 2011009 (pdf)
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| Barrier Integrity Significance: Dec 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO CONTROL THE WORK HOURS OF A COVERED WORKER.
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| The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 26.205(c) and (d) for the licensee's failure to schedule and control the work hours of a covered worker performing surveillance testing on containment isolation valves during the refueling outage. Specifically, an engineer performing local leak rate testing during the refueling outage was scheduled for successive 12-hour shifts and was inappropriately excluded from the work hour limits specified in 10 CFR 26.205(d)(1) and 10 CFR 26.205(d)(2). The licensee removed the engineer from covered work activities for the remainder of the refueling outage and reviewed the work activities of other engineers to ensure that any engineer performing covered work appropriately met work hour limits.
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| The finding was of more than minor signficance since the failure to schedule and control the work hours of a worker performing covered work, if left uncorrected, would become a more significant safety concern because it could reasonably result in human performance errors that could affect the function of safety-related structures, systems, and components. Since the issue involved leak rate testing on containment isolation valves performed during the refueling outage, the inspectors concluded that this issue was associated with the Barrier Integrity Cornerstone. The finding was a licensee performance deficiency of very low safety significance because it did not represent an actual open pathway in the physical integrity of the reactor containment. The inspectors concluded that this finding affected the cross-cutting area of human performance. Specifically, the engineer did not meet expectations regarding the performance of
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| covered work activities because he did not challenge directions given to him by the leak rate test team supervisor and the leak rate test team supervisor did not meet expectations to ensure that the engineer was in compliance with the 10 CFR 26.205 (a) work requirements. Therefore, the inspectors concluded that the licensee's work practices which support human performance were less than effective. (IMC 0310, H.4(b))
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| Inspection Report# : 2011005 (pdf)
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| Significance: Dec 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation UNACCEPTABLE PRECONDITIONING OF REACTOR CORE ISOLATION COOLING SYSTEM CHECK VALVE PRIOR TO LEAK RATE TEST MEASUREMENT The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings." The licensee failed to establish an adequate procedure to perform required leak rate testing for the reactor core isolation cooling turbine exhaust check valve.
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| Specifically, the surveillance test procedure resulted in unacceptable preconditioning of the valve prior to an as-found leak rate test measurement. The licensee entered this issue into its corrective action program for evaluation and initiated a corrective action to revise the test procedure.
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| The finding was of more than minor significance since it was associated with the Procedure Quality Cornerstone attribute for the Containment and adversely affected the Barrier Integrity Cornerstone objective to provide rasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events.
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| Because the preconditioning altered the as-found condition of the check valve, the data collected through the performance of the surveillance test was not fully indicative of the true valve performance trend. Therefore, this performance deficiency had a direct effect on the licensee's ability to fully assess the past operability, as well as the ability to trend as-found data for the prupose of assessing the reliability of the check valve. The finding was a licensee performance deficiency of very low safety significance because it did not involve an actual open pathway in the physical integrity of the reactor containment. The inspectors concluded that this finding affected the cross-cutting area of problem identification and resolution. Specifically, the licensee did not implement operating experience into station processes, procedures, and training in that the licensee did not update/revise the surveillance test procedure consistent with NRC guidance and its corporate technical positon to prevent unacceptable preconditioning of the check valve.
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| (IMC 0310, P.2(b))
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| Inspection Report# : 2011005 (pdf)
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| Significance: Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO MEET TECHNICAL SPECIFICATION 3.7.3 FOR OPERABILITY OF CONTROL ROOM VENTILATION SYSTEM The inspectors identified a finding of very low safety significance with an associated non-cited violation of Technical Specification (TS) 3.7.3, "Control Room Ventilation System," following the discovery of a crack on the Train B Control Room ventilation (VC) system return fan hub during investigation of the cause for high noise and vibration levels observed on May 23, 2011. The licensee failed to correctly evaluate the operability of the Train B VC system return fan in a timely manner to prevent exceeding the TS allowed outage time for entry into Mode 3. The licensee replaced the fan and returned it to an operable status.
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| The failure to correctly evaluate a degraded/nonconforming condition potentially affecting the operability of structures, systems, and components (SSCs) required to be operable by TS would become a more significant safety concern if left uncorrected because it could reasonably result in an unrecognized condition of an SSC failing to fulfill a safety-related function. The finding was, therefore, of more than minor significance. Because the Control Room ventilation system supports the radiological barrier function to protect operators inside the Control Room following certain design basis accidents, the inspectors concluded that this issue was associated with the Barrier Integrity Cornerstone. The finding was a licensee performance deficiency of very low safety significance because it involved only a degradation of the radiological barrier function provided for the Control Room. The inspectors concluded that
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| this finding affected the cross-cutting area of human performance. Specifically, licensee decision making to delay inspection of the fan hub and blades until after a new fan was delivered on site to confirm the initial operability determination was not conservative and not consistent with demonstrating that nuclear safety is an overriding priority.
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| (IMC 0310, H.1(b))
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| Inspection Report# : 2011004 (pdf)
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| Significance: Sep 30, 2011 Identified By: NRC Item Type: FIN Finding FAILURE TO EVALUATE OPERABILITY OF CONTROL ROOM VENTILATION SYSTEM FOR DEGRADED FLOW CONDITION The inspectors identified a finding of very low safety significance. The licensee failed to appropriately evaluate the operability of Control Room Ventilation Train A after identifying a degraded/nonconforming system flow condition while performing surveillance testing on April 1, 2011, that could have affected the ability of the system to perform its safety function. No violation of regulatory requirements was identified. The licensee initiated corrective actions to provide "read & sign" training for licensed operators and a procedure change to add an acceptance criterion for filtered flow rate in the surveillance test procedure.
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| The failure to correctly evaluate a degraded/nonconforming condition potentially affecting the operability of structures, systems, and components (SSCs) required to be operable by Technical Specifications (TS) would become a more significant safety concern if left uncorrected because it could reasonably result in an unrecognized condition of an SSC failing to fulfill a safety-related function. The finding was therefore of more than minor significance. Because the Control Room ventilation system supports the radiological barrier function to protect operators inside the Control Room following certain design basis accidents, the inspectors concluded that this issue was associated with the Barrier Integrity Cornerstone. The finding was a licensee performance deficiency of very low safety significance because it involved only a degradation of the radiological barrier function provided for the Control Room. The inspectors concluded that this finding affected the cross-cutting area of human performance. Specifically, licensee decision making using a systematic process to evaluate the operability of an SSC required to be operable by TS when a degraded/nonconforming condition was identified was not appropriately implemented as designed by licensed senior reactor operators. (IMC 0310 H.1(a))
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| Inspection Report# : 2011004 (pdf)
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| Significance: Mar 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO MEET SURVEILLANCE TESTING REQUIREMENT FOR HYDROGEN IGNITERS IN ACCESSIBLE AREAS OF THE PRIMARY CONTAINMENT AND DRYWELL.
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| The inspectors identified a finding of very low safety significance with an associated non-cited violation of Technical Specification Surveillance Requirement (TSSR) 3.6.3.2.4. The licensee failed to verify that each required hydrogen igniter in accessible areas of the Primary Containment and Drywell develops a surface temperature of greater than or equal to 1700 degrees Fahrenheit (°F) every 24 months. The licensee performed a risk assessment of the missed surveillance in accordance with TSSR 3.0.3, which determined that completion of the surveillance could be delayed up to the 24 month surveillance interval without a significant increase in plant risk. The licensee also completed an operability evaluation for the TS nonconformance and concluded that there was reasonable assurance that the affected hydrogen igniters were operable based on the results of surveillance testing to measure voltage/current draw.
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| The finding was of more than minor significance because it was associated with the Human Performance attribute for the Containment and adversely affected the Barrier Integrity Cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the licensee did not correctly evaluate a change to perform the surveillance test with the unit at power beginning in March 2002. It was not recognized that TSSR 3.6.3.2.4 would not be met for accessible hydrogen igniters in the Drywell and 755 Elevation Steam Tunnel when performing the test with the unit at power and the licensee incorrectly believed that performance of the current/voltage surveillance test procedure for inaccessible igniters was an appropriate substitute, contrary to existing procedural guidance. The finding was a licensee performance deficiency of very low safety significance because it did not involve an actual reduction in the function of hydrogen igniters in the
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| Primary Containment and Drywell. The inspectors concluded that because the scheduling change to perform the surveillance with the unit at power took place prior to surveillance testing beginning in March 2002, it did not necessarily reflect current licensee performance and no cross-cutting aspect was identified.
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| Inspection Report# : 2011002 (pdf)
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| Significance: Mar 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE TESTING CONTROLS TO PERFORM SURVEILLANCE TESTING OF HYDROGEN IGNITERS IN THE PRIMARY CONTAINMENT AND DRYWELL.
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| The inspectors identified a finding of very low safety significance with an associated non-cited violation of 10 CFR 50, Appendix B, Criterion XI, Test Control. The licensee failed to establish a test program adequate to assure testing of hydrogen igniters in accessible areas of the Primary Containment and Drywell pursuant to TSSR 3.6.3.2.4. The licensee entered this violation into its corrective action program to investigate the cause and to identify appropriate corrective actions.
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| The finding was of more than minor significance because it was associated with the Procedure Quality attribute for the Containment and adversely affected the Barrier Integrity Cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The finding was a licensee performance deficiency of very low safety significance because it did not involve an actual reduction in the function of hydrogen igniters in the Primary Containment and Drywell. The inspectors concluded that this finding affected the cross-cutting aspect of human performance. Specifically, adequate licensee resources involving personnel and procedures did not support successful human performance. CPS 9867.05 was not appropriate to the circumstances because it contained errors and did not provide adequate testing controls for the performance of the surveillance test (IMC 0310 H.2(c)).
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| Inspection Report# : 2011002 (pdf)
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| Emergency Preparedness Significance: Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation MISSING RESPIRATOR SPECTACLE KITS The inspectors identified a finding of very low safety significance and associated NCV of 10 CFR 50.54(q) for the failure to provide spectacle adapter kits for all eyeglass wearers (i.e, non-soft contact wearers) who were key emergency response organization (ERO) personnel that were potentially required to wear a self-contained breathing apparatus (SCBA) in order to fulfill emergency response functions. The licensee's corrective actions included revising procedures that govern the training and qualification of licensed operators to include steps that ensure licensed operators and other ERO members who require corrective lenses are provided SCBA lens inserts.
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| The finding was more than minor because it was associated with the Emergency Planning Cornerstone and, if left uncorrected, the performance deficiency has the potential to lead to a more significant safety concern, in that, emergency responders having inadequate vision could challenge the licensee's state of operational readiness and emergency response capabilities. The finding was assessed using IMC 0609, Attachment B , "Emergency Preparedness Significance Determination Process" and determined to be of very low safety significance because this failure to comply represented a planning standard issue, however, it did not result in a risk significant planning standard nor was it indicative of a planning standard functional failure. The failure to make provisions for respirator vision corrective lenses to licensed operators that required corrective lenses as a condition of their license was caused by a program weakness. Consequently, the cause of this finding has a cross-cutting aspect in the area of human performance. Specifically, the licensee did not ensure that equipment was available for key emergency response personnel. (IMC 0310,H.2(d))
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| Inspection Report# : 2011004 (pdf)
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| Significance: SL-IV Jun 22, 2011 Identified By: NRC Item Type: NCV NonCited Violation CHANGES TO EAL BASIS DECREASED THE EFFECTIVENESS OF THE PLAN WITHOUT PRIOR NRC APPROVAL (TRADITIONAL ENFORCEMENT PORTION)
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| The inspector identified a finding of very low safety significance involving a Severity Level IV NCV of 10 CFR 50.54 (q) for failing to obtain prior approval for an emergency plan change which decreased the effectiveness of the plan.
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| Specifically, the licensee modified the Emergency Action Level (EAL) Basis in EAL HU6, Revision 12, which indefinitely extended the start of the 15 minute emergency classification clock beyond a credible notification that a fire is occurring or indication of a valid fire detection system alarm. This change decreased the effectiveness of the emergency plan by reducing the capability to perform a risk significant planning function in a timely manner.
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| The violation affected the NRCs ability to perform its regulatory function because it involved implementing a change that decreased the effectiveness of the emergency plan without NRC approval. Therefore, this issue was evaluated using Traditional Enforcement. The NRC determined that a Severity Level IV violation was appropriate due to the reduction of the capability to perform a risk significant planning standard function in a timely manner. The licensee entered this issue into its corrective action program and revised the EAL basis to restore compliance.
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| The related performance deficiency is tracked as item 2010-502-02.
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| Inspection Report# : 2010502 (pdf)
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| Significance: Jun 22, 2011 Identified By: NRC Item Type: NCV NonCited Violation CHANGES TO EAL BASIS DECREASED THE EFFECTIVENESS OF THE PLAN WITHOUT PRIOR NRC APPROVAL (PERFORMANCE DEFICIENCY PORTION)
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| The inspector identified a finding of very low safety significance involving a Severity Level IV NCV of 10 CFR 50.54 (q) for failing to obtain prior approval for an emergency plan change which decreased the effectiveness of the plan.
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| Specifically, the licensee modified the Emergency Action Level (EAL) Basis in EAL HU6, Revision 12, which indefinitely extended the start of the 15 minute emergency classification clock beyond a credible notification that a fire is occurring or indication of a valid fire detection system alarm. This change decreased the effectiveness of the emergency plan by reducing the capability to perform a risk significant planning function in a timely manner.
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| The finding was more than minor using IMC 0612, because it is associated with the emergency preparedness cornerstone attribute of procedure quality for EAL and emergency plan changes, and it adversely affected the cornerstone objective of ensuring that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Therefore, the performance deficiency was a finding.
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| Using IMC 0609, Appendix B, the inspector determined that the finding had a very low safety significance because the finding is a failure to comply with 10 CFR 50.54(q) involving the risk significant planning standard 50.47(b)(4),
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| which, in this case, met the example of a Green finding because it involved one Unusual Event classification (EAL HU6).
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| The associated traditional enforcement is tracked as item 2010-502-01.
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| Inspection Report# : 2010502 (pdf)
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| Occupational Radiation Safety Significance: Dec 31, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation FAILURE TO IMPLEMENT APPROPRIATE RADIOLOGICAL CONTROLS FOR THE REMOVAL OF
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| INSULATION IN A POSTED HIGH CONTAMINATION AREA A self-revealed finding of very low safety significance and an associated Non-Cited Violation of Technical Specification 5.4.1.a was identified. Specifically, the licensee failed to implement appropriate radiological controls for the removal of insulation in a posted high contamination area. The issue was entered in the licensee's corrective action program as AR 01297713. The licensee's immediate corrective actions placed the job on hold, assessed the radiological significance for the issue, and suspended qualifications for the radiation protection technician (RPT) involved.
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| The finding is more than minor because the performance deficiency is associated with the Program and Process attribute of the Occupational Radiation Safety Cornerstone and adversely affected the cornerstone objective to ensure the adequate protection of the worker's health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Specifically, the failure to implement the radiological controls established in the radiation worker permit (RWP) as-low-reasonably-achievable (ALARA) file caused workers to receive additional, unplanned dose to the workers. The finding was assessed using the Occupational Radiation Safety, Public Radiation Safety and was determined to be of very-low safety significance because this was not related to ALARA, did not result in an overexposure, or a substantial potential for overexposure, nor was the ability to assess dose compromised.
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| The radiological controls specified in RWP 10012059 for this activity were not implemented because the RPT assumed the scope of work and failed to review the RWP ALARA requirements before the briefing. Consequently, the inspectors determined that the cause of this incident involved a cross-cutting component in the human performance area for work practices. Specifically personnel work practices did not support human performance. (IMC 0310, H.4(a))
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| Inspection Report# : 2011005 (pdf)
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| Significance: Dec 31, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation FAILURE TO IMPLEMENT APPROPRIATE RADIOLOGICAL CONTROLS AFTER RADIATION PROTECTION IDENTIFIED THAT A WORKER WAS POTENTIALLY CONTAMINATED DUE TO INAPPROPRIATE PROTECTIVE CLOTHING.
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| A self-revealed finding of very low safety significance and an associated Non-Cited Violation of Technical Specification 5.4.1.a was identified. Specifically, the licensee failed to implement appropriate radiological controls after radiation protection identified that the worker was potentially contaminated due to the inappropriate protective clothing. This issue was entered in the licensee's corrective action program as AR 01017724. The licensee's corrective actions included the replacement of all contamination monitors used at the site. The new contamination monitors have a radon subtract feature designed to mitigate the large number nuisance alarms caused by radon interference at this site.
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| The finding is more than minor because if left uncorrected the performance deficiency would have the potential to lead to a more significant safety concern. Specifically, bypassing every level of defense could result in additional dose to worker outside the radiological control area. The finding was assessed using the Significance Determination Process and was determined to be of very-low safety significance because these radioactive material control issues were not related to transportation and dose to members of the public was less than 0.005 rem. The inspectors observed the operation of the new contamination monitors and response of radiation protection technicians assigned to monitor authorized exit points during a refueling outage. The new monitors did not exhibit nuisance alarms and the technicians treated every alarm as a potential contamination event until proven otherwise with another instrument. Futhermore, these technicans informed the inspectors the briefing received before the outage by the radiation protection manager about alarm response expectations. The inspectors determined that the events involved in this performance deficiency are not indicative of current performance. Consequently, the inspectors did not assess the performance deficiency for cross-cutting aspects.
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| Inspection Report# : 2011005 (pdf)
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| Public Radiation Safety
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| Significance: Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IMPLEMENT PACKAGE DESIGN SPECIFICATIONS The inspectors identified a finding of very low-safety-significance and an associated non-cited violation of 10 CFR 71.5 for the failure to implement package design specifications. Specifically, the licensee failed to ensure the proper closure of a DOT 7A Type A package as required by Department of Transportation (DOT) regulations for packaging contained within 49 CFR 173. As a part of their corrective actions, the licensee completed a detailed review of all radioactive material shipments for the past 36 months to ensure Package Certification documents for other packages used as a Type A container satisfied requirements.
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| The finding was more than minor because it affected the Public Radiation Safety Cornerstone objective to ensure adequate protection of public health and safety from exposure to radioactive materials during transit. Specifically, the failure to correctly close a DOT Type A package could lead to a more significant safety concern by increasing the potential for a package breach occuring during transit. Using IMC 0609, Attachment D for the Public Radiation Safety Significance Determination Process (SDP) the inspectors determined the finding to be of very low safety significance.
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| This deficiency has a cross-cutting aspect in Human Performance (Resources). (IMC 0310 H.2(b))
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| Inspection Report# : 2011004 (pdf)
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| Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.
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| Miscellaneous Last modified : March 02, 2012
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| Clinton 1Q/2012 Plant Inspection Findings Initiating Events Significance: Mar 31, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation FAILURE TO INCORPORATE OPERATING EXPERIENCE INTO PREVENTIVE MAINTENANCE ACTIVITIES.
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| A self-revealed finding of very low safety significance was identified with an associated Non-Cited Violation of 10 CFR 50.65, "Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants." The licensee failed to incorporate operating experience into its preventive maintenance practices associated with steam bypass system control circuit cards. Specifically, during two operating experience driven initiatives performed by the licensee in 2001 and 2007, and once again on September 24, 2011, the licnesee failed to implement any preventive maintenance activity for critical component circuit cards, which resulted in age-related failure and a reactor scram on November 29, 2011. The licensee initiated corrective actions to replace system circuit cards, perform periodic replacement/refurbishment maintenance activities, and trend circuit card performance during routine calibration.
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| The finding was of more than minor significance because it was sufficiently similar to Inspection Manual Chapter 0612, "Power Reactor Inspection Reports," Appendix E, "Examples of Minor Issues," Example 7 (c), in that this violation of 10 CFR 50.65(a)(3) had a consequence "...such as equipment problems attributable to failure to take industry operating experience into account when practicable." The finding was a licensee performance deficiency of very low safety significance because it (1) did not contribute to the likelihood of a loss-of-collant accident initiator, (2) did not contribute to both the likelihood of a reactor scram AND the likelihood that mitigation equipment or functions would not be available, and (3) did not increase the likelihood of a fire or internal/external flooding event.
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| The inspectors concluded that this finding affected the cross-cutting area of human performance. Specifically, in the area of work control, the licensee did not appropriately coordinate work activities by incorporating actions to plan work activities to support long-term equipment reliability by scheduling maintenance as more preventive than reactive. (IMC 0310, H.3(b))
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| Inspection Report# : 2012002 (pdf)
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| Significance: Jun 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO MEET SURVEILLANCE TESTING REQUIREMENT FOR REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES The inspectors identified a finding of very low safety significance (Green) with an associated Non Cited Violation of Technical Specification Surveillance Requirement (TSSR) 3.4.6.1. The licensee failed to correctly incorporate the required test pressure limits of the TSSR into the surveillance test procedure and subsequently tested multiple reactor coolant system (RCS) pressure isolation valves (PIVs) at pressures greater than the maximum test pressure of 1025 pounds-per-square-inch gage, invalidating the testing. The licensee performed a risk assessment of the missed surveillance in accordance with TSSR 3.0.3, which determined that completion of the surveillance could be delayed up to the 24 month surveillance interval without a significant increase in plant risk. The licensee also completed an operability evaluation for the TS nonconformance and concluded that there was reasonable assurance that the affected RCS PIVs were operable based on engineering judgment.
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| The finding was of more than minor significance because it affected the Initiating Events Cornerstone and was associated with the Procedure Quality attribute. Specifically, the licensee did not correctly incorporate the required test pressure limits of TSSR 3.4.6.1 into the surveillance test procedure. This resulted in testing multiple RCS PIVs at pressures greater than the maximum test pressure of 1025 psig. The finding was determined to be a licensee performance deficiency of very low safety significance because the finding would not result in exceeding the TS limit
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| for RCS leakage and would not have likely affected mitigation systems resulting in a loss of safety function. The inspectors concluded that because the licensees missed opportunity to correct the test pressure discrepancy in its surveillance test procedure occurred in January 2005 and no other more recent opportunities reasonably existed to identify and correct the problem, this issue would not be reflective of current licensee performance and no cross-cutting aspect was identified.
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| Inspection Report# : 2011003 (pdf)
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| Significance: Jun 03, 2011 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PERFORM EFFECTIVENESS REVIEW.
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| The inspectors identified a finding of very low safety significance with an associated NCV of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings." The licensee failed to perform an effectiveness review (EFR) to ensure that corrective actions (CAs) taken to prevent recurrence of a significant condition adverse to quality were actually effective to preclude repetition. The licensee entered this violation into its corrective action program as ARs 1221616, 1221661, and 1223806 to investigate the cause and to identify appropriate CAs.
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| The finding was of more than minor significance because it was similar to Example 4a in IMC 0612, "Power Inspection Reports," Appendix E, "Examples of Minor Issues," in that, the licensee routinely failed to perform EFR evaluations on similar CAs related to significant conditions adverse to quality. The finding was a licensee performance deficiency of very low safety significance due to answering 'no' to all questions under the Initiating Events Cornerstone column of IMC 0609 Attachment 4, "Phase 1 - Initial Screening and Characterization of Findings," The inspectors concluded that this finding affected the cross-cutting aspect of problem identification and resolution. Specifically, the licensee failed to thoroughly evaluate problems to include conducting EFRs of CAs to ensure that problems were resolved. [IMC 0310 P.1(c)]
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| Inspection Report# : 2011008 (pdf)
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| Mitigating Systems Significance: Dec 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PROPERLY APPLY AN APPROVED ASME CODE CASE The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50.55a due to the licensee's failure to adequately apply American Society of Mechanical Engineers Section XI Code Case N-513-3 when it evaluated a degraded section of safety related shutdown service water system piping for operability. Specifically, the licensee failed to perform required daily walkdowns to confirm its analysis of conditions used in its operability evaluation remained valid. After this issue was identified by the inspectors, the licensee promptly resumed the daily compensatory action to verify the leak rate until the piping system was repaired.
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| The finding was of more than minor significance because it was associated with the Protection Against External Factors attribute of the Mitigating Systems Cornerstone, and it directly affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Additionally, if left uncorrected, improper application of an approved code case would become a more significant safety concern in that it could result in the failure to identify inoperable safety related piping. The finding was a licensee performance deficiency of very low safety significance because it did not result in an actual loss of safety function of a single train for greater than its Technical Specification allowed outage time. The inspectors concluded that there was no specific performance characteristic that was a significant cause to the performance deficiency in this instance; therefore no cross-cutting aspect was identified.
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| Inspection Report# : 2011005 (pdf)
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| Significance: Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PERFORM CODE REQUIRED CAUSE AND EFFECT FAILURE EVALUATIONS FOR DIESEL STARTING AIR AND FUEL OIL SYSTEM RELIEF VALVES.
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| The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50.55a. The licensee failed to perform American Society of Mechanical Engineers (ASME) Code required cause and effect failure evaluations for set pressure test failures of diesel generator (DG) starting air and fuel oil system relief valves. The licensee entered this issue into its corrective action program for evaluation and subsequently completed an engineering evaluation to address past operability of the associated DG starting air and fuel oil systems due to the relief valve test failures. The licensee also moved up its schedule to test the remaining relief valves.
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| The finding was of more than minor significance because it could lead to a more significant safety concern if left uncorrected. Specifically, the failure to perform Code required cause and effect evaluations for relief valve set pressure test failures could lead to a generic problem with valves in the same or other valve groups remaining uncorrected with a potential impact on operability of safety significant mitigating systems. Because the DG starting air and fuel oil systems are relied upon to support DG operability, the inspectors concluded that this issue was associated with the Mitigating Systems Cornerstone. The finding was determined to be a licensee performance deficiency of very low safety significance because the finding: (1) was not a design or qualification deficiency; (2) did not represent an actual loss of safety function of a system; (3) did not represent an actual loss of safety function of a single train or greater than its Technical Specification (TS) allowed outage time; (4) did not represent an actual loss of safety function of one or more non-TS trains of equipment designated as risk significant; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors concluded that the finding affected the cross-cutting area of human performance in that the licensee's work practices did not ensure adequate supervisory and management oversight of work activities, such that nuclear safety was supported.
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| Specifically, the relief valve test failures were left unresolved and were not evaluated as required by the Code for an extended period of time with several failed tests. (IMC 0310, H.4(C))
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| Inspection Report# : 2011004 (pdf)
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| Significance: Sep 30, 2011 Identified By: NRC Item Type: FIN Finding FAILURE TO CORRECT A CONDITION ADVERSE TO QUALITY FOR IMPROPERLY IMPLEMENTED ENGINEERING CORRECTIVE ACTIONS The inspectors identified a finding of very low safety significance due to the licensee's failure to effectively implement corrective actions for a condition adverse to quality described in Apparent Cause Evaluation 1095413, "NOS [Nuclear Oversight] Identified Improperly Implemented Engineering Corrective Actions Cause Repeat Operational Challenges." No violation of regulatory requirements was identified. The licensee entered this issue into its corrective action program to investigate the cause and to identify appropriate corrective actions.
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| The finding was of more than minor significance because it was associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone and directly affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.
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| Specifically, improperly implemented engineering corrective actions could result in additional repeat operational equipment challenges. The finding was of very low safety significance because the issue: (1) was not a design or qualification deficiency; (2) did not represent an actual loss of safety function of a system; (3) did not represent an actual loss of safety function of a single train for greater than its Technical Specification (TS) allowed outage time; (4) did not represent an actual loss of safety function of one or more non-TS trains of equipment designated as risk significant; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors concluded that this finding affected the cross-cutting area of problem identification and resolution. Specifically, the licensee failed to take appropriate corrective actions to address known deficiencies in its process for tracking and closing work orders that implement corrective actions. The actions taken were neither lasting nor effective. (IMC 0310, P.1(d))
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| Inspection Report# : 2011004 (pdf)
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| Significance: Jun 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation DEFICIENCIES WITH RCIC ROOM HEAT-UP ANALYSES The inspectors identified a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, having very low safety significance for the failure to include all of the applicable heat loads in the reactor core isolation cooling (RCIC) room heat up calculation and not having a calculation of record for the RCIC room heat up under a station blackout (SBO) scenario. The licensee entered this issue into the corrective action program and performed preliminary calculations to verify that the issues did not exceed any design limits.
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| The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of Equipment Performance, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding screened as very low safety significance because the licensee determined the RCIC room cooler was capable of removing the additional heat load; and RCIC room temperature remained within the design limits without the room cooler during a SBO scenario. The inspectors determined that this finding did not represent current licensee performance and no cross-cutting aspect was assigned.
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| Inspection Report# : 2011003 (pdf)
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| Significance: Jun 03, 2011 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO MAINTAIN A QUALITY RECORD AS EVIDENCE OF AN ACTIVITY AFFECTING QUALITY OF SAFETY-RELATED EQUIPMENT DUE TO INAPPROPRIATE CORRECTIVE ACTIONS The inspectors identified a finding of very low safety significance with an associated NCV of 10 CFR Part 50, Appendix B, Criterion XVII, Quality Assurance Records. Specifically, the licensee failed to maintain a quality record documenting a nondestructive examination (NDE) of a safety-related spreader beam lifting device. After losing the original NDE report, the licensees corrective action (CA) was to recreate the report from memory and maintain the recreated report as the quality record. Upon review and questioning from the NRC, the licensee was able to locate the missing NDE report in the records archive. This issue was entered into the licensees CAP as AR1223723.
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| The inspectors determined the finding was more than minor because, if left uncorrected, failure to maintain a quality record as evidence of an activity affecting quality of safety related equipment due to inappropriate disposition of CAs pertaining to missing/lost quality records could become a more significant safety concern. This finding was of very low safety significance because this finding did not represent an actual loss of any safety function of the Mitigation Systems. The inspectors concluded that this finding affected the cross-cutting aspect of Problem Identification and Resolution. Specifically, the licensee did not take appropriate corrective actions to address a lost quality record. [IMC 0310 P.1(d)]
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| Inspection Report# : 2011008 (pdf)
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| Significance: Jun 03, 2011 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ACCOUNT FOR CABLE RESISTANCE IN OPERABILITY DETERMINATIONS.
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| The inspectors identified a finding of very low safety significance with an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," related to calculational errors found in the licensee's operability determination. Specifically, on four separate operability determinations, the licensee failed to account for the cable resistance when determining the maximum allowable contact resistance associated with the second level undervoltage (UV) relays for the 4.16 kV Buses. The licensee entered this violation into its corrective action program as Action Requests 1226340 and 1224313 and performed a preliminary calculation which determined that the error reduced the available margin in the circuit resistance but did not change the overall conclusions for the past operability calls made for the four different occasions.
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| The inspectors determined that this finding was more than minor because it was associated with the Mitigating Systems Cornerstone attribute of design control and adversely affected the cornerstone objective of ensuring availability and reliability of systems that respond to initiating events to prevent undesirable consquences. This finding was of very low safety significance (Green) because the licensee was able to demonstrate that the operability calls that were previously made relating to the second level UV relays were still valid and acceptable. The inspectors condcluded that this finding affected the cross-cutting aspect of human performance. Specifically, the licensee failed to use conservative assumptions in decision making related to immediate operability determinations of conditions adverse to quality. [IMC 0310 H.1(b)]
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| Inspection Report# : 2011008 (pdf)
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| Barrier Integrity Significance: Mar 31, 2012 Identified By: NRC Item Type: NCV NonCited Violation UNACCEPTABLE PRECONDITIONING OF LOW PRESSURE COOLANT INJECTION FROM RESIDUAL HEAT REMOVAL 'A' CHECK VALVE PRIOR TO LEAK RATE TEST MEASUREMENT The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings." The licensee failed to establish an adequate procedure to perform required leak rate testing for the Low Pressure Coolant Injection from Residual Heat Removal
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| 'A' Check Valve. Specifically, the surveillance test procedure resulted in unacceptable preconditioning of the valve prior to a leak rate test measurement due to improper test sequencing. In addition, the licensee failed to correctly evaluate a failed leak rate test of the valve. The licensee entered this issue into its corrective action program for evaluation and initiated corrective actions to revise the test procedure and train engineering personnel.
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| The finding was of more than minor significance since it was associated with the Procedure Quality attribute for the containment and adversely affected the Barrier Integrity Cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Because the preconditioning altered the as-found condition of the check valve, the data collected through the performance of the surveillance test was not fully indicative of the true valve performance trend. Additionally, the licensee's failure to correctly evaluate the initial failed leak rate test would become a more significant safety concern if left uncorrected because it could reasonably result in an unrecognized condition with a check valve failing to fulfill a safety related function. Therefore, this performance deficiency had a direct effect on the licensee's ability to fully assess the past operability, as well as the ability to trend as-found data for the purpose of assessing the reliability of the check valve.
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| The finding was a licensee performance deficiency of very low safety significance because it would not result in exceeding the Technical Specification limit for reactor coolant system leakage and would not have likely affected mitigation systems resulting in a loss of safety function. In addition, the finding did not represent an actual open pathway in the physical integrity of the reactor containment. Based on consultation and review with the Regional Senior Reactor Analyst, the inspectors concluded that the finding did not result in an increase in the likelihood of an initiating event such as an inter-system loss-of-coolant accident or a containment bypass event because the redundant isolation valve and closed loop system piping passed leak rate measurement test during the refueling outage with considerable margin. The inspectors concluded that this finding affected the cross cutting area of human performance.
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| Specifically, the licensee did not have adequately trained and knowledgeable personnel available to correctly evaluate the cause of the initial failed leak rate measurement test and to ensure that appropriate actions to correct the test sequence in the procedure were identified.
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| (IMC 0310,H.2(b))
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| Inspection Report# : 2012002 (pdf)
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| Significance: Dec 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO CONTROL THE WORK HOURS OF A COVERED WORKER.
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| The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 26.205(c) and (d) for the licensee's failure to schedule and control the work hours of a covered worker performing surveillance testing on containment isolation valves during the refueling outage. Specifically, an engineer performing local leak rate testing during the refueling outage was scheduled for successive 12-hour shifts and was inappropriately excluded from the work hour limits specified in 10 CFR 26.205(d)(1) and 10 CFR 26.205(d)(2). The licensee removed the engineer from covered work activities for the remainder of the refueling outage and reviewed the work activities of other engineers to ensure that any engineer performing covered work appropriately met work hour limits.
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| The finding was of more than minor signficance since the failure to schedule and control the work hours of a worker performing covered work, if left uncorrected, would become a more significant safety concern because it could reasonably result in human performance errors that could affect the function of safety-related structures, systems, and components. Since the issue involved leak rate testing on containment isolation valves performed during the refueling outage, the inspectors concluded that this issue was associated with the Barrier Integrity Cornerstone. The finding was a licensee performance deficiency of very low safety significance because it did not represent an actual open pathway in the physical integrity of the reactor containment. The inspectors concluded that this finding affected the cross-cutting area of human performance. Specifically, the engineer did not meet expectations regarding the performance of covered work activities because he did not challenge directions given to him by the leak rate test team supervisor and the leak rate test team supervisor did not meet expectations to ensure that the engineer was in compliance with the 10 CFR 26.205 (a) work requirements. Therefore, the inspectors concluded that the licensee's work practices which support human performance were less than effective. (IMC 0310, H.4(b))
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| Inspection Report# : 2011005 (pdf)
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| Significance: Dec 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation UNACCEPTABLE PRECONDITIONING OF REACTOR CORE ISOLATION COOLING SYSTEM CHECK VALVE PRIOR TO LEAK RATE TEST MEASUREMENT The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings." The licensee failed to establish an adequate procedure to perform required leak rate testing for the reactor core isolation cooling turbine exhaust check valve.
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| Specifically, the surveillance test procedure resulted in unacceptable preconditioning of the valve prior to an as-found leak rate test measurement. The licensee entered this issue into its corrective action program for evaluation and initiated a corrective action to revise the test procedure.
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| The finding was of more than minor significance since it was associated with the Procedure Quality Cornerstone attribute for the Containment and adversely affected the Barrier Integrity Cornerstone objective to provide rasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events.
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| Because the preconditioning altered the as-found condition of the check valve, the data collected through the performance of the surveillance test was not fully indicative of the true valve performance trend. Therefore, this performance deficiency had a direct effect on the licensee's ability to fully assess the past operability, as well as the ability to trend as-found data for the prupose of assessing the reliability of the check valve. The finding was a licensee performance deficiency of very low safety significance because it did not involve an actual open pathway in the physical integrity of the reactor containment. The inspectors concluded that this finding affected the cross-cutting area of problem identification and resolution. Specifically, the licensee did not implement operating experience into station processes, procedures, and training in that the licensee did not update/revise the surveillance test procedure consistent with NRC guidance and its corporate technical positon to prevent unacceptable preconditioning of the check valve.
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| (IMC 0310, P.2(b))
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| Inspection Report# : 2011005 (pdf)
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| Significance: Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO MEET TECHNICAL SPECIFICATION 3.7.3 FOR OPERABILITY OF CONTROL ROOM
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| VENTILATION SYSTEM The inspectors identified a finding of very low safety significance with an associated non-cited violation of Technical Specification (TS) 3.7.3, "Control Room Ventilation System," following the discovery of a crack on the Train B Control Room ventilation (VC) system return fan hub during investigation of the cause for high noise and vibration levels observed on May 23, 2011. The licensee failed to correctly evaluate the operability of the Train B VC system return fan in a timely manner to prevent exceeding the TS allowed outage time for entry into Mode 3. The licensee replaced the fan and returned it to an operable status.
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| The failure to correctly evaluate a degraded/nonconforming condition potentially affecting the operability of structures, systems, and components (SSCs) required to be operable by TS would become a more significant safety concern if left uncorrected because it could reasonably result in an unrecognized condition of an SSC failing to fulfill a safety-related function. The finding was, therefore, of more than minor significance. Because the Control Room ventilation system supports the radiological barrier function to protect operators inside the Control Room following certain design basis accidents, the inspectors concluded that this issue was associated with the Barrier Integrity Cornerstone. The finding was a licensee performance deficiency of very low safety significance because it involved only a degradation of the radiological barrier function provided for the Control Room. The inspectors concluded that this finding affected the cross-cutting area of human performance. Specifically, licensee decision making to delay inspection of the fan hub and blades until after a new fan was delivered on site to confirm the initial operability determination was not conservative and not consistent with demonstrating that nuclear safety is an overriding priority.
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| (IMC 0310, H.1(b))
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| Inspection Report# : 2011004 (pdf)
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| Significance: Sep 30, 2011 Identified By: NRC Item Type: FIN Finding FAILURE TO EVALUATE OPERABILITY OF CONTROL ROOM VENTILATION SYSTEM FOR DEGRADED FLOW CONDITION The inspectors identified a finding of very low safety significance. The licensee failed to appropriately evaluate the operability of Control Room Ventilation Train A after identifying a degraded/nonconforming system flow condition while performing surveillance testing on April 1, 2011, that could have affected the ability of the system to perform its safety function. No violation of regulatory requirements was identified. The licensee initiated corrective actions to provide "read & sign" training for licensed operators and a procedure change to add an acceptance criterion for filtered flow rate in the surveillance test procedure.
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| The failure to correctly evaluate a degraded/nonconforming condition potentially affecting the operability of structures, systems, and components (SSCs) required to be operable by Technical Specifications (TS) would become a more significant safety concern if left uncorrected because it could reasonably result in an unrecognized condition of an SSC failing to fulfill a safety-related function. The finding was therefore of more than minor significance. Because the Control Room ventilation system supports the radiological barrier function to protect operators inside the Control Room following certain design basis accidents, the inspectors concluded that this issue was associated with the Barrier Integrity Cornerstone. The finding was a licensee performance deficiency of very low safety significance because it involved only a degradation of the radiological barrier function provided for the Control Room. The inspectors concluded that this finding affected the cross-cutting area of human performance. Specifically, licensee decision making using a systematic process to evaluate the operability of an SSC required to be operable by TS when a degraded/nonconforming condition was identified was not appropriately implemented as designed by licensed senior reactor operators. (IMC 0310 H.1(a))
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| Inspection Report# : 2011004 (pdf)
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| Emergency Preparedness Significance: Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation
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| MISSING RESPIRATOR SPECTACLE KITS The inspectors identified a finding of very low safety significance and associated NCV of 10 CFR 50.54(q) for the failure to provide spectacle adapter kits for all eyeglass wearers (i.e, non-soft contact wearers) who were key emergency response organization (ERO) personnel that were potentially required to wear a self-contained breathing apparatus (SCBA) in order to fulfill emergency response functions. The licensee's corrective actions included revising procedures that govern the training and qualification of licensed operators to include steps that ensure licensed operators and other ERO members who require corrective lenses are provided SCBA lens inserts.
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| The finding was more than minor because it was associated with the Emergency Planning Cornerstone and, if left uncorrected, the performance deficiency has the potential to lead to a more significant safety concern, in that, emergency responders having inadequate vision could challenge the licensee's state of operational readiness and emergency response capabilities. The finding was assessed using IMC 0609, Attachment B , "Emergency Preparedness Significance Determination Process" and determined to be of very low safety significance because this failure to comply represented a planning standard issue, however, it did not result in a risk significant planning standard nor was it indicative of a planning standard functional failure. The failure to make provisions for respirator vision corrective lenses to licensed operators that required corrective lenses as a condition of their license was caused by a program weakness. Consequently, the cause of this finding has a cross-cutting aspect in the area of human performance. Specifically, the licensee did not ensure that equipment was available for key emergency response personnel. (IMC 0310,H.2(d))
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| Inspection Report# : 2011004 (pdf)
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| Significance: SL-IV Jun 22, 2011 Identified By: NRC Item Type: NCV NonCited Violation CHANGES TO EAL BASIS DECREASED THE EFFECTIVENESS OF THE PLAN WITHOUT PRIOR NRC APPROVAL (TRADITIONAL ENFORCEMENT PORTION)
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| The inspector identified a finding of very low safety significance involving a Severity Level IV NCV of 10 CFR 50.54 (q) for failing to obtain prior approval for an emergency plan change which decreased the effectiveness of the plan.
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| Specifically, the licensee modified the Emergency Action Level (EAL) Basis in EAL HU6, Revision 12, which indefinitely extended the start of the 15 minute emergency classification clock beyond a credible notification that a fire is occurring or indication of a valid fire detection system alarm. This change decreased the effectiveness of the emergency plan by reducing the capability to perform a risk significant planning function in a timely manner.
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| The violation affected the NRCs ability to perform its regulatory function because it involved implementing a change that decreased the effectiveness of the emergency plan without NRC approval. Therefore, this issue was evaluated using Traditional Enforcement. The NRC determined that a Severity Level IV violation was appropriate due to the reduction of the capability to perform a risk significant planning standard function in a timely manner. The licensee entered this issue into its corrective action program and revised the EAL basis to restore compliance.
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| The related performance deficiency is tracked as item 2010-502-02.
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| Inspection Report# : 2010502 (pdf)
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| Significance: Jun 22, 2011 Identified By: NRC Item Type: NCV NonCited Violation CHANGES TO EAL BASIS DECREASED THE EFFECTIVENESS OF THE PLAN WITHOUT PRIOR NRC APPROVAL (PERFORMANCE DEFICIENCY PORTION)
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| The inspector identified a finding of very low safety significance involving a Severity Level IV NCV of 10 CFR 50.54 (q) for failing to obtain prior approval for an emergency plan change which decreased the effectiveness of the plan.
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| Specifically, the licensee modified the Emergency Action Level (EAL) Basis in EAL HU6, Revision 12, which indefinitely extended the start of the 15 minute emergency classification clock beyond a credible notification that a fire is occurring or indication of a valid fire detection system alarm. This change decreased the effectiveness of the emergency plan by reducing the capability to perform a risk significant planning function in a timely manner.
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| The finding was more than minor using IMC 0612, because it is associated with the emergency preparedness cornerstone attribute of procedure quality for EAL and emergency plan changes, and it adversely affected the
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| cornerstone objective of ensuring that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Therefore, the performance deficiency was a finding.
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| Using IMC 0609, Appendix B, the inspector determined that the finding had a very low safety significance because the finding is a failure to comply with 10 CFR 50.54(q) involving the risk significant planning standard 50.47(b)(4),
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| which, in this case, met the example of a Green finding because it involved one Unusual Event classification (EAL HU6).
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| The associated traditional enforcement is tracked as item 2010-502-01.
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| Inspection Report# : 2010502 (pdf)
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| Occupational Radiation Safety Significance: Dec 31, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation FAILURE TO IMPLEMENT APPROPRIATE RADIOLOGICAL CONTROLS FOR THE REMOVAL OF INSULATION IN A POSTED HIGH CONTAMINATION AREA A self-revealed finding of very low safety significance and an associated Non-Cited Violation of Technical Specification 5.4.1.a was identified. Specifically, the licensee failed to implement appropriate radiological controls for the removal of insulation in a posted high contamination area. The issue was entered in the licensee's corrective action program as AR 01297713. The licensee's immediate corrective actions placed the job on hold, assessed the radiological significance for the issue, and suspended qualifications for the radiation protection technician (RPT) involved.
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| The finding is more than minor because the performance deficiency is associated with the Program and Process attribute of the Occupational Radiation Safety Cornerstone and adversely affected the cornerstone objective to ensure the adequate protection of the worker's health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Specifically, the failure to implement the radiological controls established in the radiation worker permit (RWP) as-low-reasonably-achievable (ALARA) file caused workers to receive additional, unplanned dose to the workers. The finding was assessed using the Occupational Radiation Safety, Public Radiation Safety and was determined to be of very-low safety significance because this was not related to ALARA, did not result in an overexposure, or a substantial potential for overexposure, nor was the ability to assess dose compromised.
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| The radiological controls specified in RWP 10012059 for this activity were not implemented because the RPT assumed the scope of work and failed to review the RWP ALARA requirements before the briefing. Consequently, the inspectors determined that the cause of this incident involved a cross-cutting component in the human performance area for work practices. Specifically personnel work practices did not support human performance. (IMC 0310, H.4(a))
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| Inspection Report# : 2011005 (pdf)
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| Significance: Dec 31, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation FAILURE TO IMPLEMENT APPROPRIATE RADIOLOGICAL CONTROLS AFTER RADIATION PROTECTION IDENTIFIED THAT A WORKER WAS POTENTIALLY CONTAMINATED DUE TO INAPPROPRIATE PROTECTIVE CLOTHING.
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| A self-revealed finding of very low safety significance and an associated Non-Cited Violation of Technical Specification 5.4.1.a was identified. Specifically, the licensee failed to implement appropriate radiological controls after radiation protection identified that the worker was potentially contaminated due to the inappropriate protective clothing. This issue was entered in the licensee's corrective action program as AR 01017724. The licensee's corrective actions included the replacement of all contamination monitors used at the site. The new contamination monitors have a radon subtract feature designed to mitigate the large number nuisance alarms caused by radon interference at this site.
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| The finding is more than minor because if left uncorrected the performance deficiency would have the potential to lead to a more significant safety concern. Specifically, bypassing every level of defense could result in additional dose to worker outside the radiological control area. The finding was assessed using the Significance Determination Process and was determined to be of very-low safety significance because these radioactive material control issues were not related to transportation and dose to members of the public was less than 0.005 rem. The inspectors observed the operation of the new contamination monitors and response of radiation protection technicians assigned to monitor authorized exit points during a refueling outage. The new monitors did not exhibit nuisance alarms and the technicians treated every alarm as a potential contamination event until proven otherwise with another instrument. Futhermore, these technicans informed the inspectors the briefing received before the outage by the radiation protection manager about alarm response expectations. The inspectors determined that the events involved in this performance deficiency are not indicative of current performance. Consequently, the inspectors did not assess the performance deficiency for cross-cutting aspects.
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| Inspection Report# : 2011005 (pdf)
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| Public Radiation Safety Significance: Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IMPLEMENT PACKAGE DESIGN SPECIFICATIONS The inspectors identified a finding of very low-safety-significance and an associated non-cited violation of 10 CFR 71.5 for the failure to implement package design specifications. Specifically, the licensee failed to ensure the proper closure of a DOT 7A Type A package as required by Department of Transportation (DOT) regulations for packaging contained within 49 CFR 173. As a part of their corrective actions, the licensee completed a detailed review of all radioactive material shipments for the past 36 months to ensure Package Certification documents for other packages used as a Type A container satisfied requirements.
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| The finding was more than minor because it affected the Public Radiation Safety Cornerstone objective to ensure adequate protection of public health and safety from exposure to radioactive materials during transit. Specifically, the failure to correctly close a DOT Type A package could lead to a more significant safety concern by increasing the potential for a package breach occuring during transit. Using IMC 0609, Attachment D for the Public Radiation Safety Significance Determination Process (SDP) the inspectors determined the finding to be of very low safety significance.
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| This deficiency has a cross-cutting aspect in Human Performance (Resources). (IMC 0310 H.2(b))
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| Inspection Report# : 2011004 (pdf)
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| Physical Protection Although the NRC is actively overseeing the Security cornerstone, the Commission has decided that certain findings pertaining to security cornerstone will not be publicly available to ensure that potentially useful information is not provided to a possible adversary. Therefore, the cover letters to security inspection reports may be viewed.
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| Miscellaneous Last modified : May 29, 2012
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| Clinton 2Q/2012 Plant Inspection Findings Initiating Events Significance: Jun 30, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation FAILURE TO ESTABLISH INSTRUCTIONS APPROPRIATE FOR INSTALLATION OF SHUTDOWN AND UPSET LEVEL INSTRUMENT REFERENCE LEG PIPE A finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was self-revealed on December 18, 2011, when an automatic reactor scram signal and loss of decay heat removal occurred due to low reactor pressure vessel (RPV) water level while lowering water level to a target level following an RPV hydrostatic pressure test. The licensee failed to establish an adequate procedure to perfrom reinstallation of common shutdown and upset level instrument reference leg piping.
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| Specifically, inadequacies with the procedure resulted in improper filling and venting of the reference leg piping causing inaccurate indication of RPV level - an error of approximately 108 inches. In addition, the licensee failed to use appropriate acceptance criteria when accepting that the instrument restoration activities had been successfully acomplished. The licensee entered this issue into its corrective action program for evaluation and initiated corrective actions to revise procedures to more rigorously control the evolution and to train personnel.
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| The finding was of more than minor significance since it was associated with the Mitigating Systems cornerstone attribute of Procedure Quality and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to establish procedures adequate to maintain correct indication of RPV water level upon the reinstallation of permanent shutdown and upset level instrument reference leg piping. The finding was determined to be a licensee performance deficiency of very low safety significance based upon a Phase 3 Significance Determination Process evaluation by the Regional Senior Reactor Analyst with a risk result of approximately 4E-7 for Core Damage Frequency and no Large Early Release Frequency contribution since the event occurred more than 8 days from the beginning of the refueling outage. The inspectors concluded that this finding affected the cross cutting area of human performance. Specifically, in the area of work control, the licensee did not ensure that personnel, equipment, procedures, and other resources were available and adequate. Complete, accurate, and up-to-date procedures and work packages were not available to ensure nuclear safety (IMC 0310 H.2(c))
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| Inspection Report# : 2012003 (pdf)
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| Significance: Mar 31, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation FAILURE TO INCORPORATE OPERATING EXPERIENCE INTO PREVENTIVE MAINTENANCE ACTIVITIES.
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| A self-revealed finding of very low safety significance was identified with an associated Non-Cited Violation of 10 CFR 50.65, "Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants." The licensee failed to incorporate operating experience into its preventive maintenance practices associated with steam bypass system control circuit cards. Specifically, during two operating experience driven initiatives performed by the licensee in 2001 and 2007, and once again on September 24, 2011, the licnesee failed to implement any preventive maintenance activity for critical component circuit cards, which resulted in age-related failure and a reactor scram on November 29, 2011. The licensee initiated corrective actions to replace system circuit cards, perform periodic replacement/refurbishment maintenance activities, and trend circuit card performance during routine calibration.
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| The finding was of more than minor significance because it was sufficiently similar to Inspection Manual Chapter 0612, "Power Reactor Inspection Reports," Appendix E, "Examples of Minor Issues," Example 7 (c), in that this
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| violation of 10 CFR 50.65(a)(3) had a consequence "...such as equipment problems attributable to failure to take industry operating experience into account when practicable." The finding was a licensee performance deficiency of very low safety significance because it (1) did not contribute to the likelihood of a loss-of-collant accident initiator, (2) did not contribute to both the likelihood of a reactor scram AND the likelihood that mitigation equipment or functions would not be available, and (3) did not increase the likelihood of a fire or internal/external flooding event.
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| The inspectors concluded that this finding affected the cross-cutting area of human performance. Specifically, in the area of work control, the licensee did not appropriately coordinate work activities by incorporating actions to plan work activities to support long-term equipment reliability by scheduling maintenance as more preventive than reactive. (IMC 0310, H.3(b))
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| Inspection Report# : 2012002 (pdf)
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| Mitigating Systems Significance: Jun 30, 2012 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ENSURE TORNADO MISSLE PROTECTION FOR SAFETY RELATED COMPONENTS The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion III, "Design Control," when permanently installed tornado missile barrier protection was removed without adequate provisions to assure that appropriate quality standards were specified and included in design documents and that deviation from such standards was controlled. The licensee failed to ensure tornado missile protection for safety related components prior to and during maintenance affecting Control Room Ventilation (VC)
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| Train 'A'. Specifically, when the permanent missile barrier was removed, the licensee failed to ensure protection for two safety related radiation monitors, 1RIX-PR009C and 1RIX-PR009D and did not satisy requirements in modification documents for protection of VC panel 0PL72JA. The licensee entered this issue into its corrective action program for evaluation and performed immediate corrective actions to resolve the design deficiencies at the time of identification.
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| The finding was of more than minor significance because it was sufficiently similar to Inspection Manual Chapter 0612, "Power Reactor Inspection Reports," Appendix E, "Examples of Minor Issues," Example 3(a) in that this modification was found to contain errors significant enough that the modification required rework to correctly resolve design basis tornado concerns. The performance deficiency was also associated with the Mitigation Systems cornerstone attribute of Equipment Performance, and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to protect safety related components during work activities that modified the installed missile barrier required by the Clinton Power Station design. The finding was a licensee performance deficiency of very low safety significance because the design deficiency was confirmed to not result in an actual loss of operability or functionality. The inspectors concluded that the finding affected the cross cutting area of human performance. Specifically, in the area of work control, the licencee did not appropriately plan work activities by incorporating job site conditions and the need for adequate planned contingencies. (IMC 0310 H.3(a))
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| Inspection Report# : 2012003 (pdf)
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| Significance: Dec 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PROPERLY APPLY AN APPROVED ASME CODE CASE The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50.55a due to the licensee's failure to adequately apply American Society of Mechanical Engineers Section XI Code Case N-513-3 when it evaluated a degraded section of safety related shutdown service water system piping for operability. Specifically, the licensee failed to perform required daily walkdowns to confirm its analysis of conditions used in its operability evaluation remained valid. After this issue was identified by the inspectors, the licensee promptly resumed the daily compensatory action to verify the leak rate until the piping system was repaired.
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| The finding was of more than minor significance because it was associated with the Protection Against External Factors attribute of the Mitigating Systems Cornerstone, and it directly affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Additionally, if left uncorrected, improper application of an approved code case would become a more significant safety concern in that it could result in the failure to identify inoperable safety related piping. The finding was a licensee performance deficiency of very low safety significance because it did not result in an actual loss of safety function of a single train for greater than its Technical Specification allowed outage time. The inspectors concluded that there was no specific performance characteristic that was a significant cause to the performance deficiency in this instance; therefore no cross-cutting aspect was identified.
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| Inspection Report# : 2011005 (pdf)
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| Significance: Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PERFORM CODE REQUIRED CAUSE AND EFFECT FAILURE EVALUATIONS FOR DIESEL STARTING AIR AND FUEL OIL SYSTEM RELIEF VALVES.
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| The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50.55a. The licensee failed to perform American Society of Mechanical Engineers (ASME) Code required cause and effect failure evaluations for set pressure test failures of diesel generator (DG) starting air and fuel oil system relief valves. The licensee entered this issue into its corrective action program for evaluation and subsequently completed an engineering evaluation to address past operability of the associated DG starting air and fuel oil systems due to the relief valve test failures. The licensee also moved up its schedule to test the remaining relief valves.
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| The finding was of more than minor significance because it could lead to a more significant safety concern if left uncorrected. Specifically, the failure to perform Code required cause and effect evaluations for relief valve set pressure test failures could lead to a generic problem with valves in the same or other valve groups remaining uncorrected with a potential impact on operability of safety significant mitigating systems. Because the DG starting air and fuel oil systems are relied upon to support DG operability, the inspectors concluded that this issue was associated with the Mitigating Systems Cornerstone. The finding was determined to be a licensee performance deficiency of very low safety significance because the finding: (1) was not a design or qualification deficiency; (2) did not represent an actual loss of safety function of a system; (3) did not represent an actual loss of safety function of a single train or greater than its Technical Specification (TS) allowed outage time; (4) did not represent an actual loss of safety function of one or more non-TS trains of equipment designated as risk significant; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors concluded that the finding affected the cross-cutting area of human performance in that the licensee's work practices did not ensure adequate supervisory and management oversight of work activities, such that nuclear safety was supported.
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| Specifically, the relief valve test failures were left unresolved and were not evaluated as required by the Code for an extended period of time with several failed tests. (IMC 0310, H.4(C))
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| Inspection Report# : 2011004 (pdf)
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| Significance: Sep 30, 2011 Identified By: NRC Item Type: FIN Finding FAILURE TO CORRECT A CONDITION ADVERSE TO QUALITY FOR IMPROPERLY IMPLEMENTED ENGINEERING CORRECTIVE ACTIONS The inspectors identified a finding of very low safety significance due to the licensee's failure to effectively implement corrective actions for a condition adverse to quality described in Apparent Cause Evaluation 1095413, "NOS [Nuclear Oversight] Identified Improperly Implemented Engineering Corrective Actions Cause Repeat Operational Challenges." No violation of regulatory requirements was identified. The licensee entered this issue into its corrective action program to investigate the cause and to identify appropriate corrective actions.
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| The finding was of more than minor significance because it was associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone and directly affected the cornerstone objective to ensure the availability,
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| reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.
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| Specifically, improperly implemented engineering corrective actions could result in additional repeat operational equipment challenges. The finding was of very low safety significance because the issue: (1) was not a design or qualification deficiency; (2) did not represent an actual loss of safety function of a system; (3) did not represent an actual loss of safety function of a single train for greater than its Technical Specification (TS) allowed outage time; (4) did not represent an actual loss of safety function of one or more non-TS trains of equipment designated as risk significant; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors concluded that this finding affected the cross-cutting area of problem identification and resolution. Specifically, the licensee failed to take appropriate corrective actions to address known deficiencies in its process for tracking and closing work orders that implement corrective actions. The actions taken were neither lasting nor effective. (IMC 0310, P.1(d))
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| Inspection Report# : 2011004 (pdf)
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| Barrier Integrity Significance: Mar 31, 2012 Identified By: NRC Item Type: NCV NonCited Violation UNACCEPTABLE PRECONDITIONING OF LOW PRESSURE COOLANT INJECTION FROM RESIDUAL HEAT REMOVAL 'A' CHECK VALVE PRIOR TO LEAK RATE TEST MEASUREMENT The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings." The licensee failed to establish an adequate procedure to perform required leak rate testing for the Low Pressure Coolant Injection from Residual Heat Removal
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| 'A' Check Valve. Specifically, the surveillance test procedure resulted in unacceptable preconditioning of the valve prior to a leak rate test measurement due to improper test sequencing. In addition, the licensee failed to correctly evaluate a failed leak rate test of the valve. The licensee entered this issue into its corrective action program for evaluation and initiated corrective actions to revise the test procedure and train engineering personnel.
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| The finding was of more than minor significance since it was associated with the Procedure Quality attribute for the containment and adversely affected the Barrier Integrity Cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Because the preconditioning altered the as-found condition of the check valve, the data collected through the performance of the surveillance test was not fully indicative of the true valve performance trend. Additionally, the licensee's failure to correctly evaluate the initial failed leak rate test would become a more significant safety concern if left uncorrected because it could reasonably result in an unrecognized condition with a check valve failing to fulfill a safety related function. Therefore, this performance deficiency had a direct effect on the licensee's ability to fully assess the past operability, as well as the ability to trend as-found data for the purpose of assessing the reliability of the check valve.
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| The finding was a licensee performance deficiency of very low safety significance because it would not result in exceeding the Technical Specification limit for reactor coolant system leakage and would not have likely affected mitigation systems resulting in a loss of safety function. In addition, the finding did not represent an actual open pathway in the physical integrity of the reactor containment. Based on consultation and review with the Regional Senior Reactor Analyst, the inspectors concluded that the finding did not result in an increase in the likelihood of an initiating event such as an inter-system loss-of-coolant accident or a containment bypass event because the redundant isolation valve and closed loop system piping passed leak rate measurement test during the refueling outage with considerable margin. The inspectors concluded that this finding affected the cross cutting area of human performance.
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| Specifically, the licensee did not have adequately trained and knowledgeable personnel available to correctly evaluate the cause of the initial failed leak rate measurement test and to ensure that appropriate actions to correct the test sequence in the procedure were identified.
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| (IMC 0310,H.2(b))
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| Inspection Report# : 2012002 (pdf)
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| Significance: Dec 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO CONTROL THE WORK HOURS OF A COVERED WORKER.
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| The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 26.205(c) and (d) for the licensee's failure to schedule and control the work hours of a covered worker performing surveillance testing on containment isolation valves during the refueling outage. Specifically, an engineer performing local leak rate testing during the refueling outage was scheduled for successive 12-hour shifts and was inappropriately excluded from the work hour limits specified in 10 CFR 26.205(d)(1) and 10 CFR 26.205(d)(2). The licensee removed the engineer from covered work activities for the remainder of the refueling outage and reviewed the work activities of other engineers to ensure that any engineer performing covered work appropriately met work hour limits.
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| The finding was of more than minor signficance since the failure to schedule and control the work hours of a worker performing covered work, if left uncorrected, would become a more significant safety concern because it could reasonably result in human performance errors that could affect the function of safety-related structures, systems, and components. Since the issue involved leak rate testing on containment isolation valves performed during the refueling outage, the inspectors concluded that this issue was associated with the Barrier Integrity Cornerstone. The finding was a licensee performance deficiency of very low safety significance because it did not represent an actual open pathway in the physical integrity of the reactor containment. The inspectors concluded that this finding affected the cross-cutting area of human performance. Specifically, the engineer did not meet expectations regarding the performance of covered work activities because he did not challenge directions given to him by the leak rate test team supervisor and the leak rate test team supervisor did not meet expectations to ensure that the engineer was in compliance with the 10 CFR 26.205 (a) work requirements. Therefore, the inspectors concluded that the licensee's work practices which support human performance were less than effective. (IMC 0310, H.4(b))
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| Inspection Report# : 2011005 (pdf)
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| Significance: Dec 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation UNACCEPTABLE PRECONDITIONING OF REACTOR CORE ISOLATION COOLING SYSTEM CHECK VALVE PRIOR TO LEAK RATE TEST MEASUREMENT The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings." The licensee failed to establish an adequate procedure to perform required leak rate testing for the reactor core isolation cooling turbine exhaust check valve.
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| Specifically, the surveillance test procedure resulted in unacceptable preconditioning of the valve prior to an as-found leak rate test measurement. The licensee entered this issue into its corrective action program for evaluation and initiated a corrective action to revise the test procedure.
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| The finding was of more than minor significance since it was associated with the Procedure Quality Cornerstone attribute for the Containment and adversely affected the Barrier Integrity Cornerstone objective to provide rasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events.
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| Because the preconditioning altered the as-found condition of the check valve, the data collected through the performance of the surveillance test was not fully indicative of the true valve performance trend. Therefore, this performance deficiency had a direct effect on the licensee's ability to fully assess the past operability, as well as the ability to trend as-found data for the prupose of assessing the reliability of the check valve. The finding was a licensee performance deficiency of very low safety significance because it did not involve an actual open pathway in the physical integrity of the reactor containment. The inspectors concluded that this finding affected the cross-cutting area of problem identification and resolution. Specifically, the licensee did not implement operating experience into station processes, procedures, and training in that the licensee did not update/revise the surveillance test procedure consistent with NRC guidance and its corporate technical positon to prevent unacceptable preconditioning of the check valve.
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| (IMC 0310, P.2(b))
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| Inspection Report# : 2011005 (pdf)
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| Significance: Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO MEET TECHNICAL SPECIFICATION 3.7.3 FOR OPERABILITY OF CONTROL ROOM VENTILATION SYSTEM The inspectors identified a finding of very low safety significance with an associated non-cited violation of Technical Specification (TS) 3.7.3, "Control Room Ventilation System," following the discovery of a crack on the Train B Control Room ventilation (VC) system return fan hub during investigation of the cause for high noise and vibration levels observed on May 23, 2011. The licensee failed to correctly evaluate the operability of the Train B VC system return fan in a timely manner to prevent exceeding the TS allowed outage time for entry into Mode 3. The licensee replaced the fan and returned it to an operable status.
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| The failure to correctly evaluate a degraded/nonconforming condition potentially affecting the operability of structures, systems, and components (SSCs) required to be operable by TS would become a more significant safety concern if left uncorrected because it could reasonably result in an unrecognized condition of an SSC failing to fulfill a safety-related function. The finding was, therefore, of more than minor significance. Because the Control Room ventilation system supports the radiological barrier function to protect operators inside the Control Room following certain design basis accidents, the inspectors concluded that this issue was associated with the Barrier Integrity Cornerstone. The finding was a licensee performance deficiency of very low safety significance because it involved only a degradation of the radiological barrier function provided for the Control Room. The inspectors concluded that this finding affected the cross-cutting area of human performance. Specifically, licensee decision making to delay inspection of the fan hub and blades until after a new fan was delivered on site to confirm the initial operability determination was not conservative and not consistent with demonstrating that nuclear safety is an overriding priority.
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| (IMC 0310, H.1(b))
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| Inspection Report# : 2011004 (pdf)
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| Significance: Sep 30, 2011 Identified By: NRC Item Type: FIN Finding FAILURE TO EVALUATE OPERABILITY OF CONTROL ROOM VENTILATION SYSTEM FOR DEGRADED FLOW CONDITION The inspectors identified a finding of very low safety significance. The licensee failed to appropriately evaluate the operability of Control Room Ventilation Train A after identifying a degraded/nonconforming system flow condition while performing surveillance testing on April 1, 2011, that could have affected the ability of the system to perform its safety function. No violation of regulatory requirements was identified. The licensee initiated corrective actions to provide "read & sign" training for licensed operators and a procedure change to add an acceptance criterion for filtered flow rate in the surveillance test procedure.
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| The failure to correctly evaluate a degraded/nonconforming condition potentially affecting the operability of structures, systems, and components (SSCs) required to be operable by Technical Specifications (TS) would become a more significant safety concern if left uncorrected because it could reasonably result in an unrecognized condition of an SSC failing to fulfill a safety-related function. The finding was therefore of more than minor significance. Because the Control Room ventilation system supports the radiological barrier function to protect operators inside the Control Room following certain design basis accidents, the inspectors concluded that this issue was associated with the Barrier Integrity Cornerstone. The finding was a licensee performance deficiency of very low safety significance because it involved only a degradation of the radiological barrier function provided for the Control Room. The inspectors concluded that this finding affected the cross-cutting area of human performance. Specifically, licensee decision making using a systematic process to evaluate the operability of an SSC required to be operable by TS when a degraded/nonconforming condition was identified was not appropriately implemented as designed by licensed senior reactor operators. (IMC 0310 H.1(a))
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| Inspection Report# : 2011004 (pdf)
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| Emergency Preparedness Significance: Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation MISSING RESPIRATOR SPECTACLE KITS The inspectors identified a finding of very low safety significance and associated NCV of 10 CFR 50.54(q) for the failure to provide spectacle adapter kits for all eyeglass wearers (i.e, non-soft contact wearers) who were key emergency response organization (ERO) personnel that were potentially required to wear a self-contained breathing apparatus (SCBA) in order to fulfill emergency response functions. The licensee's corrective actions included revising procedures that govern the training and qualification of licensed operators to include steps that ensure licensed operators and other ERO members who require corrective lenses are provided SCBA lens inserts.
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| The finding was more than minor because it was associated with the Emergency Planning Cornerstone and, if left uncorrected, the performance deficiency has the potential to lead to a more significant safety concern, in that, emergency responders having inadequate vision could challenge the licensee's state of operational readiness and emergency response capabilities. The finding was assessed using IMC 0609, Attachment B , "Emergency Preparedness Significance Determination Process" and determined to be of very low safety significance because this failure to comply represented a planning standard issue, however, it did not result in a risk significant planning standard nor was it indicative of a planning standard functional failure. The failure to make provisions for respirator vision corrective lenses to licensed operators that required corrective lenses as a condition of their license was caused by a program weakness. Consequently, the cause of this finding has a cross-cutting aspect in the area of human performance. Specifically, the licensee did not ensure that equipment was available for key emergency response personnel. (IMC 0310,H.2(d))
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| Inspection Report# : 2011004 (pdf)
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| Occupational Radiation Safety Significance: Dec 31, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation FAILURE TO IMPLEMENT APPROPRIATE RADIOLOGICAL CONTROLS FOR THE REMOVAL OF INSULATION IN A POSTED HIGH CONTAMINATION AREA A self-revealed finding of very low safety significance and an associated Non-Cited Violation of Technical Specification 5.4.1.a was identified. Specifically, the licensee failed to implement appropriate radiological controls for the removal of insulation in a posted high contamination area. The issue was entered in the licensee's corrective action program as AR 01297713. The licensee's immediate corrective actions placed the job on hold, assessed the radiological significance for the issue, and suspended qualifications for the radiation protection technician (RPT) involved.
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| The finding is more than minor because the performance deficiency is associated with the Program and Process attribute of the Occupational Radiation Safety Cornerstone and adversely affected the cornerstone objective to ensure the adequate protection of the worker's health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Specifically, the failure to implement the radiological controls established in the radiation worker permit (RWP) as-low-reasonably-achievable (ALARA) file caused workers to receive additional, unplanned dose to the workers. The finding was assessed using the Occupational Radiation Safety, Public Radiation Safety and was determined to be of very-low safety significance because this was not related to ALARA, did not result in an overexposure, or a substantial potential for overexposure, nor was the ability to assess dose compromised.
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| The radiological controls specified in RWP 10012059 for this activity were not implemented because the RPT assumed the scope of work and failed to review the RWP ALARA requirements before the briefing. Consequently, the inspectors determined that the cause of this incident involved a cross-cutting component in the human
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| performance area for work practices. Specifically personnel work practices did not support human performance. (IMC 0310, H.4(a))
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| Inspection Report# : 2011005 (pdf)
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| Significance: Dec 31, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation FAILURE TO IMPLEMENT APPROPRIATE RADIOLOGICAL CONTROLS AFTER RADIATION PROTECTION IDENTIFIED THAT A WORKER WAS POTENTIALLY CONTAMINATED DUE TO INAPPROPRIATE PROTECTIVE CLOTHING.
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| A self-revealed finding of very low safety significance and an associated Non-Cited Violation of Technical Specification 5.4.1.a was identified. Specifically, the licensee failed to implement appropriate radiological controls after radiation protection identified that the worker was potentially contaminated due to the inappropriate protective clothing. This issue was entered in the licensee's corrective action program as AR 01017724. The licensee's corrective actions included the replacement of all contamination monitors used at the site. The new contamination monitors have a radon subtract feature designed to mitigate the large number nuisance alarms caused by radon interference at this site.
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| The finding is more than minor because if left uncorrected the performance deficiency would have the potential to lead to a more significant safety concern. Specifically, bypassing every level of defense could result in additional dose to worker outside the radiological control area. The finding was assessed using the Significance Determination Process and was determined to be of very-low safety significance because these radioactive material control issues were not related to transportation and dose to members of the public was less than 0.005 rem. The inspectors observed the operation of the new contamination monitors and response of radiation protection technicians assigned to monitor authorized exit points during a refueling outage. The new monitors did not exhibit nuisance alarms and the technicians treated every alarm as a potential contamination event until proven otherwise with another instrument. Futhermore, these technicans informed the inspectors the briefing received before the outage by the radiation protection manager about alarm response expectations. The inspectors determined that the events involved in this performance deficiency are not indicative of current performance. Consequently, the inspectors did not assess the performance deficiency for cross-cutting aspects.
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| Inspection Report# : 2011005 (pdf)
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| Public Radiation Safety Significance: Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IMPLEMENT PACKAGE DESIGN SPECIFICATIONS The inspectors identified a finding of very low-safety-significance and an associated non-cited violation of 10 CFR 71.5 for the failure to implement package design specifications. Specifically, the licensee failed to ensure the proper closure of a DOT 7A Type A package as required by Department of Transportation (DOT) regulations for packaging contained within 49 CFR 173. As a part of their corrective actions, the licensee completed a detailed review of all radioactive material shipments for the past 36 months to ensure Package Certification documents for other packages used as a Type A container satisfied requirements.
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| The finding was more than minor because it affected the Public Radiation Safety Cornerstone objective to ensure adequate protection of public health and safety from exposure to radioactive materials during transit. Specifically, the failure to correctly close a DOT Type A package could lead to a more significant safety concern by increasing the potential for a package breach occuring during transit. Using IMC 0609, Attachment D for the Public Radiation Safety Significance Determination Process (SDP) the inspectors determined the finding to be of very low safety significance.
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| This deficiency has a cross-cutting aspect in Human Performance (Resources). (IMC 0310 H.2(b))
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| Inspection Report# : 2011004 (pdf)
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| Security Although the Security Cornerstone is included in the Reactor Oversight Process assessment program, the Commission has decided that specific information related to findings and performance indicators pertaining to the Security Cornerstone will not be publicly available to ensure that security information is not provided to a possible adversary.
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| Other than the fact that a finding or performance indicator is Green or Greater-Than-Green, security related information will not be displayed on the public web page. Therefore, the cover letters to security inspection reports may be viewed.
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| Miscellaneous Last modified : September 12, 2012
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| 3Q/2012 Inspection Findings - Clinton Clinton 3Q/2012 Plant Inspection Findings Initiating Events Significance: Jun 30, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation FAILURE TO ESTABLISH INSTRUCTIONS APPROPRIATE FOR INSTALLATION OF SHUTDOWN AND UPSET LEVEL INSTRUMENT REFERENCE LEG PIPE A finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was self-revealed on December 18, 2011, when an automatic reactor scram signal and loss of decay heat removal occurred due to low reactor pressure vessel (RPV) water level while lowering water level to a target level following an RPV hydrostatic pressure test. The licensee failed to establish an adequate procedure to perfrom reinstallation of common shutdown and upset level instrument reference leg piping.
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| Specifically, inadequacies with the procedure resulted in improper filling and venting of the reference leg piping causing inaccurate indication of RPV level - an error of approximately 108 inches. In addition, the licensee failed to use appropriate acceptance criteria when accepting that the instrument restoration activities had been successfully acomplished. The licensee entered this issue into its corrective action program for evaluation and initiated corrective actions to revise procedures to more rigorously control the evolution and to train personnel.
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| The finding was of more than minor significance since it was associated with the Mitigating Systems cornerstone attribute of Procedure Quality and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to establish procedures adequate to maintain correct indication of RPV water level upon the reinstallation of permanent shutdown and upset level instrument reference leg piping. The finding was determined to be a licensee performance deficiency of very low safety significance based upon a Phase 3 Significance Determination Process evaluation by the Regional Senior Reactor Analyst with a risk result of approximately 4E-7 for Core Damage Frequency and no Large Early Release Frequency contribution since the event occurred more than 8 days from the beginning of the refueling outage. The inspectors concluded that this finding affected the cross cutting area of human performance. Specifically, in the area of work control, the licensee did not ensure that personnel, equipment, procedures, and other resources were available and adequate. Complete, accurate, and up-to-date procedures and work packages were not available to ensure nuclear safety (IMC 0310 H.2(c))
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| Inspection Report# : 2012003 (pdf)
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| Significance: Mar 31, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation FAILURE TO INCORPORATE OPERATING EXPERIENCE INTO PREVENTIVE MAINTENANCE ACTIVITIES.
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| A self-revealed finding of very low safety significance was identified with an associated Non-Cited Violation of 10 CFR 50.65, "Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants." The licensee failed to incorporate operating experience into its preventive maintenance practices associated with steam bypass system control circuit cards. Specifically, during two operating experience driven initiatives performed by the licensee in 2001 and 2007, and once again on September 24, 2011, the licnesee failed to implement any preventive Page 1 of 8
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| 3Q/2012 Inspection Findings - Clinton maintenance activity for critical component circuit cards, which resulted in age-related failure and a reactor scram on November 29, 2011. The licensee initiated corrective actions to replace system circuit cards, perform periodic replacement/refurbishment maintenance activities, and trend circuit card performance during routine calibration.
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| The finding was of more than minor significance because it was sufficiently similar to Inspection Manual Chapter 0612, "Power Reactor Inspection Reports," Appendix E, "Examples of Minor Issues," Example 7 (c), in that this violation of 10 CFR 50.65(a)(3) had a consequence "...such as equipment problems attributable to failure to take industry operating experience into account when practicable." The finding was a licensee performance deficiency of very low safety significance because it (1) did not contribute to the likelihood of a loss-of-collant accident initiator, (2) did not contribute to both the likelihood of a reactor scram AND the likelihood that mitigation equipment or functions would not be available, and (3) did not increase the likelihood of a fire or internal/external flooding event.
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| The inspectors concluded that this finding affected the cross-cutting area of human performance. Specifically, in the area of work control, the licensee did not appropriately coordinate work activities by incorporating actions to plan work activities to support long-term equipment reliability by scheduling maintenance as more preventive than reactive. (IMC 0310, H.3(b))
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| Inspection Report# : 2012002 (pdf)
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| Mitigating Systems Significance: TBD Sep 30, 2012 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO CORRECTLY ASSEMBLE DIESEL GENERATOR VENTILATION SYSTEM DAMPER RESULTED IN INOPERABLE DIESEL GENERATOR A finding of very low safety significance with an associated non-cited violation of 10 CFR 50, Appendix B, Criteria V, "Instructions, Procedures, and Drawings" was self-revealed on March 1, 2012 when the Division 1 diesel generator (DG) ventilation system supply damper was discovered failed closed with the ventilation supply fan running during a Division 1 DG surveillance test. The damper failure occurred due to the licensee's failure to establish an adequate procedure to perform maintenance. Specifically, the maintenance procedure did not contain an appropriate verification step to ensure that locknuts on the damper hydramotor coupling were tightly fastened. As a result, vibration of the coupling during operation over time caused the coupling to separate such that the damper would not open. The licensee entered this issue into its corrective action program for evaluation, repaired the damper, and initiated corrective actions to revise the maintenance procedure.
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| The finding was of more than minor significance since it was associated with the Procedure Quality attribute and adversely affected the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the damper failure rendered the Division 1 DG inoperable. Although the finding involved an actual loss of function of a single train for greater than its Technical Specification allowed outage time, it was determined to be of very low safety significance during a detailed quantitative Significance Determination Process review since the delta core damage frequency and delta large early release frequency were both determined to be megligible based upon crediting operator recovery actions to restore DG room ventilation. The inspectors condluded that this finding affected the cross-cutting area of human performance since adequate licensee resources involving personnel and procedures did not support successful human performance. Specifically, the maintenance procedure did not contain adequate instruction to ensure that locknuts on the damper hydramotor coupling were tightly fastened.
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| Inspection Report# : 2012004 (pdf)
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| Page 2 of 8
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| 3Q/2012 Inspection Findings - Clinton Significance: Sep 30, 2012 Identified By: NRC Item Type: FIN Finding FAILURE TO PERFORM ADEQUATE PAST OPERABILITY EVALUATIONS FOR EMERGENCY CORE COOLING SYSTEM RELIEF VALVES The inspectors identified a finding of very low safety significance associated with the licensee's failure to correctly evaluate the past operability of two emergency core cooling system (ECCS) relief valves that failed bench testing following replacement during the C1R13 refueling outage. No violation of regulatory requirements was identified because revised evaluations by the licensee determined that the valves would have satisfied their safety functions. The licensee entered this issue into its corrective action program for evaluation and initated corrective actions to revise the past operability evaluations to correct gross errors in the original evaluations.
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| The finding was of more than minor significance since the failure to correctly evaluate a degraded/nonconforming condition potentially affecting the operability of structures, systems, and components (SSC) required to be operable by Technical Specifications (TS) would become a more significant safety concern if left uncorrected because it could reasonably result in an unrecognized condition of an SSC failing to fulfill a safety-related function. The finding was a licensee performance deficiency of very low safety significance because it: (1) was not a design or qualification deficiency; (2) did not represent an actual loss of function of a system; (3) did not represent an actual loss of function of a single train or two separate trains for greater than its TS allowed outage time; (4) did not represent an actual loss of function of one or more non-TS trains of equipment designated as high safety significant; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors concluded that this finding affected the cross-cutting area of human performance since licensee engineering staff failed to thoroughly and correctly evaluate past operability of the two ECCS relief valves due to inattention to detail. Human error prevention techniques were not appropriately employed to support human performance. The most significant concerns were that the independent technical reviewer did not independently validate information contained in the past operability evaluations by reviewing the valve test records and, that neither the independent technical reviewer nor the engineering supervisory reviewer challended the unwarranted past operability conclusion reached for the 1E12-F025C test failure.
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| Inspection Report# : 2012004 (pdf)
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| Significance: Jun 30, 2012 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ENSURE TORNADO MISSLE PROTECTION FOR SAFETY RELATED COMPONENTS The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion III, "Design Control," when permanently installed tornado missile barrier protection was removed without adequate provisions to assure that appropriate quality standards were specified and included in design documents and that deviation from such standards was controlled. The licensee failed to ensure tornado missile protection for safety related components prior to and during maintenance affecting Control Room Ventilation (VC)
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| Train 'A'. Specifically, when the permanent missile barrier was removed, the licensee failed to ensure protection for two safety related radiation monitors, 1RIX-PR009C and 1RIX-PR009D and did not satisy requirements in modification documents for protection of VC panel 0PL72JA. The licensee entered this issue into its corrective action program for evaluation and performed immediate corrective actions to resolve the design deficiencies at the time of identification.
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| The finding was of more than minor significance because it was sufficiently similar to Inspection Manual Chapter 0612, "Power Reactor Inspection Reports," Appendix E, "Examples of Minor Issues," Example 3(a) in that this modification was found to contain errors significant enough that the modification required rework to correctly resolve design basis tornado concerns. The performance deficiency was also associated with the Mitigation Systems Page 3 of 8
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| 3Q/2012 Inspection Findings - Clinton cornerstone attribute of Equipment Performance, and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to protect safety related components during work activities that modified the installed missile barrier required by the Clinton Power Station design. The finding was a licensee performance deficiency of very low safety significance because the design deficiency was confirmed to not result in an actual loss of operability or functionality. The inspectors concluded that the finding affected the cross cutting area of human performance. Specifically, in the area of work control, the licencee did not appropriately plan work activities by incorporating job site conditions and the need for adequate planned contingencies. (IMC 0310 H.3(a))
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| Inspection Report# : 2012003 (pdf)
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| Significance: Dec 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PROPERLY APPLY AN APPROVED ASME CODE CASE The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50.55a due to the licensee's failure to adequately apply American Society of Mechanical Engineers Section XI Code Case N-513-3 when it evaluated a degraded section of safety related shutdown service water system piping for operability. Specifically, the licensee failed to perform required daily walkdowns to confirm its analysis of conditions used in its operability evaluation remained valid. After this issue was identified by the inspectors, the licensee promptly resumed the daily compensatory action to verify the leak rate until the piping system was repaired.
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| The finding was of more than minor significance because it was associated with the Protection Against External Factors attribute of the Mitigating Systems Cornerstone, and it directly affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Additionally, if left uncorrected, improper application of an approved code case would become a more significant safety concern in that it could result in the failure to identify inoperable safety related piping. The finding was a licensee performance deficiency of very low safety significance because it did not result in an actual loss of safety function of a single train for greater than its Technical Specification allowed outage time. The inspectors concluded that there was no specific performance characteristic that was a significant cause to the performance deficiency in this instance; therefore no cross-cutting aspect was identified.
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| Inspection Report# : 2011005 (pdf)
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| Barrier Integrity Significance: Mar 31, 2012 Identified By: NRC Item Type: NCV NonCited Violation UNACCEPTABLE PRECONDITIONING OF LOW PRESSURE COOLANT INJECTION FROM RESIDUAL HEAT REMOVAL 'A' CHECK VALVE PRIOR TO LEAK RATE TEST MEASUREMENT The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings." The licensee failed to establish an adequate procedure to perform required leak rate testing for the Low Pressure Coolant Injection from Residual Heat Removal
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| 'A' Check Valve. Specifically, the surveillance test procedure resulted in unacceptable preconditioning of the valve prior to a leak rate test measurement due to improper test sequencing. In addition, the licensee failed to correctly evaluate a failed leak rate test of the valve. The licensee entered this issue into its corrective action program for evaluation and initiated corrective actions to revise the test procedure and train engineering personnel.
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| Page 4 of 8
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| 3Q/2012 Inspection Findings - Clinton The finding was of more than minor significance since it was associated with the Procedure Quality attribute for the containment and adversely affected the Barrier Integrity Cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Because the preconditioning altered the as-found condition of the check valve, the data collected through the performance of the surveillance test was not fully indicative of the true valve performance trend. Additionally, the licensee's failure to correctly evaluate the initial failed leak rate test would become a more significant safety concern if left uncorrected because it could reasonably result in an unrecognized condition with a check valve failing to fulfill a safety related function. Therefore, this performance deficiency had a direct effect on the licensee's ability to fully assess the past operability, as well as the ability to trend as-found data for the purpose of assessing the reliability of the check valve.
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| The finding was a licensee performance deficiency of very low safety significance because it would not result in exceeding the Technical Specification limit for reactor coolant system leakage and would not have likely affected mitigation systems resulting in a loss of safety function. In addition, the finding did not represent an actual open pathway in the physical integrity of the reactor containment. Based on consultation and review with the Regional Senior Reactor Analyst, the inspectors concluded that the finding did not result in an increase in the likelihood of an initiating event such as an inter-system loss-of-coolant accident or a containment bypass event because the redundant isolation valve and closed loop system piping passed leak rate measurement test during the refueling outage with considerable margin. The inspectors concluded that this finding affected the cross cutting area of human performance.
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| Specifically, the licensee did not have adequately trained and knowledgeable personnel available to correctly evaluate the cause of the initial failed leak rate measurement test and to ensure that appropriate actions to correct the test sequence in the procedure were identified.
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| (IMC 0310,H.2(b))
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| Inspection Report# : 2012002 (pdf)
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| Significance: Dec 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO CONTROL THE WORK HOURS OF A COVERED WORKER.
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| The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 26.205(c) and (d) for the licensee's failure to schedule and control the work hours of a covered worker performing surveillance testing on containment isolation valves during the refueling outage. Specifically, an engineer performing local leak rate testing during the refueling outage was scheduled for successive 12-hour shifts and was inappropriately excluded from the work hour limits specified in 10 CFR 26.205(d)(1) and 10 CFR 26.205(d)(2). The licensee removed the engineer from covered work activities for the remainder of the refueling outage and reviewed the work activities of other engineers to ensure that any engineer performing covered work appropriately met work hour limits.
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| The finding was of more than minor signficance since the failure to schedule and control the work hours of a worker performing covered work, if left uncorrected, would become a more significant safety concern because it could reasonably result in human performance errors that could affect the function of safety-related structures, systems, and components. Since the issue involved leak rate testing on containment isolation valves performed during the refueling outage, the inspectors concluded that this issue was associated with the Barrier Integrity Cornerstone. The finding was a licensee performance deficiency of very low safety significance because it did not represent an actual open pathway in the physical integrity of the reactor containment. The inspectors concluded that this finding affected the cross-cutting area of human performance. Specifically, the engineer did not meet expectations regarding the performance of covered work activities because he did not challenge directions given to him by the leak rate test team supervisor and the leak rate test team supervisor did not meet expectations to ensure that the engineer was in compliance with the 10 CFR 26.205 (a) work requirements. Therefore, the inspectors concluded that the licensee's work practices which support human performance were less than effective. (IMC 0310, H.4(b))
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| Page 5 of 8
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| 3Q/2012 Inspection Findings - Clinton Inspection Report# : 2011005 (pdf)
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| Significance: Dec 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation UNACCEPTABLE PRECONDITIONING OF REACTOR CORE ISOLATION COOLING SYSTEM CHECK VALVE PRIOR TO LEAK RATE TEST MEASUREMENT The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings." The licensee failed to establish an adequate procedure to perform required leak rate testing for the reactor core isolation cooling turbine exhaust check valve.
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| Specifically, the surveillance test procedure resulted in unacceptable preconditioning of the valve prior to an as-found leak rate test measurement. The licensee entered this issue into its corrective action program for evaluation and initiated a corrective action to revise the test procedure.
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| The finding was of more than minor significance since it was associated with the Procedure Quality Cornerstone attribute for the Containment and adversely affected the Barrier Integrity Cornerstone objective to provide rasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events.
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| Because the preconditioning altered the as-found condition of the check valve, the data collected through the performance of the surveillance test was not fully indicative of the true valve performance trend. Therefore, this performance deficiency had a direct effect on the licensee's ability to fully assess the past operability, as well as the ability to trend as-found data for the prupose of assessing the reliability of the check valve. The finding was a licensee performance deficiency of very low safety significance because it did not involve an actual open pathway in the physical integrity of the reactor containment. The inspectors concluded that this finding affected the cross-cutting area of problem identification and resolution. Specifically, the licensee did not implement operating experience into station processes, procedures, and training in that the licensee did not update/revise the surveillance test procedure consistent with NRC guidance and its corporate technical positon to prevent unacceptable preconditioning of the check valve.
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| (IMC 0310, P.2(b))
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| Inspection Report# : 2011005 (pdf)
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| Emergency Preparedness Occupational Radiation Safety Significance: Dec 31, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation FAILURE TO IMPLEMENT APPROPRIATE RADIOLOGICAL CONTROLS FOR THE REMOVAL OF INSULATION IN A POSTED HIGH CONTAMINATION AREA A self-revealed finding of very low safety significance and an associated Non-Cited Violation of Technical Specification 5.4.1.a was identified. Specifically, the licensee failed to implement appropriate radiological controls for the removal of insulation in a posted high contamination area. The issue was entered in the licensee's corrective action program as AR 01297713. The licensee's immediate corrective actions placed the job on hold, assessed the radiological significance for the issue, and suspended qualifications for the radiation protection technician (RPT)
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| Page 6 of 8
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| 3Q/2012 Inspection Findings - Clinton involved.
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| The finding is more than minor because the performance deficiency is associated with the Program and Process attribute of the Occupational Radiation Safety Cornerstone and adversely affected the cornerstone objective to ensure the adequate protection of the worker's health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Specifically, the failure to implement the radiological controls established in the radiation worker permit (RWP) as-low-reasonably-achievable (ALARA) file caused workers to receive additional, unplanned dose to the workers. The finding was assessed using the Occupational Radiation Safety, Public Radiation Safety and was determined to be of very-low safety significance because this was not related to ALARA, did not result in an overexposure, or a substantial potential for overexposure, nor was the ability to assess dose compromised.
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| The radiological controls specified in RWP 10012059 for this activity were not implemented because the RPT assumed the scope of work and failed to review the RWP ALARA requirements before the briefing. Consequently, the inspectors determined that the cause of this incident involved a cross-cutting component in the human performance area for work practices. Specifically personnel work practices did not support human performance. (IMC 0310, H.4(a))
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| Inspection Report# : 2011005 (pdf)
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| Significance: Dec 31, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation FAILURE TO IMPLEMENT APPROPRIATE RADIOLOGICAL CONTROLS AFTER RADIATION PROTECTION IDENTIFIED THAT A WORKER WAS POTENTIALLY CONTAMINATED DUE TO INAPPROPRIATE PROTECTIVE CLOTHING.
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| A self-revealed finding of very low safety significance and an associated Non-Cited Violation of Technical Specification 5.4.1.a was identified. Specifically, the licensee failed to implement appropriate radiological controls after radiation protection identified that the worker was potentially contaminated due to the inappropriate protective clothing. This issue was entered in the licensee's corrective action program as AR 01017724. The licensee's corrective actions included the replacement of all contamination monitors used at the site. The new contamination monitors have a radon subtract feature designed to mitigate the large number nuisance alarms caused by radon interference at this site.
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| The finding is more than minor because if left uncorrected the performance deficiency would have the potential to lead to a more significant safety concern. Specifically, bypassing every level of defense could result in additional dose to worker outside the radiological control area. The finding was assessed using the Significance Determination Process and was determined to be of very-low safety significance because these radioactive material control issues were not related to transportation and dose to members of the public was less than 0.005 rem. The inspectors observed the operation of the new contamination monitors and response of radiation protection technicians assigned to monitor authorized exit points during a refueling outage. The new monitors did not exhibit nuisance alarms and the technicians treated every alarm as a potential contamination event until proven otherwise with another instrument. Futhermore, these technicans informed the inspectors the briefing received before the outage by the radiation protection manager about alarm response expectations. The inspectors determined that the events involved in this performance deficiency are not indicative of current performance. Consequently, the inspectors did not assess the performance deficiency for cross-cutting aspects.
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| Inspection Report# : 2011005 (pdf)
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| Public Radiation Safety Page 7 of 8
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| 3Q/2012 Inspection Findings - Clinton Security Although the Security Cornerstone is included in the Reactor Oversight Process assessment program, the Commission has decided that specific information related to findings and performance indicators pertaining to the Security Cornerstone will not be publicly available to ensure that security information is not provided to a possible adversary.
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| Other than the fact that a finding or performance indicator is Green or Greater-Than-Green, security related information will not be displayed on the public web page. Therefore, the cover letters to security inspection reports may be viewed.
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| Miscellaneous Last modified : November 30, 2012 Page 8 of 8
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| 4Q/2012 Inspection Findings - Clinton Clinton 4Q/2012 Plant Inspection Findings Initiating Events Significance: Dec 31, 2012 Identified By: Self-Revealing Item Type: FIN Finding FAILURE TO COMPLETE AN ADEQUATE EXTENT CONDITION REVIEW AND TO CORRECT A PREVIOUSLY IDENTIFIED DESIGN PROBLEM RESULTED IN A TRIP OF THE EMERGENCY RESERVE AUXILIARY TRANSFORMER A finding of very low safety significance was self-revealed when the emergency reserve auxiliary transformer (ERAT) tripped during troubleshooting activities to isolate a direct current system ground following heavy rainfall.
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| The ERAT trip occurred due to the presence of a latent design error identified on seal-in relays in the ERAT's control circuitry and the licensee's failure to adequately evaluate and correct it during its extent of condition review of the problem after it was identified in September 2002. The licensee restored the ERAT to service and implemented a modification to correct the latent design problem. Because the ERAT is not safety related, no violation of regulatory requirements was identified.
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| The finding was of more than minor safety significance because it was sufficiently similar to several examples in Inspection Manual Chapter 0612, "Power Reactor Inspection Reports," Appendix E, "Examples of Minor Issues,"
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| wherein licnesees failed to adequately correct conditions adverse to quality and the consequences had some safety impact. The performance deficiency was also associated with the Equipment Performance attribute and adversely affected the Initiating Events Cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, when the ERAT tripped safety related 4160 volt bus 1A1, which had been powered by the ERAT, momentarily lost power. With the momentary loss of power several plant safety systems were affected including a loss of secondary containment differential pressure. The finding was a licensee performance deficiency of very low safety significance because it: (1) did not involve a loss-of-coolant accident initiator; (2) did not cause a reactor trip AND the loss of mitigation equipment; (3) did not involve the complete or partial loss of a support system that contributes to the likelihood of, or cause, an initiating event AND affect mitigation equipment; and (4) did not increase the frequency of a fire or internal flooding initiating event. While the finding did involve a partial loss of a support system (i.e., offsite power) that contributes to the likelihood of an initiating event, mitigation equipment was not adversely affected by the momentary loss of power. The inspectors concluded that because the licensee's missed opportunity to correct the latent design error occurred in 2002 and no other more recent opportunities reasonably existed to identify and correct the problem, this issue would not be reflective of current licensee performance and no cross-cutting aspect was identified.
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| Inspection Report# : 2012005 (pdf)
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| Significance: Mar 31, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation FAILURE TO INCORPORATE OPERATING EXPERIENCE INTO PREVENTIVE MAINTENANCE ACTIVITIES.
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| A self-revealed finding of very low safety significance was identified with an associated Non-Cited Violation of 10 CFR 50.65, "Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants." The licensee failed to incorporate operating experience into its preventive maintenance practices associated with steam bypass system control circuit cards. Specifically, during two operating experience driven initiatives performed by the licensee in 2001 and 2007, and once again on September 24, 2011, the licnesee failed to implement any preventive maintenance activity for critical component circuit cards, which resulted in age-related failure and a reactor scram on November 29, 2011. The licensee initiated corrective actions to replace system circuit cards, perform periodic replacement/refurbishment maintenance activities, and trend circuit card performance during routine calibration.
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| Page 1 of 6
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| 4Q/2012 Inspection Findings - Clinton The finding was of more than minor significance because it was sufficiently similar to Inspection Manual Chapter 0612, "Power Reactor Inspection Reports," Appendix E, "Examples of Minor Issues," Example 7 (c), in that this violation of 10 CFR 50.65(a)(3) had a consequence "...such as equipment problems attributable to failure to take industry operating experience into account when practicable." The finding was a licensee performance deficiency of very low safety significance because it: (1) did not contribute to the likelihood of a loss-of-collant accident initiator, (2) did not contribute to both the likelihood of a reactor scram AND the likelihood that mitigation equipment or functions would not be available, and (3) did not increase the likelihood of a fire or internal/external flooding event.
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| The inspectors concluded that this finding affected the cross-cutting area of human performance. Specifically, in the area of work control, the licensee did not appropriately coordinate work activities by incorporating actions to plan work activities to support long-term equipment reliability by scheduling maintenance as more preventive than reactive. (IMC 0310, H.3(b))
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| Inspection Report# : 2012002 (pdf)
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| Mitigating Systems Significance: Dec 31, 2012 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO SATISFY 10 CFR 50.73 REPORTING REQUIREMENTS FOR A CONDITION PROHIBITED BY TECHNICAL SPECIFICATIONS.
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| The inspectors identified a finding of very low safety significance (Green) with an associated Severity Level IV Non-Cited Vilation of the NRC's reporting requirements in 10 CFR 50.73, "Licensee Event Report System." The licensee failed to submit a required Licensee Event Report (LER) within 60 days after the discovery of an event that was reportable in accordance with 10 CFR 50.73(a)(2)(i)(B) as a condition which was prohibited by the plant's Technical Specifications (TS) and 10 CFR 50.73(a)(2)(v)(B) as a condition that could have prevented the fulfillment of a safety function. The condition involved an inoperable diesel generator (DG) for longer than the TS completion time for restoration. The licensee subsequently submitted the required LER.
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| Because this violation of the NRC's reporting requirements affected the NRC's ability to perform its regulatory function, the inspectors evaluated the violation using the traditional enforcement process in accordance with the NRC Enforcement Policy and assessed the significance of the underlying issue using the Significance Determination Process. The finding was of more than minor significance because the NRC relies on licensees to identify and report conditions or events meeting the criteria specified in the TS and the regulations in order to perform its regulatory function and, therefore if left uncorrected it could lead to a more significant safety concern. The inspectors previously determined that the underlying issue (i.e., the failure to correctly assembly a DG ventilation system damper that resulted in an inoperable DG) was a finding of very low safety significance during a detailed risk evaluation.
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| Consistent with the guidance in Section 6.9, Paragraph d.9, of the NRC Enforcement Policy, the violation associated with this finding was determined to be a Severity Level IV Violation. This finding affected the cross-cutting area of human performance. Specifically, the licnesee's decision making process while evaluating the reportability of the condition with respect to the reporting requirements in 10 CFR 50.73 was inadequate. IMC 0310 H.1(a))
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| Inspection Report# : 2012005 (pdf)
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| Significance: Sep 30, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation FAILURE TO CORRECTLY ASSEMBLE DIESEL GENERATOR VENTILATION SYSTEM DAMPER RESULTED IN INOPERABLE DIESEL GENERATOR A finding of very low safety significance with an associated non-cited violation of 10 CFR 50, Appendix B, Criteria V, "Instructions, Procedures, and Drawings" was self-revealed on March 1, 2012 when the Division 1 diesel generator (DG) ventilation system supply damper was discovered failed closed with the ventilation supply fan running during a Division 1 DG surveillance test. The damper failure occurred due to the licensee's failure to establish an adequate Page 2 of 6
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| 4Q/2012 Inspection Findings - Clinton procedure to perform maintenance. Specifically, the maintenance procedure did not contain an appropriate verification step to ensure that locknuts on the damper hydramotor coupling were tightly fastened. As a result, vibration of the coupling during operation over time caused the coupling to separate such that the damper would not open. The licensee entered this issue into its corrective action program for evaluation, repaired the damper, and initiated corrective actions to revise the maintenance procedure.
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| The finding was of more than minor significance since it was associated with the Procedure Quality attribute and adversely affected the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the damper failure rendered the Division 1 DG inoperable. Although the finding involved an actual loss of function of a single train for greater than its Technical Specification allowed outage time, it was determined to be of very low safety significance during a detailed quantitative Significance Determination Process review since the delta core damage frequency and delta large early release frequency were both determined to be negligible based upon crediting operator recovery actions to restore DG room ventilation. The inspectors condluded that this finding affected the cross-cutting area of human performance since adequate licensee resources involving personnel and procedures did not support successful human performance. Specifically, the maintenance procedure did not contain adequate instruction to ensure that locknuts on the damper hydramotor coupling were tightly fastened. (IMC 0310, H.2(a))
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| Inspection Report# : 2012004 (pdf)
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| Significance: Sep 30, 2012 Identified By: NRC Item Type: FIN Finding FAILURE TO PERFORM ADEQUATE PAST OPERABILITY EVALUATIONS FOR EMERGENCY CORE COOLING SYSTEM RELIEF VALVES The inspectors identified a finding of very low safety significance associated with the licensee's failure to correctly evaluate the past operability of two emergency core cooling system (ECCS) relief valves that failed bench testing following replacement during the C1R13 refueling outage. No violation of regulatory requirements was identified because revised evaluations by the licensee determined that the valves would have satisfied their safety functions. The licensee entered this issue into its corrective action program for evaluation and initated corrective actions to revise the past operability evaluations to correct gross errors in the original evaluations.
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| The finding was of more than minor significance since the failure to correctly evaluate a degraded/nonconforming condition potentially affecting the operability of structures, systems, and components (SSC) required to be operable by Technical Specifications (TS) would become a more significant safety concern if left uncorrected because it could reasonably result in an unrecognized condition of an SSC failing to fulfill a safety-related function. The finding was a licensee performance deficiency of very low safety significance because it: (1) was not a design or qualification deficiency; (2) did not represent an actual loss of function of a system; (3) did not represent an actual loss of function of a single train or two separate trains for greater than its TS allowed outage time; (4) did not represent an actual loss of function of one or more non-TS trains of equipment designated as high safety significant; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors concluded that this finding affected the cross-cutting area of human performance since licensee engineering staff failed to thoroughly and correctly evaluate past operability of the two ECCS relief valves due to inattention to detail. Human error prevention techniques were not appropriately employed to support human performance. The most significant concerns were that the independent technical reviewer did not independently validate information contained in the past operability evaluations by reviewing the valve test records and, that neither the independent technical reviewer nor the engineering supervisory reviewer challenged the unwarranted past operability conclusion reached for the 1E12-F025C test failure. (IMC 0310,H.4 (a)
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| Inspection Report# : 2012004 (pdf)
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| Significance: Jun 30, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation FAILURE TO ESTABLISH INSTRUCTIONS APPROPRIATE FOR INSTALLATION OF SHUTDOWN AND UPSET LEVEL INSTRUMENT REFERENCE LEG PIPE Page 3 of 6
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| 4Q/2012 Inspection Findings - Clinton A finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was self-revealed on December 18, 2011, when an automatic reactor scram signal and loss of decay heat removal occurred due to low reactor pressure vessel (RPV) water level while lowering water level to a target level following an RPV hydrostatic pressure test. The licensee failed to establish an adequate procedure to perfrom reinstallation of common shutdown and upset level instrument reference leg piping.
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| Specifically, inadequacies with the procedure resulted in improper filling and venting of the reference leg piping causing inaccurate indication of RPV level - an error of approximately 108 inches. In addition, the licensee failed to use appropriate acceptance criteria when accepting that the instrument restoration activities had been successfully acomplished. The licensee entered this issue into its corrective action program for evaluation and initiated corrective actions to revise procedures to more rigorously control the evolution and to train personnel.
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| The finding was of more than minor significance since it was associated with the Mitigating Systems cornerstone attribute of Procedure Quality and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to establish procedures adequate to maintain correct indication of RPV water level upon the reinstallation of permanent shutdown and upset level instrument reference leg piping. The finding was determined to be a licensee performance deficiency of very low safety significance based upon a Phase 3 Significance Determination Process evaluation by the Regional Senior Reactor Analyst with a risk result of approximately 4E-7 for Core Damage Frequency and no Large Early Release Frequency contribution since the event occurred more than 8 days from the beginning of the refueling outage. The inspectors concluded that this finding affected the cross cutting area of human performance. Specifically, in the area of work control, the licensee did not ensure that personnel, equipment, procedures, and other resources were available and adequate. Complete, accurate, and up-to-date procedures and work packages were not available to ensure nuclear safety (IMC 0310 H.2(c))
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| Inspection Report# : 2012003 (pdf)
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| Significance: Jun 30, 2012 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ENSURE TORNADO MISSLE PROTECTION FOR SAFETY RELATED COMPONENTS The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion III, "Design Control," when permanently installed tornado missile barrier protection was removed without adequate provisions to assure that appropriate quality standards were specified and included in design documents and that deviation from such standards was controlled. The licensee failed to ensure tornado missile protection for safety related components prior to and during maintenance affecting Control Room Ventilation (VC)
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| Train 'A'. Specifically, when the permanent missile barrier was removed, the licensee failed to ensure protection for two safety related radiation monitors, 1RIX-PR009C and 1RIX-PR009D and did not satisy requirements in modification documents for protection of VC panel 0PL72JA. The licensee entered this issue into its corrective action program for evaluation and performed immediate corrective actions to resolve the design deficiencies at the time of identification.
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| The finding was of more than minor significance because it was sufficiently similar to Inspection Manual Chapter 0612, "Power Reactor Inspection Reports," Appendix E, "Examples of Minor Issues," Example 3(a) in that this modification was found to contain errors significant enough that the modification required rework to correctly resolve design basis tornado concerns. The performance deficiency was also associated with the Mitigation Systems cornerstone attribute of Equipment Performance, and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to protect safety related components during work activities that modified the installed missile barrier required by the Clinton Power Station design. The finding was a licensee performance deficiency of very low safety significance because the design deficiency was confirmed to not result in an actual loss of operability or functionality. The inspectors concluded that the finding affected the cross cutting area of human performance. Specifically, in the area of work control, the licencee did not appropriately plan work activities by incorporating job site conditions and the need for adequate planned contingencies. (IMC 0310 H.3(a))
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| Inspection Report# : 2012003 (pdf)
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| Page 4 of 6
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| 4Q/2012 Inspection Findings - Clinton Barrier Integrity Significance: Dec 31, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation FAILURE TO PERFORM PREVENTIVE MAINTENANCE ON STANDBY GAS TREATMENT SYSTEM RELAY 0UAY-VG506D.
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| A finding of very low safety significance with an associated Non-Cited Violation of Technical Specification (TS) 5.4.1.a. was self-revealed when the age-related failure of Standy Gas Treatment (VG) system relay 0UAY-VG506D caused the removal of VG Train A electric heater 0VG04AA from operation, an entry into TS 3.5.4.3 due to the inoperability of VG Train A, and an unplanned on-line plant risk condition increase from Green to Yellow. The relay failure occurred due to the licensee's failure to perform any replacement preventive maintenance on the component throughout the history of plant operation. During two separate independent reviews performed by the licensee on July 15, 2011, and on August 24, 2011, the licensee failed to correctly classify the component in accordance with its preventive maintenance procedure. This resulted in no replacement maintenance activity ever being performed for the relay and its eventual failure on August 22, 2012. The licensee initiated corrective actions to replace the relay and put in place the appropriate preventive maintenance actions.
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| The finding was of more than minor safety significance because it was sufficiently similar to several examples in Inspection Manual Chapter 0612, "Power Reactor Inspection Reports," Appendix E, "Examples of Minor Issues,"
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| wherein licensees failed to adequately implement procedural requirements and the consequences had some safety impact. The performance deficiency was also associated with the SSC [Systems, Structures, and Components] and Barrier Performance attribute and adversely affected the Barrier Integrity Cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events.
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| Specifically, the age-related failure of 0UAY-VG506D on August 22, 2012 rendered VG Train A inoperable and caused an unplanned increase in the plant's on-line risk condition from Green to Yellow. The finding was a licensee performance deficiency of very low safety significance because it only represented a degradation of the radiological barrier function provided for the Auxiliary Building and the Fuel Building and was not a complete loss of the barrier function provided by the VG system since VG Train B remained operable. This finding affected the cross-cutting area of human performance. Specifically, in the area of work control, the licensee did not appropriately coordinate work activities by incorporating actions to plan work activities to support longterm equipment reliability by scheduling maintenance as more preventive than reactive. (IMC 0310, H.3(b))
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| Inspection Report# : 2012005 (pdf)
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| Significance: Mar 31, 2012 Identified By: NRC Item Type: NCV NonCited Violation UNACCEPTABLE PRECONDITIONING OF LOW PRESSURE COOLANT INJECTION FROM RESIDUAL HEAT REMOVAL 'A' CHECK VALVE PRIOR TO LEAK RATE TEST MEASUREMENT The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings." The licensee failed to establish an adequate procedure to perform required leak rate testing for the Low Pressure Coolant Injection from Residual Heat Removal
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| 'A' Check Valve. Specifically, the surveillance test procedure resulted in unacceptable preconditioning of the valve prior to a leak rate test measurement due to improper test sequencing. In addition, the licensee failed to correctly evaluate a failed leak rate test of the valve. The licensee entered this issue into its corrective action program for evaluation and initiated corrective actions to revise the test procedure and train engineering personnel.
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| The finding was of more than minor significance since it was associated with the Procedure Quality attribute for the containment and adversely affected the Barrier Integrity Cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Because the preconditioning altered the as-found condition of the check valve, the data collected through the performance of the surveillance test was not fully indicative of the true valve performance trend. Additionally, the licensee's failure to correctly evaluate the initial failed leak rate test would become a more significant safety concern if left uncorrected because it could reasonably result in an unrecognized condition with a check valve failing to fulfill a safety related Page 5 of 6
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| 4Q/2012 Inspection Findings - Clinton function. Therefore, this performance deficiency had a direct effect on the licensee's ability to fully assess the past operability, as well as the ability to trend as-found data for the purpose of assessing the reliability of the check valve.
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| The finding was a licensee performance deficiency of very low safety significance because it would not result in exceeding the Technical Specification limit for reactor coolant system leakage and would not have likely affected mitigation systems resulting in a loss of safety function. In addition, the finding did not represent an actual open pathway in the physical integrity of the reactor containment. Based on consultation and review with the Regional Senior Reactor Analyst, the inspectors concluded that the finding did not result in an increase in the likelihood of an initiating event such as an inter-system loss-of-coolant accident or a containment bypass event because the redundant isolation valve and closed loop system piping passed leak rate measurement test during the refueling outage with considerable margin. The inspectors concluded that this finding affected the cross cutting area of human performance.
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| Specifically, the licensee did not have adequately trained and knowledgeable personnel available to correctly evaluate the cause of the initial failed leak rate measurement test and to ensure that appropriate actions to correct the test sequence in the procedure were identified.
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| (IMC 0310,H.2(b))
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| Inspection Report# : 2012002 (pdf)
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| Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Security Although the Security Cornerstone is included in the Reactor Oversight Process assessment program, the Commission has decided that specific information related to findings and performance indicators pertaining to the Security Cornerstone will not be publicly available to ensure that security information is not provided to a possible adversary.
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| Other than the fact that a finding or performance indicator is Green or Greater-Than-Green, security related information will not be displayed on the public web page. Therefore, the cover letters to security inspection reports may be viewed.
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| Miscellaneous Last modified : February 28, 2013 Page 6 of 6
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| 1Q/2013 Inspection Findings - Clinton Clinton 1Q/2013 Plant Inspection Findings Initiating Events Significance: Dec 31, 2012 Identified By: Self-Revealing Item Type: FIN Finding FAILURE TO COMPLETE AN ADEQUATE EXTENT CONDITION REVIEW AND TO CORRECT A PREVIOUSLY IDENTIFIED DESIGN PROBLEM RESULTED IN A TRIP OF THE EMERGENCY RESERVE AUXILIARY TRANSFORMER A finding of very low safety significance was self-revealed when the emergency reserve auxiliary transformer (ERAT) tripped during troubleshooting activities to isolate a direct current system ground following heavy rainfall.
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| The ERAT trip occurred due to the presence of a latent design error identified on seal-in relays in the ERAT's control circuitry and the licensee's failure to adequately evaluate and correct it during its extent of condition review of the problem after it was identified in September 2002. The licensee restored the ERAT to service and implemented a modification to correct the latent design problem. Because the ERAT is not safety related, no violation of regulatory requirements was identified.
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| The finding was of more than minor safety significance because it was sufficiently similar to several examples in Inspection Manual Chapter 0612, "Power Reactor Inspection Reports," Appendix E, "Examples of Minor Issues,"
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| wherein licnesees failed to adequately correct conditions adverse to quality and the consequences had some safety impact. The performance deficiency was also associated with the Equipment Performance attribute and adversely affected the Initiating Events Cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, when the ERAT tripped safety related 4160 volt bus 1A1, which had been powered by the ERAT, momentarily lost power. With the momentary loss of power several plant safety systems were affected including a loss of secondary containment differential pressure. The finding was a licensee performance deficiency of very low safety significance because it: (1) did not involve a loss-of-coolant accident initiator; (2) did not cause a reactor trip AND the loss of mitigation equipment; (3) did not involve the complete or partial loss of a support system that contributes to the likelihood of, or cause, an initiating event AND affect mitigation equipment; and (4) did not increase the frequency of a fire or internal flooding initiating event. While the finding did involve a partial loss of a support system (i.e., offsite power) that contributes to the likelihood of an initiating event, mitigation equipment was not adversely affected by the momentary loss of power. The inspectors concluded that because the licensee's missed opportunity to correct the latent design error occurred in 2002 and no other more recent opportunities reasonably existed to identify and correct the problem, this issue would not be reflective of current licensee performance and no cross-cutting aspect was identified.
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| Inspection Report# : 2012005 (pdf)
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| Mitigating Systems Significance: Mar 31, 2013 Identified By: NRC Item Type: NCV NonCited Violation Page 1 of 7
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| 1Q/2013 Inspection Findings - Clinton FAILURE TO PERFORM ADEQUATE MOV PREVENTATIVE MAINTENANCE RESULTED IN INOPERABLE RCIC SYSTEM A finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criteria V, "Instructions, Procedures, and Drawings" was self-revealed when safety-related motor operated valve 1E51-F031, reactor core isolation cooling (RCIC) system suppression pool suction valve, failed to fully close during surveillance testing on October 29, 2012. The valve failure occured due to the licensee's failure to establish an adequate procedure to perform preventive maintenance on it. Specifically, the maintenance procedure did not contain a requirement to stroke a motor operated valve during the performance of periodic stem lubrication activities. The licensee entered this issue into its corrective action program for evaluation and initiated corrective actions to revise the maintenance procedure.
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| The finding was of more than minor significance since it was associated with the Procedure Quality attribute and adversely affected the Mitigation Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the valve failure rendered the RCIC system inoperable. This finding is of very low safety significance because it: (1) was not a design or qualification deficiency; (2) did not represent an actual loss of function of a system; (3) did not represent an actual loss of function of a single train or two seprate trains for greater than its Technical Specification (TS) allowed outage time; (4) did not represent an actual loss of function of one or more non-TS trains of equipment designated as high safety significant; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors concluded that this finding affected the cross-cutting area of human performance since adequate licensee resources involving personnel and procedures did not support successful human performance. Specifically, the maintenance procedure was not appropriate to the circumstances because it did not contain adequate instructions to ensure that motor operated valve stems were adequately lubricated. (IMC 0310 H.2 (c))
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| Inspection Report# : 2013002 (pdf)
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| Significance: Mar 31, 2013 Identified By: NRC Item Type: FIN Finding FAILURE TO PERFORM ADEQUATE PAST OPERABILITY EVALUATION The inspectors identified a finding of very low safety significance associated with the licensee's failure to correctly evaluate the past operability of safety-related motor operator valve 1E51-F031, reactor core isolation cooling system suppression pool suction valve, which failed quarterly surveillance testing on October 29, 2012. No violation of regulatory requirements was identified. The licensee entered this issue into its corrective action program for evaluation and initiated corrective actions to revise the past operability evaluation.
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| The finding was of more than minor significance since the failure to correctly evaluate a degraded/nonconforming condition potentially affecting the operability of structures, systems, and components (SSC) required to be operable by Technical specification (TS) would become a more significant safety concern, if left uncorrected, because it could reasonably result in an unrecognized condition of an SSC failing to fulfill a safety-related function. The finding was a licensee performance deficiency of very low safety significance because it: (1) was not a design or qualification deficiency; (2) did not represent an actual loss of function of a system; (3) did not represent an actual loss of function of a single train or two separate trains for greater than its TS allowed outage time; (4) did not represent an actual loss of function of one or more non-TS trains of equipment designated as high safety significant; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors concluded that this finding affected the cross-cutting area of human performance. Specifically, the licensee failed to use conservative assumptions in decision making while evaluating past operability of the valve by assuming that the time of inoperability was the same as the time of discovery for a time dependent failure mechanism (i.e., hardened grease) since no firm evidence to support operability was obtained by testing. (IMC 0310 H.1(b))
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| Inspection Report# : 2013002 (pdf)
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| Page 2 of 7
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| 1Q/2013 Inspection Findings - Clinton Significance: Dec 31, 2012 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO SATISFY 10 CFR 50.73 REPORTING REQUIREMENTS FOR A CONDITION PROHIBITED BY TECHNICAL SPECIFICATIONS.
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| The inspectors identified a finding of very low safety significance (Green) with an associated Severity Level IV Non-Cited Vilation of the NRC's reporting requirements in 10 CFR 50.73, "Licensee Event Report System." The licensee failed to submit a required Licensee Event Report (LER) within 60 days after the discovery of an event that was reportable in accordance with 10 CFR 50.73(a)(2)(i)(B) as a condition which was prohibited by the plant's Technical Specifications (TS) and 10 CFR 50.73(a)(2)(v)(B) as a condition that could have prevented the fulfillment of a safety function. The condition involved an inoperable diesel generator (DG) for longer than the TS completion time for restoration. The licensee subsequently submitted the required LER.
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| Because this violation of the NRC's reporting requirements affected the NRC's ability to perform its regulatory function, the inspectors evaluated the violation using the traditional enforcement process in accordance with the NRC Enforcement Policy and assessed the significance of the underlying issue using the Significance Determination Process. The finding was of more than minor significance because the NRC relies on licensees to identify and report conditions or events meeting the criteria specified in the TS and the regulations in order to perform its regulatory function and, therefore if left uncorrected it could lead to a more significant safety concern. The inspectors previously determined that the underlying issue (i.e., the failure to correctly assembly a DG ventilation system damper that resulted in an inoperable DG) was a finding of very low safety significance during a detailed risk evaluation.
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| Consistent with the guidance in Section 6.9, Paragraph d.9, of the NRC Enforcement Policy, the violation associated with this finding was determined to be a Severity Level IV Violation. This finding affected the cross-cutting area of human performance. Specifically, the licnesee's decision making process while evaluating the reportability of the condition with respect to the reporting requirements in 10 CFR 50.73 was inadequate. IMC 0310 H.1(a))
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| Inspection Report# : 2012005 (pdf)
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| Significance: Sep 30, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation FAILURE TO CORRECTLY ASSEMBLE DIESEL GENERATOR VENTILATION SYSTEM DAMPER RESULTED IN INOPERABLE DIESEL GENERATOR A finding of very low safety significance with an associated non-cited violation of 10 CFR 50, Appendix B, Criteria V, "Instructions, Procedures, and Drawings" was self-revealed on March 1, 2012 when the Division 1 diesel generator (DG) ventilation system supply damper was discovered failed closed with the ventilation supply fan running during a Division 1 DG surveillance test. The damper failure occurred due to the licensee's failure to establish an adequate procedure to perform maintenance. Specifically, the maintenance procedure did not contain an appropriate verification step to ensure that locknuts on the damper hydramotor coupling were tightly fastened. As a result, vibration of the coupling during operation over time caused the coupling to separate such that the damper would not open. The licensee entered this issue into its corrective action program for evaluation, repaired the damper, and initiated corrective actions to revise the maintenance procedure.
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| The finding was of more than minor significance since it was associated with the Procedure Quality attribute and adversely affected the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the damper failure rendered the Division 1 DG inoperable. Although the finding involved an actual loss of function of a single train for greater than its Technical Specification allowed outage time, it was determined to be of very low safety significance during a detailed quantitative Significance Determination Process review since the delta core damage frequency and delta large early release frequency were both determined to be negligible based upon crediting operator Page 3 of 7
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| 1Q/2013 Inspection Findings - Clinton recovery actions to restore DG room ventilation. The inspectors condluded that this finding affected the cross-cutting area of human performance since adequate licensee resources involving personnel and procedures did not support successful human performance. Specifically, the maintenance procedure did not contain adequate instruction to ensure that locknuts on the damper hydramotor coupling were tightly fastened. (IMC 0310, H.2(a))
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| Inspection Report# : 2012004 (pdf)
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| Significance: Sep 30, 2012 Identified By: NRC Item Type: FIN Finding FAILURE TO PERFORM ADEQUATE PAST OPERABILITY EVALUATIONS FOR EMERGENCY CORE COOLING SYSTEM RELIEF VALVES The inspectors identified a finding of very low safety significance associated with the licensee's failure to correctly evaluate the past operability of two emergency core cooling system (ECCS) relief valves that failed bench testing following replacement during the C1R13 refueling outage. No violation of regulatory requirements was identified because revised evaluations by the licensee determined that the valves would have satisfied their safety functions. The licensee entered this issue into its corrective action program for evaluation and initated corrective actions to revise the past operability evaluations to correct gross errors in the original evaluations.
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| The finding was of more than minor significance since the failure to correctly evaluate a degraded/nonconforming condition potentially affecting the operability of structures, systems, and components (SSC) required to be operable by Technical Specifications (TS) would become a more significant safety concern if left uncorrected because it could reasonably result in an unrecognized condition of an SSC failing to fulfill a safety-related function. The finding was a licensee performance deficiency of very low safety significance because it: (1) was not a design or qualification deficiency; (2) did not represent an actual loss of function of a system; (3) did not represent an actual loss of function of a single train or two separate trains for greater than its TS allowed outage time; (4) did not represent an actual loss of function of one or more non-TS trains of equipment designated as high safety significant; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors concluded that this finding affected the cross-cutting area of human performance since licensee engineering staff failed to thoroughly and correctly evaluate past operability of the two ECCS relief valves due to inattention to detail. Human error prevention techniques were not appropriately employed to support human performance. The most significant concerns were that the independent technical reviewer did not independently validate information contained in the past operability evaluations by reviewing the valve test records and, that neither the independent technical reviewer nor the engineering supervisory reviewer challenged the unwarranted past operability conclusion reached for the 1E12-F025C test failure. (IMC 0310,H.4 (a)
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| Inspection Report# : 2012004 (pdf)
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| Significance: Jun 30, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation FAILURE TO ESTABLISH INSTRUCTIONS APPROPRIATE FOR INSTALLATION OF SHUTDOWN AND UPSET LEVEL INSTRUMENT REFERENCE LEG PIPE A finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was self-revealed on December 18, 2011, when an automatic reactor scram signal and loss of decay heat removal occurred due to low reactor pressure vessel (RPV) water level while lowering water level to a target level following an RPV hydrostatic pressure test. The licensee failed to establish an adequate procedure to perfrom reinstallation of common shutdown and upset level instrument reference leg piping.
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| Specifically, inadequacies with the procedure resulted in improper filling and venting of the reference leg piping causing inaccurate indication of RPV level - an error of approximately 108 inches. In addition, the licensee failed to use appropriate acceptance criteria when accepting that the instrument restoration activities had been successfully Page 4 of 7
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| 1Q/2013 Inspection Findings - Clinton acomplished. The licensee entered this issue into its corrective action program for evaluation and initiated corrective actions to revise procedures to more rigorously control the evolution and to train personnel.
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| The finding was of more than minor significance since it was associated with the Mitigating Systems cornerstone attribute of Procedure Quality and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to establish procedures adequate to maintain correct indication of RPV water level upon the reinstallation of permanent shutdown and upset level instrument reference leg piping. The finding was determined to be a licensee performance deficiency of very low safety significance based upon a Phase 3 Significance Determination Process evaluation by the Regional Senior Reactor Analyst with a risk result of approximately 4E-7 for Core Damage Frequency and no Large Early Release Frequency contribution since the event occurred more than 8 days from the beginning of the refueling outage. The inspectors concluded that this finding affected the cross cutting area of human performance. Specifically, in the area of work control, the licensee did not ensure that personnel, equipment, procedures, and other resources were available and adequate. Complete, accurate, and up-to-date procedures and work packages were not available to ensure nuclear safety (IMC 0310 H.2(c))
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| Inspection Report# : 2012003 (pdf)
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| Significance: Jun 30, 2012 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ENSURE TORNADO MISSLE PROTECTION FOR SAFETY RELATED COMPONENTS The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion III, "Design Control," when permanently installed tornado missile barrier protection was removed without adequate provisions to assure that appropriate quality standards were specified and included in design documents and that deviation from such standards was controlled. The licensee failed to ensure tornado missile protection for safety related components prior to and during maintenance affecting Control Room Ventilation (VC)
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| Train 'A'. Specifically, when the permanent missile barrier was removed, the licensee failed to ensure protection for two safety related radiation monitors, 1RIX-PR009C and 1RIX-PR009D and did not satisy requirements in modification documents for protection of VC panel 0PL72JA. The licensee entered this issue into its corrective action program for evaluation and performed immediate corrective actions to resolve the design deficiencies at the time of identification.
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| The finding was of more than minor significance because it was sufficiently similar to Inspection Manual Chapter 0612, "Power Reactor Inspection Reports," Appendix E, "Examples of Minor Issues," Example 3(a) in that this modification was found to contain errors significant enough that the modification required rework to correctly resolve design basis tornado concerns. The performance deficiency was also associated with the Mitigation Systems cornerstone attribute of Equipment Performance, and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to protect safety related components during work activities that modified the installed missile barrier required by the Clinton Power Station design. The finding was a licensee performance deficiency of very low safety significance because the design deficiency was confirmed to not result in an actual loss of operability or functionality. The inspectors concluded that the finding affected the cross cutting area of human performance. Specifically, in the area of work control, the licencee did not appropriately plan work activities by incorporating job site conditions and the need for adequate planned contingencies. (IMC 0310 H.3(a))
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| Inspection Report# : 2012003 (pdf)
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| Barrier Integrity Page 5 of 7
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| 1Q/2013 Inspection Findings - Clinton Significance: Dec 31, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation FAILURE TO PERFORM PREVENTIVE MAINTENANCE ON STANDBY GAS TREATMENT SYSTEM RELAY 0UAY-VG506D.
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| A finding of very low safety significance with an associated Non-Cited Violation of Technical Specification (TS) 5.4.1.a. was self-revealed when the age-related failure of Standy Gas Treatment (VG) system relay 0UAY-VG506D caused the removal of VG Train A electric heater 0VG04AA from operation, an entry into TS 3.5.4.3 due to the inoperability of VG Train A, and an unplanned on-line plant risk condition increase from Green to Yellow. The relay failure occurred due to the licensee's failure to perform any replacement preventive maintenance on the component throughout the history of plant operation. During two separate independent reviews performed by the licensee on July 15, 2011, and on August 24, 2011, the licensee failed to correctly classify the component in accordance with its preventive maintenance procedure. This resulted in no replacement maintenance activity ever being performed for the relay and its eventual failure on August 22, 2012. The licensee initiated corrective actions to replace the relay and put in place the appropriate preventive maintenance actions.
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| The finding was of more than minor safety significance because it was sufficiently similar to several examples in Inspection Manual Chapter 0612, "Power Reactor Inspection Reports," Appendix E, "Examples of Minor Issues,"
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| wherein licensees failed to adequately implement procedural requirements and the consequences had some safety impact. The performance deficiency was also associated with the SSC [Systems, Structures, and Components] and Barrier Performance attribute and adversely affected the Barrier Integrity Cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events.
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| Specifically, the age-related failure of 0UAY-VG506D on August 22, 2012 rendered VG Train A inoperable and caused an unplanned increase in the plant's on-line risk condition from Green to Yellow. The finding was a licensee performance deficiency of very low safety significance because it only represented a degradation of the radiological barrier function provided for the Auxiliary Building and the Fuel Building and was not a complete loss of the barrier function provided by the VG system since VG Train B remained operable. This finding affected the cross-cutting area of human performance. Specifically, in the area of work control, the licensee did not appropriately coordinate work activities by incorporating actions to plan work activities to support longterm equipment reliability by scheduling maintenance as more preventive than reactive. (IMC 0310, H.3(b))
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| Inspection Report# : 2012005 (pdf)
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| Emergency Preparedness Occupational Radiation Safety Significance: Mar 31, 2013 Identified By: NRC Item Type: NCV NonCited Violation INCOMPLETE ED DOSE RATE ALARM EVALUATION The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 20.1501(a) for the failure to perform surveys to ensure compliance with 10 CFR 20.1201 shallow-dose equivalent (SDE) limits for five individuals during the fourth quarter 2011 due to contamination build-up on the workers' gloves.
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| This issue was entered into the licensee's corrective action program as AR 01335298 and AR 01454976. Corrective Page 6 of 7
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| 1Q/2013 Inspection Findings - Clinton actions include performing an apparent cause evaluation and performing dose assessments for the individuals involved.
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| The performance deficiency was determined to be of more than minor safety significance in accordance with IMC 0612, Appendix B, "Issue Screening," because it was associated with the Program And Process Attribute of the Occupational Radiation Safety Cornerstone and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation, in that not performing an adequate SDE assessment affected the licensee's ability to monitor, control, and limit radiation exposures. The inspectors also reviewed the guidance in IMC 0612, Appendix E, "Examples of Minor issues," and did not find any similar examples. In accordance with IMC 0609, Appendix C, "Occupational Radiation Safety Significance Determination Process," the inspectors determined that the finding had very low safety significance because the finding did not involve: (1)
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| ALARA planning and controls, (2) a radiological overexposure, (3) a substantial potential for an overexposure, or (4) a compromised ability to assess dose. The primary cause of this finding was related to the cross-cutting aspect of human performance with the component of work practices. The specific aspect was that the licensee ensures supervisory and management oversight of work activities, including contractors, such that nuclear safety is supported (IMC 0310 H.4(c))
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| Inspection Report# : 2013002 (pdf)
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| Public Radiation Safety Security Although the Security Cornerstone is included in the Reactor Oversight Process assessment program, the Commission has decided that specific information related to findings and performance indicators pertaining to the Security Cornerstone will not be publicly available to ensure that security information is not provided to a possible adversary.
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| Other than the fact that a finding or performance indicator is Green or Greater-Than-Green, security related information will not be displayed on the public web page. Therefore, the cover letters to security inspection reports may be viewed.
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| Miscellaneous Last modified : June 04, 2013 Page 7 of 7
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| 2Q/2013 Inspection Findings - Clinton Clinton 2Q/2013 Plant Inspection Findings Initiating Events Significance: Dec 31, 2012 Identified By: Self-Revealing Item Type: FIN Finding FAILURE TO COMPLETE AN ADEQUATE EXTENT CONDITION REVIEW AND TO CORRECT A PREVIOUSLY IDENTIFIED DESIGN PROBLEM RESULTED IN A TRIP OF THE EMERGENCY RESERVE AUXILIARY TRANSFORMER A finding of very low safety significance was self-revealed when the emergency reserve auxiliary transformer (ERAT) tripped during troubleshooting activities to isolate a direct current system ground following heavy rainfall.
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| The ERAT trip occurred due to the presence of a latent design error identified on seal-in relays in the ERAT's control circuitry and the licensee's failure to adequately evaluate and correct it during its extent of condition review of the problem after it was identified in September 2002. The licensee restored the ERAT to service and implemented a modification to correct the latent design problem. Because the ERAT is not safety related, no violation of regulatory requirements was identified.
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| The finding was of more than minor safety significance because it was sufficiently similar to several examples in Inspection Manual Chapter 0612, "Power Reactor Inspection Reports," Appendix E, "Examples of Minor Issues,"
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| wherein licnesees failed to adequately correct conditions adverse to quality and the consequences had some safety impact. The performance deficiency was also associated with the Equipment Performance attribute and adversely affected the Initiating Events Cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, when the ERAT tripped safety related 4160 volt bus 1A1, which had been powered by the ERAT, momentarily lost power. With the momentary loss of power several plant safety systems were affected including a loss of secondary containment differential pressure. The finding was a licensee performance deficiency of very low safety significance because it: (1) did not involve a loss-of-coolant accident initiator; (2) did not cause a reactor trip AND the loss of mitigation equipment; (3) did not involve the complete or partial loss of a support system that contributes to the likelihood of, or cause, an initiating event AND affect mitigation equipment; and (4) did not increase the frequency of a fire or internal flooding initiating event. While the finding did involve a partial loss of a support system (i.e., offsite power) that contributes to the likelihood of an initiating event, mitigation equipment was not adversely affected by the momentary loss of power. The inspectors concluded that because the licensee's missed opportunity to correct the latent design error occurred in 2002 and no other more recent opportunities reasonably existed to identify and correct the problem, this issue would not be reflective of current licensee performance and no cross-cutting aspect was identified.
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| Inspection Report# : 2012005 (pdf)
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| Mitigating Systems Significance: Mar 31, 2013 Identified By: NRC Item Type: NCV NonCited Violation Page 1 of 7
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| 2Q/2013 Inspection Findings - Clinton FAILURE TO PERFORM ADEQUATE MOV PREVENTATIVE MAINTENANCE RESULTED IN INOPERABLE RCIC SYSTEM A finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criteria V, "Instructions, Procedures, and Drawings" was self-revealed when safety-related motor operated valve 1E51-F031, reactor core isolation cooling (RCIC) system suppression pool suction valve, failed to fully close during surveillance testing on October 29, 2012. The valve failure occured due to the licensee's failure to establish an adequate procedure to perform preventive maintenance on it. Specifically, the maintenance procedure did not contain a requirement to stroke a motor operated valve during the performance of periodic stem lubrication activities. The licensee entered this issue into its corrective action program for evaluation and initiated corrective actions to revise the maintenance procedure.
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| The finding was of more than minor significance since it was associated with the Procedure Quality attribute and adversely affected the Mitigation Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the valve failure rendered the RCIC system inoperable. This finding is of very low safety significance because it: (1) was not a design or qualification deficiency; (2) did not represent an actual loss of function of a system; (3) did not represent an actual loss of function of a single train or two seprate trains for greater than its Technical Specification (TS) allowed outage time; (4) did not represent an actual loss of function of one or more non-TS trains of equipment designated as high safety significant; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors concluded that this finding affected the cross-cutting area of human performance since adequate licensee resources involving personnel and procedures did not support successful human performance. Specifically, the maintenance procedure was not appropriate to the circumstances because it did not contain adequate instructions to ensure that motor operated valve stems were adequately lubricated. (IMC 0310 H.2 (c))
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| Inspection Report# : 2013002 (pdf)
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| Significance: Mar 31, 2013 Identified By: NRC Item Type: FIN Finding FAILURE TO PERFORM ADEQUATE PAST OPERABILITY EVALUATION The inspectors identified a finding of very low safety significance associated with the licensee's failure to correctly evaluate the past operability of safety-related motor operator valve 1E51-F031, reactor core isolation cooling system suppression pool suction valve, which failed quarterly surveillance testing on October 29, 2012. No violation of regulatory requirements was identified. The licensee entered this issue into its corrective action program for evaluation and initiated corrective actions to revise the past operability evaluation.
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| The finding was of more than minor significance since the failure to correctly evaluate a degraded/nonconforming condition potentially affecting the operability of structures, systems, and components (SSC) required to be operable by Technical specification (TS) would become a more significant safety concern, if left uncorrected, because it could reasonably result in an unrecognized condition of an SSC failing to fulfill a safety-related function. The finding was a licensee performance deficiency of very low safety significance because it: (1) was not a design or qualification deficiency; (2) did not represent an actual loss of function of a system; (3) did not represent an actual loss of function of a single train or two separate trains for greater than its TS allowed outage time; (4) did not represent an actual loss of function of one or more non-TS trains of equipment designated as high safety significant; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors concluded that this finding affected the cross-cutting area of human performance. Specifically, the licensee failed to use conservative assumptions in decision making while evaluating past operability of the valve by assuming that the time of inoperability was the same as the time of discovery for a time dependent failure mechanism (i.e., hardened grease) since no firm evidence to support operability was obtained by testing. (IMC 0310 H.1(b))
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| Inspection Report# : 2013002 (pdf)
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| Page 2 of 7
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| 2Q/2013 Inspection Findings - Clinton Significance: Dec 31, 2012 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO SATISFY 10 CFR 50.73 REPORTING REQUIREMENTS FOR A CONDITION PROHIBITED BY TECHNICAL SPECIFICATIONS.
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| The inspectors identified a finding of very low safety significance (Green) with an associated Severity Level IV Non-Cited Vilation of the NRC's reporting requirements in 10 CFR 50.73, "Licensee Event Report System." The licensee failed to submit a required Licensee Event Report (LER) within 60 days after the discovery of an event that was reportable in accordance with 10 CFR 50.73(a)(2)(i)(B) as a condition which was prohibited by the plant's Technical Specifications (TS) and 10 CFR 50.73(a)(2)(v)(B) as a condition that could have prevented the fulfillment of a safety function. The condition involved an inoperable diesel generator (DG) for longer than the TS completion time for restoration. The licensee subsequently submitted the required LER.
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| Because this violation of the NRC's reporting requirements affected the NRC's ability to perform its regulatory function, the inspectors evaluated the violation using the traditional enforcement process in accordance with the NRC Enforcement Policy and assessed the significance of the underlying issue using the Significance Determination Process. The finding was of more than minor significance because the NRC relies on licensees to identify and report conditions or events meeting the criteria specified in the TS and the regulations in order to perform its regulatory function and, therefore if left uncorrected it could lead to a more significant safety concern. The inspectors previously determined that the underlying issue (i.e., the failure to correctly assembly a DG ventilation system damper that resulted in an inoperable DG) was a finding of very low safety significance during a detailed risk evaluation.
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| Consistent with the guidance in Section 6.9, Paragraph d.9, of the NRC Enforcement Policy, the violation associated with this finding was determined to be a Severity Level IV Violation. This finding affected the cross-cutting area of human performance. Specifically, the licnesee's decision making process while evaluating the reportability of the condition with respect to the reporting requirements in 10 CFR 50.73 was inadequate. IMC 0310 H.1(a))
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| Inspection Report# : 2012005 (pdf)
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| Significance: Sep 30, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation FAILURE TO CORRECTLY ASSEMBLE DIESEL GENERATOR VENTILATION SYSTEM DAMPER RESULTED IN INOPERABLE DIESEL GENERATOR A finding of very low safety significance with an associated non-cited violation of 10 CFR 50, Appendix B, Criteria V, "Instructions, Procedures, and Drawings" was self-revealed on March 1, 2012 when the Division 1 diesel generator (DG) ventilation system supply damper was discovered failed closed with the ventilation supply fan running during a Division 1 DG surveillance test. The damper failure occurred due to the licensee's failure to establish an adequate procedure to perform maintenance. Specifically, the maintenance procedure did not contain an appropriate verification step to ensure that locknuts on the damper hydramotor coupling were tightly fastened. As a result, vibration of the coupling during operation over time caused the coupling to separate such that the damper would not open. The licensee entered this issue into its corrective action program for evaluation, repaired the damper, and initiated corrective actions to revise the maintenance procedure.
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| The finding was of more than minor significance since it was associated with the Procedure Quality attribute and adversely affected the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the damper failure rendered the Division 1 DG inoperable. Although the finding involved an actual loss of function of a single train for greater than its Technical Specification allowed outage time, it was determined to be of very low safety significance during a detailed quantitative Significance Determination Process review since the delta core damage frequency and delta large early release frequency were both determined to be negligible based upon crediting operator Page 3 of 7
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| 2Q/2013 Inspection Findings - Clinton recovery actions to restore DG room ventilation. The inspectors condluded that this finding affected the cross-cutting area of human performance since adequate licensee resources involving personnel and procedures did not support successful human performance. Specifically, the maintenance procedure did not contain adequate instruction to ensure that locknuts on the damper hydramotor coupling were tightly fastened. (IMC 0310, H.2(a))
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| Inspection Report# : 2012004 (pdf)
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| Significance: Sep 30, 2012 Identified By: NRC Item Type: FIN Finding FAILURE TO PERFORM ADEQUATE PAST OPERABILITY EVALUATIONS FOR EMERGENCY CORE COOLING SYSTEM RELIEF VALVES The inspectors identified a finding of very low safety significance associated with the licensee's failure to correctly evaluate the past operability of two emergency core cooling system (ECCS) relief valves that failed bench testing following replacement during the C1R13 refueling outage. No violation of regulatory requirements was identified because revised evaluations by the licensee determined that the valves would have satisfied their safety functions. The licensee entered this issue into its corrective action program for evaluation and initated corrective actions to revise the past operability evaluations to correct gross errors in the original evaluations.
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| The finding was of more than minor significance since the failure to correctly evaluate a degraded/nonconforming condition potentially affecting the operability of structures, systems, and components (SSC) required to be operable by Technical Specifications (TS) would become a more significant safety concern if left uncorrected because it could reasonably result in an unrecognized condition of an SSC failing to fulfill a safety-related function. The finding was a licensee performance deficiency of very low safety significance because it: (1) was not a design or qualification deficiency; (2) did not represent an actual loss of function of a system; (3) did not represent an actual loss of function of a single train or two separate trains for greater than its TS allowed outage time; (4) did not represent an actual loss of function of one or more non-TS trains of equipment designated as high safety significant; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors concluded that this finding affected the cross-cutting area of human performance since licensee engineering staff failed to thoroughly and correctly evaluate past operability of the two ECCS relief valves due to inattention to detail. Human error prevention techniques were not appropriately employed to support human performance. The most significant concerns were that the independent technical reviewer did not independently validate information contained in the past operability evaluations by reviewing the valve test records and, that neither the independent technical reviewer nor the engineering supervisory reviewer challenged the unwarranted past operability conclusion reached for the 1E12-F025C test failure. (IMC 0310,H.4 (a)
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| Inspection Report# : 2012004 (pdf)
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| Barrier Integrity Significance: Dec 31, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation FAILURE TO PERFORM PREVENTIVE MAINTENANCE ON STANDBY GAS TREATMENT SYSTEM RELAY 0UAY-VG506D.
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| A finding of very low safety significance with an associated Non-Cited Violation of Technical Specification (TS) 5.4.1.a. was self-revealed when the age-related failure of Standy Gas Treatment (VG) system relay 0UAY-VG506D caused the removal of VG Train A electric heater 0VG04AA from operation, an entry into TS 3.5.4.3 due to the Page 4 of 7
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| 2Q/2013 Inspection Findings - Clinton inoperability of VG Train A, and an unplanned on-line plant risk condition increase from Green to Yellow. The relay failure occurred due to the licensee's failure to perform any replacement preventive maintenance on the component throughout the history of plant operation. During two separate independent reviews performed by the licensee on July 15, 2011, and on August 24, 2011, the licensee failed to correctly classify the component in accordance with its preventive maintenance procedure. This resulted in no replacement maintenance activity ever being performed for the relay and its eventual failure on August 22, 2012. The licensee initiated corrective actions to replace the relay and put in place the appropriate preventive maintenance actions.
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| The finding was of more than minor safety significance because it was sufficiently similar to several examples in Inspection Manual Chapter 0612, "Power Reactor Inspection Reports," Appendix E, "Examples of Minor Issues,"
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| wherein licensees failed to adequately implement procedural requirements and the consequences had some safety impact. The performance deficiency was also associated with the SSC [Systems, Structures, and Components] and Barrier Performance attribute and adversely affected the Barrier Integrity Cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events.
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| Specifically, the age-related failure of 0UAY-VG506D on August 22, 2012 rendered VG Train A inoperable and caused an unplanned increase in the plant's on-line risk condition from Green to Yellow. The finding was a licensee performance deficiency of very low safety significance because it only represented a degradation of the radiological barrier function provided for the Auxiliary Building and the Fuel Building and was not a complete loss of the barrier function provided by the VG system since VG Train B remained operable. This finding affected the cross-cutting area of human performance. Specifically, in the area of work control, the licensee did not appropriately coordinate work activities by incorporating actions to plan work activities to support longterm equipment reliability by scheduling maintenance as more preventive than reactive. (IMC 0310, H.3(b))
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| Inspection Report# : 2012005 (pdf)
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| Emergency Preparedness Occupational Radiation Safety Significance: Jun 30, 2013 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW PROCEDURES RESULTED IN THE UNPLANNED INTAKE OF RADIOACTIVE MATERIAL BY FIVE WORKERS.
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| A self-revealing finding of very low safety significance (Green) and associated Non-Cited Violation of Technical Specification 5.4.1.a for the failure to follow procedures associated with the Radiation Work Permit (RWP) on March 28, 2013. The issue resulted in the unplanned intake of radioactive material by five workers. RWP 10014553, "2013 RW HRA/LHRA," Revision 0, established the requirement for the usage of high efficiency particulate air vacuums during the cleanup of a legacy radioactive resin spill. The licensee replaced this cleanup method with manual resin removal during the cleanup contrary to the conditions set in the RWP. This is a performance deficiency, which was within the licensee's ability to foresee and should have been prevented. The issue was entered into the licensee's corrective action program as Action Request 01494203. The licensee completed actions to ensure worker compliance with radiation protection program procedures.
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| The performance deficiency was determined to be more than minor safety significance in accordance with Inspection Manual Chapter (IMC) 0612, Appendix B, "Issue Screening," because it was associated with the program and process Page 5 of 7
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| 2Q/2013 Inspection Findings - Clinton attribute of the Occupational Radiation Safety Cornerstone and affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation, in that, the workers received additonal and unplanned dose from the intake of radioactive materials, The significance was determined in accordance with IMC 0609, Appendix C, "Occupational Radiation Safety Significance Determination Process." The inspectors determined the finding has very low safety significance (Green) because the finding did not involve: (1) As Low As Reasonably Achievable (ALARA) planning or work controls involving excessive occupational collective dose, (2) an overexposure, (3) a substantial potential for overexposure, or (4) compromised ability to assess dose. The primary cause of this finding was related to the cross-cutting aspect of human performance with the component of decision making. The licensee failed to use conservative assumptions in decision making and failed to adopt a requirment to demonstrate that the proposed action is safe in order to proceed. H. 1(b).
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| Inspection Report# : 2013003 (pdf)
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| Significance: Mar 31, 2013 Identified By: NRC Item Type: NCV NonCited Violation INCOMPLETE ED DOSE RATE ALARM EVALUATION The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 20.1501(a) for the failure to perform surveys to ensure compliance with 10 CFR 20.1201 shallow-dose equivalent (SDE) limits for five individuals during the fourth quarter 2011 due to contamination build-up on the workers' gloves.
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| This issue was entered into the licensee's corrective action program as AR 01335298 and AR 01454976. Corrective actions include performing an apparent cause evaluation and performing dose assessments for the individuals involved.
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| The performance deficiency was determined to be of more than minor safety significance in accordance with IMC 0612, Appendix B, "Issue Screening," because it was associated with the Program And Process Attribute of the Occupational Radiation Safety Cornerstone and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation, in that not performing an adequate SDE assessment affected the licensee's ability to monitor, control, and limit radiation exposures. The inspectors also reviewed the guidance in IMC 0612, Appendix E, "Examples of Minor issues," and did not find any similar examples. In accordance with IMC 0609, Appendix C, "Occupational Radiation Safety Significance Determination Process," the inspectors determined that the finding had very low safety significance because the finding did not involve: (1)
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| ALARA planning and controls, (2) a radiological overexposure, (3) a substantial potential for an overexposure, or (4) a compromised ability to assess dose. The primary cause of this finding was related to the cross-cutting aspect of human performance with the component of work practices. The specific aspect was that the licensee ensures supervisory and management oversight of work activities, including contractors, such that nuclear safety is supported (IMC 0310 H.4(c))
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| Inspection Report# : 2013002 (pdf)
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| Public Radiation Safety Security Although the Security Cornerstone is included in the Reactor Oversight Process assessment program, the Commission has decided that specific information related to findings and performance indicators pertaining to the Security Cornerstone will not be publicly available to ensure that security information is not provided to a possible adversary.
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| Page 6 of 7
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| 2Q/2013 Inspection Findings - Clinton Other than the fact that a finding or performance indicator is Green or Greater-Than-Green, security related information will not be displayed on the public web page. Therefore, the cover letters to security inspection reports may be viewed.
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| Miscellaneous Last modified : September 03, 2013 Page 7 of 7
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| 3Q/2013 Inspection Findings - Clinton Clinton 3Q/2013 Plant Inspection Findings Initiating Events Significance: Dec 31, 2012 Identified By: Self-Revealing Item Type: FIN Finding FAILURE TO COMPLETE AN ADEQUATE EXTENT CONDITION REVIEW AND TO CORRECT A PREVIOUSLY IDENTIFIED DESIGN PROBLEM RESULTED IN A TRIP OF THE EMERGENCY RESERVE AUXILIARY TRANSFORMER A finding of very low safety significance was self-revealed when the emergency reserve auxiliary transformer (ERAT) tripped during troubleshooting activities to isolate a direct current system ground following heavy rainfall.
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| The ERAT trip occurred due to the presence of a latent design error identified on seal-in relays in the ERAT's control circuitry and the licensee's failure to adequately evaluate and correct it during its extent of condition review of the problem after it was identified in September 2002. The licensee restored the ERAT to service and implemented a modification to correct the latent design problem. Because the ERAT is not safety related, no violation of regulatory requirements was identified.
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| The finding was of more than minor safety significance because it was sufficiently similar to several examples in Inspection Manual Chapter 0612, "Power Reactor Inspection Reports," Appendix E, "Examples of Minor Issues,"
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| wherein licnesees failed to adequately correct conditions adverse to quality and the consequences had some safety impact. The performance deficiency was also associated with the Equipment Performance attribute and adversely affected the Initiating Events Cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, when the ERAT tripped safety related 4160 volt bus 1A1, which had been powered by the ERAT, momentarily lost power. With the momentary loss of power several plant safety systems were affected including a loss of secondary containment differential pressure. The finding was a licensee performance deficiency of very low safety significance because it: (1) did not involve a loss-of-coolant accident initiator; (2) did not cause a reactor trip AND the loss of mitigation equipment; (3) did not involve the complete or partial loss of a support system that contributes to the likelihood of, or cause, an initiating event AND affect mitigation equipment; and (4) did not increase the frequency of a fire or internal flooding initiating event. While the finding did involve a partial loss of a support system (i.e., offsite power) that contributes to the likelihood of an initiating event, mitigation equipment was not adversely affected by the momentary loss of power. The inspectors concluded that because the licensee's missed opportunity to correct the latent design error occurred in 2002 and no other more recent opportunities reasonably existed to identify and correct the problem, this issue would not be reflective of current licensee performance and no cross-cutting aspect was identified.
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| Inspection Report# : 2012005 (pdf)
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| Mitigating Systems Significance: Sep 30, 2013 Identified By: NRC Item Type: NCV NonCited Violation Page 1 of 6
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| 3Q/2013 Inspection Findings - Clinton FAILURE TO FOLLOW PROCEDURE AND APPROPRIATELY DOCUMENT BASIS FOR IMMEDIATE OPERABILITY OF THE DIVISION 2 EMERGENCY DIESEL GENERATOR An NRC identified non-cited violation of 10CFR50, Appendix B, Criterion V, Instructions, Procedures and Drawings for the failure to follow procedure OP-AA-108-115, "Operability Determinations", Revision 11, and document the basis that a reasonable expectation of operability existed after an immediate operability determination. Specifically, after the control room received a report of a crack on the after cooler ducting of the Division 2 emergency diesel generator the licensee failed to document their basis that a reasonable expectation of operability existed for the Division 2 emergency diesel generator. The licensee documented this issue in the corrective action prgram as Action Request 015401540.
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| The inspectors determined that the licensee failing to follow the station procedure for operability determinations was a performance deficiency. Specifically, the licensee failed to follow the station procedure for operability determinations and appropriately document the decision and the basis that a reasonable expectation of operability existed for the Division 2 emergency diesel generator. The performance deficiency is more than minor because if immediate operability determination and either the basis that a reasonable expectation of operability exists or the declaration that the system, structure or component is inoperable is not appropriately documented it could lead to a more significant safety concern. Using Manual Chapter 0609, Attachment 4 "Initial Characterization of Findings," and Appendix A "The Significance Determination Process for Findings at Power" the finding was screened against the mitigating systems cornerstone and determined to be of very low safety significance (Green) because the finding was/did not: 1) a deficiency affecting the design or qualification of a mitigating structure, system or component, 2) represent a loss of system and/or function, 3) represent an actual loss of function of a single train for greater than its technical specification allowed outage time, 4) represent an actual loss of function of one or more non-technical specifications trains of equipment designated as high safety-significant for greater than 24 hours and 5) did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding or severe weather event.
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| The finding was determined to have a cross-cutting aspect in the area of human performance, associated with the decision making component, in that the licnesee decisions failed to demonstrate that nuclear safety is an overriding priority. Specifically, the licensee failed to use their systematic process, when faced with an unexpected plant condition of the Division 2 emergency diesel generator to ensure safety was maintained.. H.1(a).
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| Inspection Report# : 2013004 (pdf)
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| Significance: Mar 31, 2013 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PERFORM ADEQUATE MOV PREVENTATIVE MAINTENANCE RESULTED IN INOPERABLE RCIC SYSTEM A finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criteria V, "Instructions, Procedures, and Drawings" was self-revealed when safety-related motor operated valve 1E51-F031, reactor core isolation cooling (RCIC) system suppression pool suction valve, failed to fully close during surveillance testing on October 29, 2012. The valve failure occured due to the licensee's failure to establish an adequate procedure to perform preventive maintenance on it. Specifically, the maintenance procedure did not contain a requirement to stroke a motor operated valve during the performance of periodic stem lubrication activities. The licensee entered this issue into its corrective action program for evaluation and initiated corrective actions to revise the maintenance procedure.
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| The finding was of more than minor significance since it was associated with the Procedure Quality attribute and adversely affected the Mitigation Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the valve failure rendered the RCIC system inoperable. This finding is of very low safety significance because it: (1) was not a design or qualification deficiency; (2) did not represent an actual loss of function of a system; (3) did not Page 2 of 6
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| 3Q/2013 Inspection Findings - Clinton represent an actual loss of function of a single train or two seprate trains for greater than its Technical Specification (TS) allowed outage time; (4) did not represent an actual loss of function of one or more non-TS trains of equipment designated as high safety significant; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors concluded that this finding affected the cross-cutting area of human performance since adequate licensee resources involving personnel and procedures did not support successful human performance. Specifically, the maintenance procedure was not appropriate to the circumstances because it did not contain adequate instructions to ensure that motor operated valve stems were adequately lubricated. (IMC 0310 H.2 (c))
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| Inspection Report# : 2013002 (pdf)
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| Significance: Mar 31, 2013 Identified By: NRC Item Type: FIN Finding FAILURE TO PERFORM ADEQUATE PAST OPERABILITY EVALUATION The inspectors identified a finding of very low safety significance associated with the licensee's failure to correctly evaluate the past operability of safety-related motor operator valve 1E51-F031, reactor core isolation cooling system suppression pool suction valve, which failed quarterly surveillance testing on October 29, 2012. No violation of regulatory requirements was identified. The licensee entered this issue into its corrective action program for evaluation and initiated corrective actions to revise the past operability evaluation.
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| The finding was of more than minor significance since the failure to correctly evaluate a degraded/nonconforming condition potentially affecting the operability of structures, systems, and components (SSC) required to be operable by Technical specification (TS) would become a more significant safety concern, if left uncorrected, because it could reasonably result in an unrecognized condition of an SSC failing to fulfill a safety-related function. The finding was a licensee performance deficiency of very low safety significance because it: (1) was not a design or qualification deficiency; (2) did not represent an actual loss of function of a system; (3) did not represent an actual loss of function of a single train or two separate trains for greater than its TS allowed outage time; (4) did not represent an actual loss of function of one or more non-TS trains of equipment designated as high safety significant; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors concluded that this finding affected the cross-cutting area of human performance. Specifically, the licensee failed to use conservative assumptions in decision making while evaluating past operability of the valve by assuming that the time of inoperability was the same as the time of discovery for a time dependent failure mechanism (i.e., hardened grease) since no firm evidence to support operability was obtained by testing. (IMC 0310 H.1(b))
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| Inspection Report# : 2013002 (pdf)
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| Significance: Dec 31, 2012 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO SATISFY 10 CFR 50.73 REPORTING REQUIREMENTS FOR A CONDITION PROHIBITED BY TECHNICAL SPECIFICATIONS.
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| The inspectors identified a finding of very low safety significance (Green) with an associated Severity Level IV Non-Cited Vilation of the NRC's reporting requirements in 10 CFR 50.73, "Licensee Event Report System." The licensee failed to submit a required Licensee Event Report (LER) within 60 days after the discovery of an event that was reportable in accordance with 10 CFR 50.73(a)(2)(i)(B) as a condition which was prohibited by the plant's Technical Specifications (TS) and 10 CFR 50.73(a)(2)(v)(B) as a condition that could have prevented the fulfillment of a safety function. The condition involved an inoperable diesel generator (DG) for longer than the TS completion time for restoration. The licensee subsequently submitted the required LER.
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| Because this violation of the NRC's reporting requirements affected the NRC's ability to perform its regulatory Page 3 of 6
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| 3Q/2013 Inspection Findings - Clinton function, the inspectors evaluated the violation using the traditional enforcement process in accordance with the NRC Enforcement Policy and assessed the significance of the underlying issue using the Significance Determination Process. The finding was of more than minor significance because the NRC relies on licensees to identify and report conditions or events meeting the criteria specified in the TS and the regulations in order to perform its regulatory function and, therefore if left uncorrected it could lead to a more significant safety concern. The inspectors previously determined that the underlying issue (i.e., the failure to correctly assembly a DG ventilation system damper that resulted in an inoperable DG) was a finding of very low safety significance during a detailed risk evaluation.
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| Consistent with the guidance in Section 6.9, Paragraph d.9, of the NRC Enforcement Policy, the violation associated with this finding was determined to be a Severity Level IV Violation. This finding affected the cross-cutting area of human performance. Specifically, the licnesee's decision making process while evaluating the reportability of the condition with respect to the reporting requirements in 10 CFR 50.73 was inadequate. IMC 0310 H.1(a))
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| Inspection Report# : 2012005 (pdf)
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| Barrier Integrity Significance: Dec 31, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation FAILURE TO PERFORM PREVENTIVE MAINTENANCE ON STANDBY GAS TREATMENT SYSTEM RELAY 0UAY-VG506D.
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| A finding of very low safety significance with an associated Non-Cited Violation of Technical Specification (TS) 5.4.1.a. was self-revealed when the age-related failure of Standy Gas Treatment (VG) system relay 0UAY-VG506D caused the removal of VG Train A electric heater 0VG04AA from operation, an entry into TS 3.5.4.3 due to the inoperability of VG Train A, and an unplanned on-line plant risk condition increase from Green to Yellow. The relay failure occurred due to the licensee's failure to perform any replacement preventive maintenance on the component throughout the history of plant operation. During two separate independent reviews performed by the licensee on July 15, 2011, and on August 24, 2011, the licensee failed to correctly classify the component in accordance with its preventive maintenance procedure. This resulted in no replacement maintenance activity ever being performed for the relay and its eventual failure on August 22, 2012. The licensee initiated corrective actions to replace the relay and put in place the appropriate preventive maintenance actions.
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| The finding was of more than minor safety significance because it was sufficiently similar to several examples in Inspection Manual Chapter 0612, "Power Reactor Inspection Reports," Appendix E, "Examples of Minor Issues,"
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| wherein licensees failed to adequately implement procedural requirements and the consequences had some safety impact. The performance deficiency was also associated with the SSC [Systems, Structures, and Components] and Barrier Performance attribute and adversely affected the Barrier Integrity Cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events.
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| Specifically, the age-related failure of 0UAY-VG506D on August 22, 2012 rendered VG Train A inoperable and caused an unplanned increase in the plant's on-line risk condition from Green to Yellow. The finding was a licensee performance deficiency of very low safety significance because it only represented a degradation of the radiological barrier function provided for the Auxiliary Building and the Fuel Building and was not a complete loss of the barrier function provided by the VG system since VG Train B remained operable. This finding affected the cross-cutting area of human performance. Specifically, in the area of work control, the licensee did not appropriately coordinate work activities by incorporating actions to plan work activities to support longterm equipment reliability by scheduling maintenance as more preventive than reactive. (IMC 0310, H.3(b))
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| Inspection Report# : 2012005 (pdf)
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| Page 4 of 6
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| 3Q/2013 Inspection Findings - Clinton Emergency Preparedness Occupational Radiation Safety Significance: Jun 30, 2013 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW PROCEDURES RESULTED IN THE UNPLANNED INTAKE OF RADIOACTIVE MATERIAL BY FIVE WORKERS.
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| A self-revealing finding of very low safety significance (Green) and associated Non-Cited Violation of Technical Specification 5.4.1.a for the failure to follow procedures associated with the Radiation Work Permit (RWP) on March 28, 2013. The issue resulted in the unplanned intake of radioactive material by five workers. RWP 10014553, "2013 RW HRA/LHRA," Revision 0, established the requirement for the usage of high efficiency particulate air vacuums during the cleanup of a legacy radioactive resin spill. The licensee replaced this cleanup method with manual resin removal during the cleanup contrary to the conditions set in the RWP. This is a performance deficiency, which was within the licensee's ability to foresee and should have been prevented. The issue was entered into the licensee's corrective action program as Action Request 01494203. The licensee completed actions to ensure worker compliance with radiation protection program procedures.
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| The performance deficiency was determined to be more than minor safety significance in accordance with Inspection Manual Chapter (IMC) 0612, Appendix B, "Issue Screening," because it was associated with the program and process attribute of the Occupational Radiation Safety Cornerstone and affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation, in that, the workers received additonal and unplanned dose from the intake of radioactive materials, The significance was determined in accordance with IMC 0609, Appendix C, "Occupational Radiation Safety Significance Determination Process." The inspectors determined the finding has very low safety significance (Green) because the finding did not involve: (1) As Low As Reasonably Achievable (ALARA) planning or work controls involving excessive occupational collective dose, (2) an overexposure, (3) a substantial potential for overexposure, or (4) compromised ability to assess dose. The primary cause of this finding was related to the cross-cutting aspect of human performance with the component of decision making. The licensee failed to use conservative assumptions in decision making and failed to adopt a requirment to demonstrate that the proposed action is safe in order to proceed. H. 1(b).
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| Inspection Report# : 2013003 (pdf)
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| Significance: Mar 31, 2013 Identified By: NRC Item Type: NCV NonCited Violation INCOMPLETE ED DOSE RATE ALARM EVALUATION The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 20.1501(a) for the failure to perform surveys to ensure compliance with 10 CFR 20.1201 shallow-dose equivalent (SDE) limits for five individuals during the fourth quarter 2011 due to contamination build-up on the workers' gloves.
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| This issue was entered into the licensee's corrective action program as AR 01335298 and AR 01454976. Corrective actions include performing an apparent cause evaluation and performing dose assessments for the individuals involved.
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| The performance deficiency was determined to be of more than minor safety significance in accordance with IMC 0612, Appendix B, "Issue Screening," because it was associated with the Program And Process Attribute of the Page 5 of 6
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| 3Q/2013 Inspection Findings - Clinton Occupational Radiation Safety Cornerstone and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation, in that not performing an adequate SDE assessment affected the licensee's ability to monitor, control, and limit radiation exposures. The inspectors also reviewed the guidance in IMC 0612, Appendix E, "Examples of Minor issues," and did not find any similar examples. In accordance with IMC 0609, Appendix C, "Occupational Radiation Safety Significance Determination Process," the inspectors determined that the finding had very low safety significance because the finding did not involve: (1)
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| ALARA planning and controls, (2) a radiological overexposure, (3) a substantial potential for an overexposure, or (4) a compromised ability to assess dose. The primary cause of this finding was related to the cross-cutting aspect of human performance with the component of work practices. The specific aspect was that the licensee ensures supervisory and management oversight of work activities, including contractors, such that nuclear safety is supported (IMC 0310 H.4(c))
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| Inspection Report# : 2013002 (pdf)
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| Public Radiation Safety Security Although the Security Cornerstone is included in the Reactor Oversight Process assessment program, the Commission has decided that specific information related to findings and performance indicators pertaining to the Security Cornerstone will not be publicly available to ensure that security information is not provided to a possible adversary.
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| Other than the fact that a finding or performance indicator is Green or Greater-Than-Green, security related information will not be displayed on the public web page. Therefore, the cover letters to security inspection reports may be viewed.
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| Miscellaneous Last modified : December 03, 2013 Page 6 of 6
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| 4Q/2013 Inspection Findings - Clinton Clinton 4Q/2013 Plant Inspection Findings Initiating Events Mitigating Systems Significance: Dec 31, 2013 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IMPLEMENT REQUIREMENTS OF STATION SCAFFOLD INSTALLATION PROCEDURE.
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| Inspectors identified a NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures and Drawings for the failure to follow station procedure MA AA-796-024, Scaffold Installation, Inspection, and Removal, Revision 8, to obtain engineering approval for seismic scaffolds not complying with specific requirements of approved station procedures during the C1R14 outage. Specifically, seismic scaffolds identified during walkdowns by the inspectors did not meet procedural requirements for required clearances from or tie off to safety-related components and did not have the required engineering evaluation and approval for acceptability. The licensee documented this issue in the corrective action program (CAP) as Issue Report (IR) 01574003 and completed the required engineering review and approval.
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| The inspectors determined that the licensees failure to follow the station procedure for scaffold installation, inspection, and removal was a performance deficiency. The performance deficiency is more than minor because it was associated with the protection against external factors attribute of the Mitigating Systems (MS) cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Using IMC 0609, Attachment 4 Initial Characterization of Findings, and Appendix G Shutdown Operations Significance Determination Process, the finding was screened against Attachment 1, Checklist 8 and found to be of very low safety significance (Green) because the finding did not: 1) increase the likelihood of a loss of reactor coolant system (RCS) inventory, 2) degrade the licensees ability to terminate a leak path or add RCS inventory when needed, 3) significantly degrade the licensees ability to recover decay heat removal once it is lost, 4) result in one or less safety relief valves being available to establish a heat removal path to the suppression pool with the vessel head on. The finding was determined to have a cross-cutting aspect in the area of human performance, associated with the resources component, in that the licensee ensures that personnel, equipment, procedures and other resources are available and adequate to assure nuclear safety. Specifically, the licensee failed to ensure that the scaffold coordinator and superintendents had the required training to assure nuclear safety while erecting seismic scaffolds. [H.2(b)]
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| Inspection Report# : 2013005 (pdf)
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| Significance: Dec 31, 2013 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ASSESS AND MANAGE RISK ASSOCIATED WITH THE PERFORMANCE OF SURVEILLANCE TESTING ON AVERAGE POWER RANGE MONITORS Page 1 of 8
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| 4Q/2013 Inspection Findings - Clinton Inspectors reviewed a self-revealing NCV of 10 CFR 50.65(a)(4) for failing to manage risk when the Division 4 Nuclear System Protection System (NSPS) inverter unexpectedly transferred from its normal direct current (DC) power source to its alternate alternating current (AC) power source during the Average Power Range Monitor (APRM) D surveillance test. Specifically, the installed operational barrier failed to protect a fuse block when a test cable connector was inadvertently dropped. This caused a momentary electrical short and resulted in the inverter to transfer power sources. The licensee documented this issue in the CAP as IR 01476647 and performed (1) a stand-down with instrument maintenance craftsmen to discuss the event and lessons learned, (2) changes to the licensees risk/hazards assessment process to include a checklist designed to aid in challenging jobsite conditions, (3) conduct of paired observations by maintenance department managers on use of the checklist, and (4) a case study with the maintenance shops using this event to highlight determining risk perception and robust protective barriers.
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| The inspectors determined that the licensees failure to adequately manage the risk associated with performance of surveillance testing for APRM D was a performance deficiency. The performance deficiency is more than minor because it was associated with the configuration control attribute of the MS cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The performance deficiency involved the licensees assessment and management of risk associated with performing maintenance in accordance with 10 CFR 50.65(a)(4); therefore the inspectors used IMC 0609, Attachment 4 Initial Characterization of Findings, and Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, and determined that a detailed risk evaluation would be required since the issue represented an actual loss of safety function of a system. The Region III Senior Reactor Analyst (SRA) completed a detailed risk evaluation using the NRCs Standardized Plant Analysis Risk (SPAR) model for Clinton Power Station (CPS), Version 8.17 and SAPHIRE Version 8.09 to calculate an Incremental Core Damage Probability Deficit (ICDPD) for the unevaluated condition. The SRA ran the SPAR model conservatively assuming that High Pressure Core Spray System (HPCS) was unavailable during the 6-hour time. The result was an ICDPD of less than 2E-08/year. In accordance with IMC 0609, Appendix K, because the ICDPD was not greater than 1E 06/year, the finding was determined to be of very low safety significance (i.e., Green). The finding was determined to have a cross cutting aspect in the area of human performance, associated with the work practices component, in that personnel work practices are used commensurate with the risk of the assigned task, such that work activities are performed safely. Specifically, the technicians did not perform adequate self or peer checks after installation of the barrier to ensure the barrier would provide protection from shorting. [H.4(a)]
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| Inspection Report# : 2013005 (pdf)
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| Significance: Dec 19, 2013 Identified By: Self-Revealing Item Type: NCV NonCited Violation Insulation Resistance Testing for Unit Substation Transformers Was Incorrectly Performed A finding of very low safety significance (Green) and associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed from an event that resulted in a reactor scram. Specifically, during troubleshooting of the Unit Substation A transformer failure on December 08, 2013, it was identified that the licensee incorrectly measured the resistance between the transformer windings instead of the winding and ground. The licensee entered this concern into its Corrective Action Program as AR 01594794, and satisfactory re-measured the insulation resistance for the un-faulted transformer 1AP11E.
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| The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of equipment performance and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding screened as very low safety significance (Green), because the inspectors answered NO to all Mitigating Systems Screening questions in Exhibit 2 of Appendix A of IMC 0609. The finding was determined to have a cross-cutting aspect in the area of human performance, associated with the work control component, in that the licensee failed to ensure supervisory and management oversight of work activities, including contractors, such that nuclear safety is supported. H.4(c).
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| 4Q/2013 Inspection Findings - Clinton Inspection Report# : 2013009 (pdf)
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| Significance: Dec 19, 2013 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Acceptance Criteria in the Insulation Resistance Test Procedure The inspectors identified a finding of very low safety significance (Green) and associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to have adequate acceptance criteria in testing procedure. Specifically, the minimum acceptable insulation resistance for transformers as specified in Procedure CPS 8440.01 did not meet the minimum vendor recommended values in accordance with the vendor manual. The licensee entered this concern into its Corrective Action Program as IR 01596730 and IR 01598375.
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| The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective of ensuring capability and reliability of systems that respond to initiating events to prevent undesirable consequences. The finding screened as very low safety significance (Green), because the inspectors answered NO to all Mitigating Systems Screening questions in Exhibit 2 of Appendix A of IMC 0609. The inspectors identified the finding had a cross-cutting aspect in the area of problem identification and resolution, associated with the corrective action program component because the licensee failed to ensure issues potentially impacting nuclear safety are promptly identified. (P.1(a))
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| Inspection Report# : 2013009 (pdf)
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| Significance: Sep 30, 2013 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW PROCEDURE AND APPROPRIATELY DOCUMENT BASIS FOR IMMEDIATE OPERABILITY OF THE DIVISION 2 EMERGENCY DIESEL GENERATOR An NRC identified non-cited violation of 10CFR50, Appendix B, Criterion V, Instructions, Procedures and Drawings for the failure to follow procedure OP-AA-108-115, "Operability Determinations", Revision 11, and document the basis that a reasonable expectation of operability existed after an immediate operability determination. Specifically, after the control room received a report of a crack on the after cooler ducting of the Division 2 emergency diesel generator the licensee failed to document their basis that a reasonable expectation of operability existed for the Division 2 emergency diesel generator. The licensee documented this issue in the corrective action prgram as Action Request 015401540.
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| The inspectors determined that the licensee failing to follow the station procedure for operability determinations was a performance deficiency. Specifically, the licensee failed to follow the station procedure for operability determinations and appropriately document the decision and the basis that a reasonable expectation of operability existed for the Division 2 emergency diesel generator. The performance deficiency is more than minor because if immediate operability determination and either the basis that a reasonable expectation of operability exists or the declaration that the system, structure or component is inoperable is not appropriately documented it could lead to a more significant safety concern. Using Manual Chapter 0609, Attachment 4 "Initial Characterization of Findings," and Appendix A "The Significance Determination Process for Findings at Power" the finding was screened against the mitigating systems cornerstone and determined to be of very low safety significance (Green) because the finding was/did not: 1) a deficiency affecting the design or qualification of a mitigating structure, system or component, 2) represent a loss of system and/or function, 3) represent an actual loss of function of a single train for greater than its technical specification allowed outage time, 4) represent an actual loss of function of one or more non-technical specifications trains of equipment designated as high safety-significant for greater than 24 hours and 5) did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding or severe weather event.
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| Page 3 of 8
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| 4Q/2013 Inspection Findings - Clinton The finding was determined to have a cross-cutting aspect in the area of human performance, associated with the decision making component, in that the licnesee decisions failed to demonstrate that nuclear safety is an overriding priority. Specifically, the licensee failed to use their systematic process, when faced with an unexpected plant condition of the Division 2 emergency diesel generator to ensure safety was maintained.. H.1(a).
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| Inspection Report# : 2013004 (pdf)
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| Significance: Aug 30, 2013 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO EVALUATE A DEGRADED/NON-CONFORMING CONDITION ON DIESEL FIRE PUMP.
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| The inspectors identified a finding of very low safety significance associated with the licensee's failure to appropriately evaluate the functionality of the 'B' Diesel Fire Pump (DFP) after identifying a degraded/non-conforming crankcase pressure condition while performing testing on June 13, 2011, and on numerous occasions thereafter, that could have affected the ability of the system to perform a function important to safety. An associated NCV of Clinton Power Station License Condition 2.F was identified. The License Condition required the licensee to implement and maintain in effect all provisions of the approved Fire Protection program as described in the Updated Final Safety Analysis Report (UFSAR). Appendix E, Section 4.0.C.8 of the UFSAR stated that the Clinton Power Station Quality Assurance Program establishes measures for corrective action on conditions adverse to fire protection.
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| Quality Assurance Topical Report (QATR), Chapter 16, Section 2.4 stated that personnel performing the evaluation function of conditions adverse to quality are responsible for considering the cause and the feasibility of corrective action to assure that the necessary quality of an item is not deteriorated. The licnesee entered the issues into the CAP and initiated corrective actions to evaluate the functionality of the 'B' DFP.
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| The failure to correctly evaluate a degraded/non-conforming condition potentially affecting the functionality of structures, systems, and components (SSCs) important to safety would become a more significant safety concern if left uncorrected because it could reasonably result in an unrecognized condition of an SSC failing to fulfill a function important to safety. In addition, the finding was associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of sytems that respond to initiating events to prevent undesirable consequences. Specifically, the degraded condition of high crankcase pressure resulted in repeat operational equipment challenges and extended periods of unavailability of the 'B' DFP. Therefore the finding was of more than minor significance. The finding was a licensee performance deficienty of very low safety significance (Green) because it inolved only a low degradation of the protection against external factors function due to a redundant train that could supply water. The inspectors concluded that this finding affected the cross-cutting area of probelm identification and resolution. Specifically, the licensee failed to thoroughly evaluate problems such that the resolutions addressed causes and extent of condition as necessary for an SSC important to safety when a degraded/non-conforming condtion was identified. [P.1(c)]
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| Inspection Report# : 2013007 (pdf)
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| Significance: Mar 31, 2013 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PERFORM ADEQUATE MOV PREVENTATIVE MAINTENANCE RESULTED IN INOPERABLE RCIC SYSTEM A finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criteria V, "Instructions, Procedures, and Drawings" was self-revealed when safety-related motor operated valve 1E51-F031, reactor core isolation cooling (RCIC) system suppression pool suction valve, failed to fully close during surveillance testing on October 29, 2012. The valve failure occured due to the licensee's failure to establish an adequate procedure to perform preventive maintenance on it. Specifically, the maintenance procedure did not contain a requirement to stroke a motor operated valve during the performance of periodic stem lubrication activities. The licensee entered this Page 4 of 8
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| 4Q/2013 Inspection Findings - Clinton issue into its corrective action program for evaluation and initiated corrective actions to revise the maintenance procedure.
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| The finding was of more than minor significance since it was associated with the Procedure Quality attribute and adversely affected the Mitigation Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the valve failure rendered the RCIC system inoperable. This finding is of very low safety significance because it: (1) was not a design or qualification deficiency; (2) did not represent an actual loss of function of a system; (3) did not represent an actual loss of function of a single train or two seprate trains for greater than its Technical Specification (TS) allowed outage time; (4) did not represent an actual loss of function of one or more non-TS trains of equipment designated as high safety significant; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors concluded that this finding affected the cross-cutting area of human performance since adequate licensee resources involving personnel and procedures did not support successful human performance. Specifically, the maintenance procedure was not appropriate to the circumstances because it did not contain adequate instructions to ensure that motor operated valve stems were adequately lubricated. (IMC 0310 H.2 (c))
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| Inspection Report# : 2013002 (pdf)
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| Significance: Mar 31, 2013 Identified By: NRC Item Type: FIN Finding FAILURE TO PERFORM ADEQUATE PAST OPERABILITY EVALUATION The inspectors identified a finding of very low safety significance associated with the licensee's failure to correctly evaluate the past operability of safety-related motor operator valve 1E51-F031, reactor core isolation cooling system suppression pool suction valve, which failed quarterly surveillance testing on October 29, 2012. No violation of regulatory requirements was identified. The licensee entered this issue into its corrective action program for evaluation and initiated corrective actions to revise the past operability evaluation.
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| The finding was of more than minor significance since the failure to correctly evaluate a degraded/nonconforming condition potentially affecting the operability of structures, systems, and components (SSC) required to be operable by Technical specification (TS) would become a more significant safety concern, if left uncorrected, because it could reasonably result in an unrecognized condition of an SSC failing to fulfill a safety-related function. The finding was a licensee performance deficiency of very low safety significance because it: (1) was not a design or qualification deficiency; (2) did not represent an actual loss of function of a system; (3) did not represent an actual loss of function of a single train or two separate trains for greater than its TS allowed outage time; (4) did not represent an actual loss of function of one or more non-TS trains of equipment designated as high safety significant; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors concluded that this finding affected the cross-cutting area of human performance. Specifically, the licensee failed to use conservative assumptions in decision making while evaluating past operability of the valve by assuming that the time of inoperability was the same as the time of discovery for a time dependent failure mechanism (i.e., hardened grease) since no firm evidence to support operability was obtained by testing. (IMC 0310 H.1(b))
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| Inspection Report# : 2013002 (pdf)
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| Barrier Integrity Significance: Dec 31, 2013 Page 5 of 8
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| 4Q/2013 Inspection Findings - Clinton Identified By: NRC Item Type: FIN Finding FAILURE TO IDENTIFY EMBEDDED OPERATOR CHALLENGE Inspectors identified a finding of very low safety significance associated with the licensees failure to identify an embedded operator challenge. Specifically, the licensee proceduralized compensatory actions which were necessary in order to maintain a negative pressure (-0.25 in. H2O) inside the fuel building when opening the inner railroad bay door. The licensee documented this issue in the CAP as IR 1589104 and subsequently screened this issue as an operator challenge.
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| The inspectors determined that the licensees failure to identify an embedded operator challenge was a performance deficiency. This finding was more than minor significance because it was associated with the Barrier Integrity Cornerstone attribute of structure, system and component (SSC) and barrier performance, and adversely affected the cornerstone objective to provide reasonable assurance that the physical design barrier of secondary containment protects the public from radionuclide releases caused by accidents or events. This finding is of very low safety significance due to answering no to all questions under the Barrier Integrity Cornerstone column of IMC 0609, , Phase 1 - Initial Screening and Characterization of Findings. The inspectors concluded that this finding affected the cross-cutting aspect of problem identification and resolution. Specifically, the licensee failed to implement its CAP with a low threshold for identifying issues and did not identify this challenge to operators completely, accurately, and in a timely manner commensurate with its safety significance. [P.1(a)]
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| Inspection Report# : 2013005 (pdf)
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| Emergency Preparedness Occupational Radiation Safety Significance: Dec 31, 2013 Identified By: NRC Item Type: FIN Finding FAILURE TO MAINTAIN RADIATION EXPOSURE ALARA DURING 1R13.
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| Inspectors reviewed a self-revealing finding due to the licensee having unplanned and unintended occupational collective radiation dose because of deficiencies in the licensees Radiological Work Planning and Work Execution Program. Specifically, the licensee failed to properly incorporate as-low-as-reasonably-achievable strategies and insights while planning and executing work activity during the C1R13 refueling outage. During the In-Service Inspection (ISI) examinations performed inside the bio-shield, the dose overage was 28.410 person-rem (68 percent higher than initial estimate). This result was caused by poor radiological planning and work execution of these tasks.
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| The licensee entered this issue into their CAP as IR 01593794 and incorporated the lesson learned into the outage planning.
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| The inspectors determined that the failure to appropriately plan and coordinate outage activities, together with the failure to properly incorporate ALARA strategies or insights while planning and executing ISI examinations inside the bio-shield during the C1R13 refueling outage was a performance deficiency. The finding was more than minor because it was associated with the program and process attribute of the Occupational Radiation Safety Cornerstone.
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| This issue affected the cornerstone objective of ensuring the adequate protection of worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. The finding is also very similar to IMC 0612, Appendix E, Examples of Minor Issues, Example 6.i. This example provides guidance Page 6 of 8
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| 4Q/2013 Inspection Findings - Clinton that an issue is not minor if the actual collective dose exceeded 5 person-rem and exceeded the planned, intended dose by more than 50 percent. The inspectors determined that this finding was of very low safety significance because CPSs 3-year rolling average collective was less than the 240 person-rem/unit referenced within IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process. This finding did not have a cross cutting aspect due to not being reflective of current performance as exemplified by improvements in the recently completed C1R14 outage.
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| Inspection Report# : 2013005 (pdf)
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| Significance: Jun 30, 2013 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW PROCEDURES RESULTED IN THE UNPLANNED INTAKE OF RADIOACTIVE MATERIAL BY FIVE WORKERS.
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| A self-revealing finding of very low safety significance (Green) and associated Non-Cited Violation of Technical Specification 5.4.1.a for the failure to follow procedures associated with the Radiation Work Permit (RWP) on March 28, 2013. The issue resulted in the unplanned intake of radioactive material by five workers. RWP 10014553, "2013 RW HRA/LHRA," Revision 0, established the requirement for the usage of high efficiency particulate air vacuums during the cleanup of a legacy radioactive resin spill. The licensee replaced this cleanup method with manual resin removal during the cleanup contrary to the conditions set in the RWP. This is a performance deficiency, which was within the licensee's ability to foresee and should have been prevented. The issue was entered into the licensee's corrective action program as Action Request 01494203. The licensee completed actions to ensure worker compliance with radiation protection program procedures.
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| The performance deficiency was determined to be more than minor safety significance in accordance with Inspection Manual Chapter (IMC) 0612, Appendix B, "Issue Screening," because it was associated with the program and process attribute of the Occupational Radiation Safety Cornerstone and affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation, in that, the workers received additonal and unplanned dose from the intake of radioactive materials, The significance was determined in accordance with IMC 0609, Appendix C, "Occupational Radiation Safety Significance Determination Process." The inspectors determined the finding has very low safety significance (Green) because the finding did not involve: (1) As Low As Reasonably Achievable (ALARA) planning or work controls involving excessive occupational collective dose, (2) an overexposure, (3) a substantial potential for overexposure, or (4) compromised ability to assess dose. The primary cause of this finding was related to the cross-cutting aspect of human performance with the component of decision making. The licensee failed to use conservative assumptions in decision making and failed to adopt a requirment to demonstrate that the proposed action is safe in order to proceed. H. 1(b).
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| Inspection Report# : 2013003 (pdf)
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| Significance: Mar 31, 2013 Identified By: NRC Item Type: NCV NonCited Violation INCOMPLETE ED DOSE RATE ALARM EVALUATION The inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 20.1501(a) for the failure to perform surveys to ensure compliance with 10 CFR 20.1201 shallow-dose equivalent (SDE) limits for five individuals during the fourth quarter 2011 due to contamination build-up on the workers' gloves.
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| This issue was entered into the licensee's corrective action program as AR 01335298 and AR 01454976. Corrective actions include performing an apparent cause evaluation and performing dose assessments for the individuals involved.
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| Page 7 of 8
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| 4Q/2013 Inspection Findings - Clinton The performance deficiency was determined to be of more than minor safety significance in accordance with IMC 0612, Appendix B, "Issue Screening," because it was associated with the Program And Process Attribute of the Occupational Radiation Safety Cornerstone and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation, in that not performing an adequate SDE assessment affected the licensee's ability to monitor, control, and limit radiation exposures. The inspectors also reviewed the guidance in IMC 0612, Appendix E, "Examples of Minor issues," and did not find any similar examples. In accordance with IMC 0609, Appendix C, "Occupational Radiation Safety Significance Determination Process," the inspectors determined that the finding had very low safety significance because the finding did not involve: (1)
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| ALARA planning and controls, (2) a radiological overexposure, (3) a substantial potential for an overexposure, or (4) a compromised ability to assess dose. The primary cause of this finding was related to the cross-cutting aspect of human performance with the component of work practices. The specific aspect was that the licensee ensures supervisory and management oversight of work activities, including contractors, such that nuclear safety is supported (IMC 0310 H.4(c))
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| Inspection Report# : 2013002 (pdf)
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| Public Radiation Safety Security Although the Security Cornerstone is included in the Reactor Oversight Process assessment program, the Commission has decided that specific information related to findings and performance indicators pertaining to the Security Cornerstone will not be publicly available to ensure that security information is not provided to a possible adversary.
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| Other than the fact that a finding or performance indicator is Green or Greater-Than-Green, security related information will not be displayed on the public web page. Therefore, the cover letters to security inspection reports may be viewed.
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| Miscellaneous Last modified : February 24, 2014 Page 8 of 8
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| 1Q/2014 Inspection Findings - Clinton Clinton 1Q/2014 Plant Inspection Findings Initiating Events Significance: Feb 14, 2014 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct Identified Combustibles Violations of very low safety significance or severity Level IV that were identified by the licensee have been reviewed by the inspectors. Corrective actions taken or planned by the licensee have been entered into the licensees Corrective Action Program. These violations and corrective action tracking numbers are listed in Section 4OA7 of this report.
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| Inspection Report# : 2014007 (pdf)
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| Mitigating Systems Significance: Dec 31, 2013 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IMPLEMENT REQUIREMENTS OF STATION SCAFFOLD INSTALLATION PROCEDURE.
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| Inspectors identified a NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures and Drawings for the failure to follow station procedure MA AA-796-024, Scaffold Installation, Inspection, and Removal, Revision 8, to obtain engineering approval for seismic scaffolds not complying with specific requirements of approved station procedures during the C1R14 outage. Specifically, seismic scaffolds identified during walkdowns by the inspectors did not meet procedural requirements for required clearances from or tie off to safety-related components and did not have the required engineering evaluation and approval for acceptability. The licensee documented this issue in the corrective action program (CAP) as Issue Report (IR) 01574003 and completed the required engineering review and approval.
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| The inspectors determined that the licensees failure to follow the station procedure for scaffold installation, inspection, and removal was a performance deficiency. The performance deficiency is more than minor because it was associated with the protection against external factors attribute of the Mitigating Systems (MS) cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Using IMC 0609, Attachment 4 Initial Characterization of Findings, and Appendix G Shutdown Operations Significance Determination Process, the finding was screened against Attachment 1, Checklist 8 and found to be of very low safety significance (Green) because the finding did not: 1) increase the likelihood of a loss of reactor coolant system (RCS) inventory, 2) degrade the licensees ability to terminate a leak path or add RCS inventory when needed, 3) significantly degrade the licensees ability to recover decay heat removal once it is lost, 4) result in one or less safety relief valves being available to establish a heat removal path to the suppression pool with the vessel head on. The finding was determined to have a cross-cutting aspect in the area of human performance, associated with the resources component, in that the licensee ensures that personnel, equipment, procedures and other resources are available and adequate to assure nuclear safety. Specifically, the licensee failed to ensure that the scaffold coordinator and superintendents had the Page 1 of 8
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| 1Q/2014 Inspection Findings - Clinton required training to assure nuclear safety while erecting seismic scaffolds. [H.2(b)]
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| Inspection Report# : 2013005 (pdf)
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| Significance: Dec 31, 2013 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ASSESS AND MANAGE RISK ASSOCIATED WITH THE PERFORMANCE OF SURVEILLANCE TESTING ON AVERAGE POWER RANGE MONITORS Inspectors reviewed a self-revealing NCV of 10 CFR 50.65(a)(4) for failing to manage risk when the Division 4 Nuclear System Protection System (NSPS) inverter unexpectedly transferred from its normal direct current (DC) power source to its alternate alternating current (AC) power source during the Average Power Range Monitor (APRM) D surveillance test. Specifically, the installed operational barrier failed to protect a fuse block when a test cable connector was inadvertently dropped. This caused a momentary electrical short and resulted in the inverter to transfer power sources. The licensee documented this issue in the CAP as IR 01476647 and performed (1) a stand-down with instrument maintenance craftsmen to discuss the event and lessons learned, (2) changes to the licensees risk/hazards assessment process to include a checklist designed to aid in challenging jobsite conditions, (3) conduct of paired observations by maintenance department managers on use of the checklist, and (4) a case study with the maintenance shops using this event to highlight determining risk perception and robust protective barriers.
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| The inspectors determined that the licensees failure to adequately manage the risk associated with performance of surveillance testing for APRM D was a performance deficiency. The performance deficiency is more than minor because it was associated with the configuration control attribute of the MS cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The performance deficiency involved the licensees assessment and management of risk associated with performing maintenance in accordance with 10 CFR 50.65(a)(4); therefore the inspectors used IMC 0609, Attachment 4 Initial Characterization of Findings, and Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, and determined that a detailed risk evaluation would be required since the issue represented an actual loss of safety function of a system. The Region III Senior Reactor Analyst (SRA) completed a detailed risk evaluation using the NRCs Standardized Plant Analysis Risk (SPAR) model for Clinton Power Station (CPS), Version 8.17 and SAPHIRE Version 8.09 to calculate an Incremental Core Damage Probability Deficit (ICDPD) for the unevaluated condition. The SRA ran the SPAR model conservatively assuming that High Pressure Core Spray System (HPCS) was unavailable during the 6-hour time. The result was an ICDPD of less than 2E-08/year. In accordance with IMC 0609, Appendix K, because the ICDPD was not greater than 1E 06/year, the finding was determined to be of very low safety significance (i.e., Green). The finding was determined to have a cross cutting aspect in the area of human performance, associated with the work practices component, in that personnel work practices are used commensurate with the risk of the assigned task, such that work activities are performed safely. Specifically, the technicians did not perform adequate self or peer checks after installation of the barrier to ensure the barrier would provide protection from shorting. [H.4(a)]
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| Inspection Report# : 2013005 (pdf)
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| Significance: Dec 19, 2013 Identified By: Self-Revealing Item Type: NCV NonCited Violation Insulation Resistance Testing for Unit Substation Transformers Was Incorrectly Performed A finding of very low safety significance (Green) and associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed from an event that resulted in a reactor scram. Specifically, during troubleshooting of the Unit Substation A transformer failure on December 08, 2013, it was identified that the licensee incorrectly measured the resistance between the transformer windings instead Page 2 of 8
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| 1Q/2014 Inspection Findings - Clinton of the winding and ground. The licensee entered this concern into its Corrective Action Program as AR 01594794, and satisfactory re-measured the insulation resistance for the un-faulted transformer 1AP11E.
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| The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of equipment performance and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding screened as very low safety significance (Green), because the inspectors answered NO to all Mitigating Systems Screening questions in Exhibit 2 of Appendix A of IMC 0609. The finding was determined to have a cross-cutting aspect in the area of human performance, associated with the work control component, in that the licensee failed to ensure supervisory and management oversight of work activities, including contractors, such that nuclear safety is supported. H.4(c).
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| Inspection Report# : 2013009 (pdf)
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| Significance: Dec 19, 2013 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Acceptance Criteria in the Insulation Resistance Test Procedure The inspectors identified a finding of very low safety significance (Green) and associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to have adequate acceptance criteria in testing procedure. Specifically, the minimum acceptable insulation resistance for transformers as specified in Procedure CPS 8440.01 did not meet the minimum vendor recommended values in accordance with the vendor manual. The licensee entered this concern into its Corrective Action Program as IR 01596730 and IR 01598375.
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| The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective of ensuring capability and reliability of systems that respond to initiating events to prevent undesirable consequences. The finding screened as very low safety significance (Green), because the inspectors answered NO to all Mitigating Systems Screening questions in Exhibit 2 of Appendix A of IMC 0609. The inspectors identified the finding had a cross-cutting aspect in the area of problem identification and resolution, associated with the corrective action program component because the licensee failed to ensure issues potentially impacting nuclear safety are promptly identified. (P.1(a))
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| Inspection Report# : 2013009 (pdf)
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| Significance: Sep 30, 2013 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW PROCEDURE AND APPROPRIATELY DOCUMENT BASIS FOR IMMEDIATE OPERABILITY OF THE DIVISION 2 EMERGENCY DIESEL GENERATOR An NRC identified non-cited violation of 10CFR50, Appendix B, Criterion V, Instructions, Procedures and Drawings for the failure to follow procedure OP-AA-108-115, "Operability Determinations", Revision 11, and document the basis that a reasonable expectation of operability existed after an immediate operability determination. Specifically, after the control room received a report of a crack on the after cooler ducting of the Division 2 emergency diesel generator the licensee failed to document their basis that a reasonable expectation of operability existed for the Division 2 emergency diesel generator. The licensee documented this issue in the corrective action prgram as Action Request 015401540.
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| The inspectors determined that the licensee failing to follow the station procedure for operability determinations was a performance deficiency. Specifically, the licensee failed to follow the station procedure for operability determinations and appropriately document the decision and the basis that a reasonable expectation of operability existed for the Division 2 emergency diesel generator. The performance deficiency is more than minor because if immediate operability determination and either the basis that a reasonable expectation of operability exists or the declaration that Page 3 of 8
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| 1Q/2014 Inspection Findings - Clinton the system, structure or component is inoperable is not appropriately documented it could lead to a more significant safety concern. Using Manual Chapter 0609, Attachment 4 "Initial Characterization of Findings," and Appendix A "The Significance Determination Process for Findings at Power" the finding was screened against the mitigating systems cornerstone and determined to be of very low safety significance (Green) because the finding was/did not: 1) a deficiency affecting the design or qualification of a mitigating structure, system or component, 2) represent a loss of system and/or function, 3) represent an actual loss of function of a single train for greater than its technical specification allowed outage time, 4) represent an actual loss of function of one or more non-technical specifications trains of equipment designated as high safety-significant for greater than 24 hours and 5) did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding or severe weather event.
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| The finding was determined to have a cross-cutting aspect in the area of human performance, associated with the decision making component, in that the licnesee decisions failed to demonstrate that nuclear safety is an overriding priority. Specifically, the licensee failed to use their systematic process, when faced with an unexpected plant condition of the Division 2 emergency diesel generator to ensure safety was maintained.. H.1(a).
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| Inspection Report# : 2013004 (pdf)
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| Significance: Aug 30, 2013 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO EVALUATE A DEGRADED/NON-CONFORMING CONDITION ON DIESEL FIRE PUMP.
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| The inspectors identified a finding of very low safety significance associated with the licensee's failure to appropriately evaluate the functionality of the 'B' Diesel Fire Pump (DFP) after identifying a degraded/non-conforming crankcase pressure condition while performing testing on June 13, 2011, and on numerous occasions thereafter, that could have affected the ability of the system to perform a function important to safety. An associated NCV of Clinton Power Station License Condition 2.F was identified. The License Condition required the licensee to implement and maintain in effect all provisions of the approved Fire Protection program as described in the Updated Final Safety Analysis Report (UFSAR). Appendix E, Section 4.0.C.8 of the UFSAR stated that the Clinton Power Station Quality Assurance Program establishes measures for corrective action on conditions adverse to fire protection.
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| Quality Assurance Topical Report (QATR), Chapter 16, Section 2.4 stated that personnel performing the evaluation function of conditions adverse to quality are responsible for considering the cause and the feasibility of corrective action to assure that the necessary quality of an item is not deteriorated. The licnesee entered the issues into the CAP and initiated corrective actions to evaluate the functionality of the 'B' DFP.
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| The failure to correctly evaluate a degraded/non-conforming condition potentially affecting the functionality of structures, systems, and components (SSCs) important to safety would become a more significant safety concern if left uncorrected because it could reasonably result in an unrecognized condition of an SSC failing to fulfill a function important to safety. In addition, the finding was associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of sytems that respond to initiating events to prevent undesirable consequences. Specifically, the degraded condition of high crankcase pressure resulted in repeat operational equipment challenges and extended periods of unavailability of the 'B' DFP. Therefore the finding was of more than minor significance. The finding was a licensee performance deficienty of very low safety significance (Green) because it inolved only a low degradation of the protection against external factors function due to a redundant train that could supply water. The inspectors concluded that this finding affected the cross-cutting area of probelm identification and resolution. Specifically, the licensee failed to thoroughly evaluate problems such that the resolutions addressed causes and extent of condition as necessary for an SSC important to safety when a degraded/non-conforming condtion was identified. [P.1(c)]
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| Inspection Report# : 2013007 (pdf)
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| Page 4 of 8
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| 1Q/2014 Inspection Findings - Clinton Barrier Integrity Significance: Dec 31, 2013 Identified By: NRC Item Type: FIN Finding FAILURE TO IDENTIFY EMBEDDED OPERATOR CHALLENGE Inspectors identified a finding of very low safety significance associated with the licensees failure to identify an embedded operator challenge. Specifically, the licensee proceduralized compensatory actions which were necessary in order to maintain a negative pressure (-0.25 in. H2O) inside the fuel building when opening the inner railroad bay door. The licensee documented this issue in the CAP as IR 1589104 and subsequently screened this issue as an operator challenge.
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| The inspectors determined that the licensees failure to identify an embedded operator challenge was a performance deficiency. This finding was more than minor significance because it was associated with the Barrier Integrity Cornerstone attribute of structure, system and component (SSC) and barrier performance, and adversely affected the cornerstone objective to provide reasonable assurance that the physical design barrier of secondary containment protects the public from radionuclide releases caused by accidents or events. This finding is of very low safety significance due to answering no to all questions under the Barrier Integrity Cornerstone column of IMC 0609, , Phase 1 - Initial Screening and Characterization of Findings. The inspectors concluded that this finding affected the cross-cutting aspect of problem identification and resolution. Specifically, the licensee failed to implement its CAP with a low threshold for identifying issues and did not identify this challenge to operators completely, accurately, and in a timely manner commensurate with its safety significance. [P.1(a)]
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| Inspection Report# : 2013005 (pdf)
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| Emergency Preparedness Occupational Radiation Safety Significance: Dec 31, 2013 Identified By: NRC Item Type: FIN Finding FAILURE TO MAINTAIN RADIATION EXPOSURE ALARA DURING 1R13.
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| Inspectors reviewed a self-revealing finding due to the licensee having unplanned and unintended occupational collective radiation dose because of deficiencies in the licensees Radiological Work Planning and Work Execution Program. Specifically, the licensee failed to properly incorporate as-low-as-reasonably-achievable strategies and insights while planning and executing work activity during the C1R13 refueling outage. During the In-Service Inspection (ISI) examinations performed inside the bio-shield, the dose overage was 28.410 person-rem (68 percent higher than initial estimate). This result was caused by poor radiological planning and work execution of these tasks.
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| The licensee entered this issue into their CAP as IR 01593794 and incorporated the lesson learned into the outage planning.
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| The inspectors determined that the failure to appropriately plan and coordinate outage activities, together with the failure to properly incorporate ALARA strategies or insights while planning and executing ISI examinations inside the Page 5 of 8
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| 1Q/2014 Inspection Findings - Clinton bio-shield during the C1R13 refueling outage was a performance deficiency. The finding was more than minor because it was associated with the program and process attribute of the Occupational Radiation Safety Cornerstone.
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| This issue affected the cornerstone objective of ensuring the adequate protection of worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. The finding is also very similar to IMC 0612, Appendix E, Examples of Minor Issues, Example 6.i. This example provides guidance that an issue is not minor if the actual collective dose exceeded 5 person-rem and exceeded the planned, intended dose by more than 50 percent. The inspectors determined that this finding was of very low safety significance because CPSs 3-year rolling average collective was less than the 240 person-rem/unit referenced within IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process. This finding did not have a cross cutting aspect due to not being reflective of current performance as exemplified by improvements in the recently completed C1R14 outage.
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| Inspection Report# : 2013005 (pdf)
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| Significance: Jun 30, 2013 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW PROCEDURES RESULTED IN THE UNPLANNED INTAKE OF RADIOACTIVE MATERIAL BY FIVE WORKERS.
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| A self-revealing finding of very low safety significance (Green) and associated Non-Cited Violation of Technical Specification 5.4.1.a for the failure to follow procedures associated with the Radiation Work Permit (RWP) on March 28, 2013. The issue resulted in the unplanned intake of radioactive material by five workers. RWP 10014553, "2013 RW HRA/LHRA," Revision 0, established the requirement for the usage of high efficiency particulate air vacuums during the cleanup of a legacy radioactive resin spill. The licensee replaced this cleanup method with manual resin removal during the cleanup contrary to the conditions set in the RWP. This is a performance deficiency, which was within the licensee's ability to foresee and should have been prevented. The issue was entered into the licensee's corrective action program as Action Request 01494203. The licensee completed actions to ensure worker compliance with radiation protection program procedures.
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| The performance deficiency was determined to be more than minor safety significance in accordance with Inspection Manual Chapter (IMC) 0612, Appendix B, "Issue Screening," because it was associated with the program and process attribute of the Occupational Radiation Safety Cornerstone and affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation, in that, the workers received additonal and unplanned dose from the intake of radioactive materials, The significance was determined in accordance with IMC 0609, Appendix C, "Occupational Radiation Safety Significance Determination Process." The inspectors determined the finding has very low safety significance (Green) because the finding did not involve: (1) As Low As Reasonably Achievable (ALARA) planning or work controls involving excessive occupational collective dose, (2) an overexposure, (3) a substantial potential for overexposure, or (4) compromised ability to assess dose. The primary cause of this finding was related to the cross-cutting aspect of human performance with the component of decision making. The licensee failed to use conservative assumptions in decision making and failed to adopt a requirment to demonstrate that the proposed action is safe in order to proceed. H. 1(b).
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| Inspection Report# : 2013003 (pdf)
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| Public Radiation Safety Page 6 of 8
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| 1Q/2014 Inspection Findings - Clinton Security Although the Security Cornerstone is included in the Reactor Oversight Process assessment program, the Commission has decided that specific information related to findings and performance indicators pertaining to the Security Cornerstone will not be publicly available to ensure that security information is not provided to a possible adversary.
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| Other than the fact that a finding or performance indicator is Green or Greater-Than-Green, security related information will not be displayed on the public web page. Therefore, the cover letters to security inspection reports may be viewed.
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| Miscellaneous Significance: N/A Mar 31, 2009 Identified By: NRC Item Type: AV Apparent Violation Apparent Violation for Exelon Plants - 1 (2009 Findings)
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| For apparent violation #1:
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| Contrary to the above, on March 31, 2009 Exelon Generation Company, LLC (Exelon) provided incomplete and inaccurate information on the status of its decommissioning funding, as required by 10 CFR 50.75 when it submitted the decommissioning funding status report. Specifically, the March 31, 2009, decommissioning funding status (DFS) report contained inaccurate and incomplete information regarding Exelons compliance with the requirements of 10 CFR 50.75. The report stated that the amount listed for each of the reactors was determined in accordance with 10 CFR 50.75(b) and the applicable formulas of 10 CFR 50.75(c). However, for each of the 23 reactors, the amount reported was a discounted value that was less than the minimum required amount specified by 10 CFR 50.75(b) and (c). The report was material to the NRC because Exelon under-reported its certified decommissioning amounts by approximately $4 billion, and the NRC staff evaluated the status of Exelons decommissioning funds based on the inaccurate reports. After identifying the inaccurate information, the NRC required parent company guarantees before the staff could make its determination that there was reasonable assurance that funds will be available for the decommissioning process.
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| Inspection Report# : 2013201 (pdf)
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| Inspection Report# : 2012012 (pdf)
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| Significance: N/A Mar 31, 2009 Identified By: NRC Item Type: AV Apparent Violation Apparent Violation for Exelon Plants - 2 (2009 Findings)
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| For apparent violation #2:
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| Contrary to the above, on March 31, 2007, and March 31, 2005, Exelon Generation Company, LLC (Exelon) provided incomplete and inaccurate information on the status of its decommissioning funding, as required by 10 CFR 50.75 when it submitted the decommissioning funding status reports. Specifically, the March 31, 2007, and March 31, 2005, decommissioning funding status (DFS) reports contained inaccurate and incomplete information regarding Exelons compliance with the requirements of 10 CFR 50.75. The reports stated that the amount listed for each of the reactors was determined in accordance with 10 CFR 50.75(b) and the applicable formulas of 10 CFR 50.75(c). However, in multiple instances, the amount reported was a discounted value that was less than the minimum required amount specified by 10 CFR 50.75(b) and (c). The reports were material to the NRC because Exelon under-reported its certified decommissioning amounts, and the NRC staff evaluated the status of Exelons decommissioning funds based on the inaccurate reports. After identifying the inaccurate information, the NRC required parent company guarantees before the staff could make its determination that there was reasonable assurance that funds will be available for the decommissioning process.
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| Page 7 of 8
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| 1Q/2014 Inspection Findings - Clinton Inspection Report# : 2012012 (pdf)
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| Inspection Report# : 2013201 (pdf)
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| Last modified : May 30, 2014 Page 8 of 8
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| 2Q/2014 Inspection Findings - Clinton Clinton 2Q/2014 Plant Inspection Findings Initiating Events Significance: Jun 30, 2014 Identified By: NRC Item Type: FIN Finding ELECTRO HYDRAULIC CONTROL SYSTEM LEAK RESULTS IN MANUAL SCRAM The inspectors documented a self-revealing, Green finding associated with a failure to provide adequate work instructions to perform repairs to the shutoff valve for 1TGCV4 main turbine control valve. Specifically, contrary to station procedure MA-AA-716-010, Maintenance Planning, Revision 21, the work instructions generated to install the shutoff valve failed to reference the appropriate cap screw size, lubricate the cap screws and install lock washers on the cap screws used to attach the shut off valve to the control valve. This allowed the cap screws to loosen and ultimately fail due to fatigue resulting in a leak of electro hydraulic control fluid of sufficient rate to require a manual scram of Unit 1 on April 26, 2013. The valve was replaced and successfully tested and the unit was restarted. The licensee documented this issue in the corrective action program (CAP) as Issue Report (IR) 01506929.
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| The performance deficiency was more than minor because it was associated with the procedure quality attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations and is therefore a finding. Using Manual Chapter 0609, Attachment 4 Initial Characterization of Findings, and Appendix A The Significance Determination Process for Findings at Power, issued June 19, 2012, the finding was screened against the initiating events cornerstone and determined to be of very low safety significance (Green) because the finding did not cause a reactor trip with a coincident loss of mitigating equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The inspectors determined that no cross cutting aspect will be assigned to this performance deficiency since it occurred in 2008 and is not indicative of current plant performance.
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| Inspection Report# : 2014003 (pdf)
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| Significance: Jun 30, 2014 Identified By: NRC Item Type: FIN Finding FAILURE TO IMPLEMENT ENGINEERING CHANGE RESULTS IN MANUAL REACTOR SCRAM The inspectors documented a self-revealing, Green finding associated with a failure to implement engineering change (EC) 380150 Upgrade Feed Water Level Control and Turbine Speed. Specifically, contrary to station procedure CC-AA-256, Process for Managing Plant Modifications Involving Microprocessor Technology, Revision 2, the licensee failed to identify, evaluate and mitigate software component critical parameters in the engineering change that installed the digital feed water system. This resulted in nonlinear reactor water level oscillations when transferring from the motor driven feed pump to the turbine driven feed pump that required the reactor operator to manually scram the reactor prior to reaching the level 8 automatic scram set point. The licensee documented this issue in the corrective action program as IR 1596987.
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| The performance deficiency was more than minor because it was associated with the design control attribute of the Page 1 of 9
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| 2Q/2014 Inspection Findings - Clinton Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations and is therefore a finding. Using Manual Chapter 0609, Attachment 4 Initial Characterization of Findings, and Appendix A The Significance Determination Process for Findings at Power, issued June 19, 2012, the finding was screened against the initiating events cornerstone and determined to be of very low safety significance (Green) because the finding did not cause a reactor trip with a coincident loss of mitigating equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The inspectors determined this finding affected the cross cutting area of human performance in the aspect of documentation where the organization creates and maintains complete, accurate and up-to date documentation. Specifically, the contractors failed to create complete documentation to be use by the licensee when evaluating the critical parameters.
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| Inspection Report# : 2014003 (pdf)
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| Significance: Feb 14, 2014 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct Identified Combustibles The inspectors identified a finding of very low safety significance and associated NCV of License Condition 2.F for the failure to remove an identified combustible. Specifically, the failure to remove a piece of wood located directly under a safety-related cable tray for a period in excess of three years was a failure to take corrective action as required by the licensees Quality Assurance Program. The licensee entered the issue into their Corrective Action Program and removed the piece of wood by the end of the inspection.
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| The finding was determined to be more than minor because the combustible material was located directly beneath a safety-related cable tray and, as such, represented a credible fire scenario. The finding was determined to be of very low safety significance (i.e., Green) because the impact of the fire would be largely limited to one train/division of equipment important to safety. The inspectors determined that the finding has a cross-cutting aspect in the area of human performance because the licensee did not ensure sufficient resources were available to support nuclear safety.
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| Specifically, the failure to remove the identified combustible was due to a lack of resources to schedule and accomplish removing the material.
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| Inspection Report# : 2014007 (pdf)
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| Mitigating Systems Significance: Jun 30, 2014 Identified By: NRC Item Type: NCV NonCited Violation FOREIGN MATERIAL IN RELAY PREVENTS EMERGENCY DIESEL GENERATOR OUTPUT BREAKER FROM CLOSING The inspectors documented a self-revealing, Green non-cited violation of Clinton Power Station Technical Specification 5.4.1, Procedures, for a failure to prevent foreign material from entering a relay associated with the Division 1 Diesel Generator. Specifically, contrary to station procedure CPS 8501.05, CV-2 Relay Inspection and Calibration with Doble Test Equipment, Revision 4, the licensee failed to verify that relay 227-DGIKA, CV-2 AB phase was clean and free of all foreign material. The foreign material prevented the relay from operating and satisfying the permissive logic required to close the Division 1 Diesel Generator output breaker resulting in having to Page 2 of 9
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| 2Q/2014 Inspection Findings - Clinton declare the Diesel Generator inoperable. The relay was replaced and successfully tested and the licensee documented this issue in the corrective action program as IR 01600935.
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| The finding was more than minor because it was associated with the procedure quality attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences and is therefore a finding.
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| Using Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012, Exhibit 2 for the Mitigating Systems Cornerstone, the inspectors answered Yes to the screening question under the Mitigating Systems Cornerstone Does the finding represent an actual loss of function of at least a single Train for > its Tech Spec Allowed Outage Time OR two separate safety systems out-of- service for >
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| its Tech Spec Allowed Outage Time?," since the finding represented an actual loss of function of at least a single Train for > its Tech Spec Allowed Outage Time of 14 days. Therefore, a detailed risk evaluation was performed using IMC 0609, Appendix A. The Senior Reactor Analysts (SRAs) evaluated the finding using the Clinton Standardized Plant Analysis Risk (SPAR) model version 8.17, Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) version 8.1.0 and concluded that the risk increase to the plant due to this finding is very low (Green). The inspectors determined this finding affected the cross cutting area of human performance in the aspect of work management where the organization implements a process of planning, controlling and executing work activities such that nuclear safety is the overriding priority. Specifically, the licensees implementation of their foreign material exclusion process for this maintenance activity lacked sufficient planning, controls and execution to prevent foreign material from entering a risk significant piece of safety related equipment.
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| Inspection Report# : 2014003 (pdf)
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| Significance: Jun 30, 2014 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO DEVELOP ADEQUATE PROCEDURES FOR PRE-PLANNING AND PERFORMING MAINTENANCE AFFECTING SAFETY-RELATED EQUIPMENT The inspectors documented a self-revealing, Green non-cited violation (NCV) of Clinton Power Station Technical Specification 5.4.1, Procedures for a failure to develop adequate procedures for properly pre-planning and performing maintenance affecting the performance of safety-related equipment which resulted in the subsequent failure of the Division 3 Diesel Room Ventilation damper hydramotor on August 15, 2013. Specifically, during pre-scheduled performance testing of the Division 3 (High Pressure Core Spray System) Emergency Diesel Generator Room Ventilation Damper hydramotor, the ventilation supply air intake damper, 1VD01YC, failed to open as a result of Damper Hydramotor 1TZVD003A experiencing an age-related degradation failure. This was due in part to the licensees failure to properly pre-plan and perform the appropriate preventive maintenance for the hydramotor due to inadequate procedures. Procedure WC-AA-113, Predefine Database Revisions, Revision 2, did not provide adequate instructions appropriate to the circumstances to properly pre-plan and perform maintenance affecting the performance of safety-related equipment. This resulted in a loss of safety function of the HPCS Diesel Generator and its supported High Pressure Core Spray system because of the low confidence that diesel room temperature would be maintained to support the diesel during an event when it would be required to perform its function. The licensee subsequently replaced the hydramotor, tested the new hydramotor successfully and restored the diesel ventilation system to operable. They documented this issue in the corrective action program as IR 1546973 and IR 1547294.
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| The performance deficiency was more than minor because it was associated with the equipment performance attribute of the mitigating systems cornerstone attribute and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences and is therefore a finding. Using Manual Chapter 0609, Appendix A, The SDP for Findings At-Power, issued June 19, 2012, Exhibit 2 for the Mitigating Systems Cornerstone. The inspectors answered Yes to the screening question under the Mitigating Screening Cornerstone Does the finding represent a loss of system and/or function? since the Page 3 of 9
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| 2Q/2014 Inspection Findings - Clinton finding resulted in a loss of safety function. Therefore, a detailed risk evaluation was performed using IMC 0609, Appendix A. The SRAs evaluated the finding using the Clinton SPAR model version 8.17, SAPHIRE version 8.1.0 and concluded that the risk increase to the plant due to this finding is very low (Green). The inspectors determined that no cross-cutting aspect will be assigned to this performance deficiency since it occurred in 2005 and is not indicative of current plant performance Inspection Report# : 2014003 (pdf)
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| Significance: Dec 31, 2013 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IMPLEMENT REQUIREMENTS OF STATION SCAFFOLD INSTALLATION PROCEDURE.
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| Inspectors identified a NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures and Drawings for the failure to follow station procedure MA AA-796-024, Scaffold Installation, Inspection, and Removal, Revision 8, to obtain engineering approval for seismic scaffolds not complying with specific requirements of approved station procedures during the C1R14 outage. Specifically, seismic scaffolds identified during walkdowns by the inspectors did not meet procedural requirements for required clearances from or tie off to safety-related components and did not have the required engineering evaluation and approval for acceptability. The licensee documented this issue in the corrective action program (CAP) as Issue Report (IR) 01574003 and completed the required engineering review and approval.
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| The inspectors determined that the licensees failure to follow the station procedure for scaffold installation, inspection, and removal was a performance deficiency. The performance deficiency is more than minor because it was associated with the protection against external factors attribute of the Mitigating Systems (MS) cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Using IMC 0609, Attachment 4 Initial Characterization of Findings, and Appendix G Shutdown Operations Significance Determination Process, the finding was screened against Attachment 1, Checklist 8 and found to be of very low safety significance (Green) because the finding did not: 1) increase the likelihood of a loss of reactor coolant system (RCS) inventory, 2) degrade the licensees ability to terminate a leak path or add RCS inventory when needed, 3) significantly degrade the licensees ability to recover decay heat removal once it is lost, 4) result in one or less safety relief valves being available to establish a heat removal path to the suppression pool with the vessel head on. The finding was determined to have a cross-cutting aspect in the area of human performance, associated with the resources component, in that the licensee ensures that personnel, equipment, procedures and other resources are available and adequate to assure nuclear safety. Specifically, the licensee failed to ensure that the scaffold coordinator and superintendents had the required training to assure nuclear safety while erecting seismic scaffolds. [H.2(b)]
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| Inspection Report# : 2013005 (pdf)
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| Significance: Dec 31, 2013 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ASSESS AND MANAGE RISK ASSOCIATED WITH THE PERFORMANCE OF SURVEILLANCE TESTING ON AVERAGE POWER RANGE MONITORS Inspectors reviewed a self-revealing NCV of 10 CFR 50.65(a)(4) for failing to manage risk when the Division 4 Nuclear System Protection System (NSPS) inverter unexpectedly transferred from its normal direct current (DC) power source to its alternate alternating current (AC) power source during the Average Power Range Monitor (APRM) D surveillance test. Specifically, the installed operational barrier failed to protect a fuse block when a test cable connector was inadvertently dropped. This caused a momentary electrical short and resulted in the inverter to Page 4 of 9
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| 2Q/2014 Inspection Findings - Clinton transfer power sources. The licensee documented this issue in the CAP as IR 01476647 and performed (1) a stand-down with instrument maintenance craftsmen to discuss the event and lessons learned, (2) changes to the licensees risk/hazards assessment process to include a checklist designed to aid in challenging jobsite conditions, (3) conduct of paired observations by maintenance department managers on use of the checklist, and (4) a case study with the maintenance shops using this event to highlight determining risk perception and robust protective barriers.
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| The inspectors determined that the licensees failure to adequately manage the risk associated with performance of surveillance testing for APRM D was a performance deficiency. The performance deficiency is more than minor because it was associated with the configuration control attribute of the MS cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The performance deficiency involved the licensees assessment and management of risk associated with performing maintenance in accordance with 10 CFR 50.65(a)(4); therefore the inspectors used IMC 0609, Attachment 4 Initial Characterization of Findings, and Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, and determined that a detailed risk evaluation would be required since the issue represented an actual loss of safety function of a system. The Region III Senior Reactor Analyst (SRA) completed a detailed risk evaluation using the NRCs Standardized Plant Analysis Risk (SPAR) model for Clinton Power Station (CPS), Version 8.17 and SAPHIRE Version 8.09 to calculate an Incremental Core Damage Probability Deficit (ICDPD) for the unevaluated condition. The SRA ran the SPAR model conservatively assuming that High Pressure Core Spray System (HPCS) was unavailable during the 6-hour time. The result was an ICDPD of less than 2E-08/year. In accordance with IMC 0609, Appendix K, because the ICDPD was not greater than 1E 06/year, the finding was determined to be of very low safety significance (i.e., Green). The finding was determined to have a cross cutting aspect in the area of human performance, associated with the work practices component, in that personnel work practices are used commensurate with the risk of the assigned task, such that work activities are performed safely. Specifically, the technicians did not perform adequate self or peer checks after installation of the barrier to ensure the barrier would provide protection from shorting. [H.4(a)]
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| Inspection Report# : 2013005 (pdf)
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| Significance: Dec 19, 2013 Identified By: Self-Revealing Item Type: NCV NonCited Violation Insulation Resistance Testing for Unit Substation Transformers Was Incorrectly Performed A finding of very low safety significance (Green) and associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed from an event that resulted in a reactor scram. Specifically, during troubleshooting of the Unit Substation A transformer failure on December 08, 2013, it was identified that the licensee incorrectly measured the resistance between the transformer windings instead of the winding and ground. The licensee entered this concern into its Corrective Action Program as AR 01594794, and satisfactory re-measured the insulation resistance for the un-faulted transformer 1AP11E.
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| The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of equipment performance and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding screened as very low safety significance (Green), because the inspectors answered NO to all Mitigating Systems Screening questions in Exhibit 2 of Appendix A of IMC 0609. The finding was determined to have a cross-cutting aspect in the area of human performance, associated with the work control component, in that the licensee failed to ensure supervisory and management oversight of work activities, including contractors, such that nuclear safety is supported. H.4(c).
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| Inspection Report# : 2013009 (pdf)
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| Significance: Dec 19, 2013 Identified By: NRC Page 5 of 9
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| 2Q/2014 Inspection Findings - Clinton Item Type: NCV NonCited Violation Inadequate Acceptance Criteria in the Insulation Resistance Test Procedure The inspectors identified a finding of very low safety significance (Green) and associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to have adequate acceptance criteria in testing procedure. Specifically, the minimum acceptable insulation resistance for transformers as specified in Procedure CPS 8440.01 did not meet the minimum vendor recommended values in accordance with the vendor manual. The licensee entered this concern into its Corrective Action Program as IR 01596730 and IR 01598375.
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| The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective of ensuring capability and reliability of systems that respond to initiating events to prevent undesirable consequences. The finding screened as very low safety significance (Green), because the inspectors answered NO to all Mitigating Systems Screening questions in Exhibit 2 of Appendix A of IMC 0609. The inspectors identified the finding had a cross-cutting aspect in the area of problem identification and resolution, associated with the corrective action program component because the licensee failed to ensure issues potentially impacting nuclear safety are promptly identified. (P.1(a))
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| Inspection Report# : 2013009 (pdf)
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| Significance: Sep 30, 2013 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO FOLLOW PROCEDURE AND APPROPRIATELY DOCUMENT BASIS FOR IMMEDIATE OPERABILITY OF THE DIVISION 2 EMERGENCY DIESEL GENERATOR An NRC identified non-cited violation of 10CFR50, Appendix B, Criterion V, Instructions, Procedures and Drawings for the failure to follow procedure OP-AA-108-115, "Operability Determinations", Revision 11, and document the basis that a reasonable expectation of operability existed after an immediate operability determination. Specifically, after the control room received a report of a crack on the after cooler ducting of the Division 2 emergency diesel generator the licensee failed to document their basis that a reasonable expectation of operability existed for the Division 2 emergency diesel generator. The licensee documented this issue in the corrective action prgram as Action Request 015401540.
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| The inspectors determined that the licensee failing to follow the station procedure for operability determinations was a performance deficiency. Specifically, the licensee failed to follow the station procedure for operability determinations and appropriately document the decision and the basis that a reasonable expectation of operability existed for the Division 2 emergency diesel generator. The performance deficiency is more than minor because if immediate operability determination and either the basis that a reasonable expectation of operability exists or the declaration that the system, structure or component is inoperable is not appropriately documented it could lead to a more significant safety concern. Using Manual Chapter 0609, Attachment 4 "Initial Characterization of Findings," and Appendix A "The Significance Determination Process for Findings at Power" the finding was screened against the mitigating systems cornerstone and determined to be of very low safety significance (Green) because the finding was/did not: 1) a deficiency affecting the design or qualification of a mitigating structure, system or component, 2) represent a loss of system and/or function, 3) represent an actual loss of function of a single train for greater than its technical specification allowed outage time, 4) represent an actual loss of function of one or more non-technical specifications trains of equipment designated as high safety-significant for greater than 24 hours and 5) did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding or severe weather event.
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| The finding was determined to have a cross-cutting aspect in the area of human performance, associated with the decision making component, in that the licnesee decisions failed to demonstrate that nuclear safety is an overriding priority. Specifically, the licensee failed to use their systematic process, when faced with an unexpected plant condition of the Division 2 emergency diesel generator to ensure safety was maintained.. H.1(a).
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| Inspection Report# : 2013004 (pdf)
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| Page 6 of 9
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| 2Q/2014 Inspection Findings - Clinton Significance: Aug 30, 2013 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO EVALUATE A DEGRADED/NON-CONFORMING CONDITION ON DIESEL FIRE PUMP.
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| The inspectors identified a finding of very low safety significance associated with the licensee's failure to appropriately evaluate the functionality of the 'B' Diesel Fire Pump (DFP) after identifying a degraded/non-conforming crankcase pressure condition while performing testing on June 13, 2011, and on numerous occasions thereafter, that could have affected the ability of the system to perform a function important to safety. An associated NCV of Clinton Power Station License Condition 2.F was identified. The License Condition required the licensee to implement and maintain in effect all provisions of the approved Fire Protection program as described in the Updated Final Safety Analysis Report (UFSAR). Appendix E, Section 4.0.C.8 of the UFSAR stated that the Clinton Power Station Quality Assurance Program establishes measures for corrective action on conditions adverse to fire protection.
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| Quality Assurance Topical Report (QATR), Chapter 16, Section 2.4 stated that personnel performing the evaluation function of conditions adverse to quality are responsible for considering the cause and the feasibility of corrective action to assure that the necessary quality of an item is not deteriorated. The licnesee entered the issues into the CAP and initiated corrective actions to evaluate the functionality of the 'B' DFP.
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| The failure to correctly evaluate a degraded/non-conforming condition potentially affecting the functionality of structures, systems, and components (SSCs) important to safety would become a more significant safety concern if left uncorrected because it could reasonably result in an unrecognized condition of an SSC failing to fulfill a function important to safety. In addition, the finding was associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of sytems that respond to initiating events to prevent undesirable consequences. Specifically, the degraded condition of high crankcase pressure resulted in repeat operational equipment challenges and extended periods of unavailability of the 'B' DFP. Therefore the finding was of more than minor significance. The finding was a licensee performance deficienty of very low safety significance (Green) because it inolved only a low degradation of the protection against external factors function due to a redundant train that could supply water. The inspectors concluded that this finding affected the cross-cutting area of probelm identification and resolution. Specifically, the licensee failed to thoroughly evaluate problems such that the resolutions addressed causes and extent of condition as necessary for an SSC important to safety when a degraded/non-conforming condtion was identified. [P.1(c)]
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| Inspection Report# : 2013007 (pdf)
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| Barrier Integrity Significance: Dec 31, 2013 Identified By: NRC Item Type: FIN Finding FAILURE TO IDENTIFY EMBEDDED OPERATOR CHALLENGE Inspectors identified a finding of very low safety significance associated with the licensees failure to identify an embedded operator challenge. Specifically, the licensee proceduralized compensatory actions which were necessary in order to maintain a negative pressure (-0.25 in. H2O) inside the fuel building when opening the inner railroad bay door. The licensee documented this issue in the CAP as IR 1589104 and subsequently screened this issue as an operator challenge.
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| The inspectors determined that the licensees failure to identify an embedded operator challenge was a performance deficiency. This finding was more than minor significance because it was associated with the Barrier Integrity Cornerstone attribute of structure, system and component (SSC) and barrier performance, and adversely affected the cornerstone objective to provide reasonable assurance that the physical design barrier of secondary containment Page 7 of 9
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| 2Q/2014 Inspection Findings - Clinton protects the public from radionuclide releases caused by accidents or events. This finding is of very low safety significance due to answering no to all questions under the Barrier Integrity Cornerstone column of IMC 0609, , Phase 1 - Initial Screening and Characterization of Findings. The inspectors concluded that this finding affected the cross-cutting aspect of problem identification and resolution. Specifically, the licensee failed to implement its CAP with a low threshold for identifying issues and did not identify this challenge to operators completely, accurately, and in a timely manner commensurate with its safety significance. [P.1(a)]
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| Inspection Report# : 2013005 (pdf)
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| Emergency Preparedness Occupational Radiation Safety Significance: Dec 31, 2013 Identified By: NRC Item Type: FIN Finding FAILURE TO MAINTAIN RADIATION EXPOSURE ALARA DURING 1R13.
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| Inspectors reviewed a self-revealing finding due to the licensee having unplanned and unintended occupational collective radiation dose because of deficiencies in the licensees Radiological Work Planning and Work Execution Program. Specifically, the licensee failed to properly incorporate as-low-as-reasonably-achievable strategies and insights while planning and executing work activity during the C1R13 refueling outage. During the In-Service Inspection (ISI) examinations performed inside the bio-shield, the dose overage was 28.410 person-rem (68 percent higher than initial estimate). This result was caused by poor radiological planning and work execution of these tasks.
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| The licensee entered this issue into their CAP as IR 01593794 and incorporated the lesson learned into the outage planning.
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| The inspectors determined that the failure to appropriately plan and coordinate outage activities, together with the failure to properly incorporate ALARA strategies or insights while planning and executing ISI examinations inside the bio-shield during the C1R13 refueling outage was a performance deficiency. The finding was more than minor because it was associated with the program and process attribute of the Occupational Radiation Safety Cornerstone.
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| This issue affected the cornerstone objective of ensuring the adequate protection of worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. The finding is also very similar to IMC 0612, Appendix E, Examples of Minor Issues, Example 6.i. This example provides guidance that an issue is not minor if the actual collective dose exceeded 5 person-rem and exceeded the planned, intended dose by more than 50 percent. The inspectors determined that this finding was of very low safety significance because CPSs 3-year rolling average collective was less than the 240 person-rem/unit referenced within IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process. This finding did not have a cross cutting aspect due to not being reflective of current performance as exemplified by improvements in the recently completed C1R14 outage.
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| Inspection Report# : 2013005 (pdf)
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| Page 8 of 9
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| 2Q/2014 Inspection Findings - Clinton Public Radiation Safety Security Although the Security Cornerstone is included in the Reactor Oversight Process assessment program, the Commission has decided that specific information related to findings and performance indicators pertaining to the Security Cornerstone will not be publicly available to ensure that security information is not provided to a possible adversary.
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| Other than the fact that a finding or performance indicator is Green or Greater-Than-Green, security related information will not be displayed on the public web page. Therefore, the cover letters to security inspection reports may be viewed.
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| Miscellaneous Last modified : August 29, 2014 Page 9 of 9
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| 3Q/2014 Inspection Findings - Clinton Clinton 3Q/2014 Plant Inspection Findings Initiating Events Significance: Sep 30, 2014 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO UPDATE THE FINAL SAFETY ANALYSIS REPORT (FSAR) - SD STRUCTURAL INTEGRITY FUNCTION The inspectors identified a Severity Level IV Non-Cited Violation of title 10 Code of Federal Regulations (CFR) 50.71(e), 'Periodic Update of the Final Safety Analysis Report' and an associated Green finding for the licensee's failure to update the Final Safety Analysis Report with a description of the basis for the steam dryer structural integrity submitted to the NRC in support of an extended power uprate license amendment. Specifically, the licensee did not update Section 3.9.5.1.1.9. "Steam Dryers," of the FSAR to include analysis and inspections of the steam dryer each refueling outage that provided the basis for steam dryer structural integrity. Consequently, the licensee had not completed an inspection of the steam dryer during the most recent refueling outage. The licensee entered this issue into the corrective action program as issue report IR 02223135 and initiated actions to evaluate the Final Safety Analysis Report for revision to include description of the structural integrity function of the steam dryer.
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| The inspectors determined that the licensee's failure to update the Final Safety Analysis Report with the basis for steam dryer structural integrity submitted to the NRC was a performance deficiency. the performance deficiency was determined to be more than minor in accordance with Inspection Manual Chapter 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," dated September 7, 2012, because, if left uncorrected, the performance deficiency would have the potential to lead a more significant safety concern and is therefore a finding. Failure to update the Final Safety Analysis Report with the basis for steam dryer structural integrity could result in a failure to maintain the structural integrity of the steam dryer. Specifically, insuffient steam dryer inspections could result in failure to detect structurally significant cracking and result in a steam dryer failure which generates debris that adversely affect the function of safety-related compoments (e.g. MSIVs). Additionally, the failure to update the Final Safety Analysis Report with the basis for steam dryer structural integrity was more than minor because it was associated with the Initiating Events Cornerstone attribute of Equipment Performance and adversely affected the Cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions.
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| Violations of 10 CFR 50.71(e) are dispositioned using the traditional enforcement process because they are considered to be violations that potentially impede or impact the regulatory process. This violation was also associated with a finding that has been evaluated by the significance determination process (SDP) and communicated with SDP color reflective of the safety impact of the deficient licensee performance. The SDP, however, does not specifically consider regulatory process impact. Thus, although related to a common regulatory concern, it is necessary to address the violation and finding using different processes to correctly reflect both the regulatory importance of the violation and the safety significance of the associated finding.
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| Using Manual Chapter 0609, Attachment 4 "Initial characterization of Findings," and Appendix A "The Significance Determination Process for findings at Power" the finding was screened against the initiating events cornerstone and determined to be of very low safety significance (Green) because the finding did not cause a reactor trip and the loss of mitigating equipment relied upon to transition from the onset of the trip to a stable. The performance deficiency associated with this finding did not reflect current licensee performance; therefore, no cross cutting aspect was identified with this finding.
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| Page 1 of 13
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| 3Q/2014 Inspection Findings - Clinton Additionally, in accordance with Section 6.1.d.3 of the NRC Enforcement Policy, this violation was categorized as Severity Level IV because the licensee's failure to update the FSAR as required by 10 CFR 50.71(e) had not yet resulted in any unacceptable change to the facility or procedures.
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| Inspection Report# : 2014004 (pdf)
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| Significance: Sep 30, 2014 Identified By: NRC Item Type: NCV NonCited Violation MODIFICATION TO STEAM DRYER TIE BARS 28 AND 30 WITHOUT A 10 CFR 50.59 SAFETY EVALUATION The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.59(d)(1), "Changes, Test, and Experiments" for the licensee's failure to perform a written evaluation, which provided the bases for the determination that a change did not require a license amendment. Specifically, the licensee made a change pursuant to 10 CFR 50.59 (c) with the installation of 1/2 inch holes adjacent to welds attaching tie bars 28 and 30 to the steam dryer vane assembly and did not provide a written evaluation to provide a basis for the determination that this change would not result in a more than minimal increase in the likelihood of occurrence of a malfunction of an system structure or component important to safety (e.g. MSIVs). The licensee entered this finding into the corrective action program as issue report IR 02223135 and identified an action to secure a detailed assessment of these degraded tie bar locations from the steam dryer vendor. The licensee also consulted with the steam dryer vendor and made a qualitative assessment that the additional unflawed and unaltered portion of the fillet welds present at the end of the tie bar 28 and 30 locations provided a reasonable basis to conclude that these tie bars would not fail and affect the operability of safety-related components.
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| The inspectors determined that the failure to provide a written evaluation, which provided the basis for the determination that a change did not require a license amendment, was a performance deficiency. Specifically, the licensee failed to provide a basis for not applying for a license amendment associated with increased likelihood of a SD failure that impacts safety-related equipment due to reduced structural support available at tie bars 28 and 30. The performance deficiency was determined to be more than minor in accordance with IMC 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," dated September 7, 2012, because it was associated with the Initiating Events cornerstone attribute of equipment performance and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. In addition, the associated violation was determined to be more than minor because the inspectors could not reasonable determine if the changes to the SD at tie bars 28 and 30 would have required NRC prior approval.
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| Violations of 10 CFR 50.59 are dispositioned using the traditional enforcement process instead of the SDP because they are considered to be violations that potentially impede or impact the regulatory process. However, if possible, the underlying technical issue is evaluated under the SDP to determine the severity of the violation. In this case, the inspectors used Manual Chapter 0609, Attachment 4 "Initial characterization of Findings," and Appendix A "The Significance Determination Process for findings a Power" the finding was screened against the initiating events cornerstone and determined to be of very low safety significance (Green) because the finding did not cause a reactor trip and the loss of mitigating equipment relied upon to transition from the onset of the trip to a stable. The performance deficiency associated with this finding did not reflect current licensee performance; therefore, no cross cutting aspect was identified with this finding.
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| In accordance with Section 6.1.d.2 of the NRC Enforcement Policy, this violation was categorized as Severity Level IV because the resulting changes were evaluated by the SDP as having very low safety significance.
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| Inspection Report# : 2014004 (pdf)
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| Page 2 of 13
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| 3Q/2014 Inspection Findings - Clinton Significance: Jul 11, 2014 Identified By: NRC Item Type: FIN Finding FAILURE TO IDENTIFY A LEVEL 1 TEST CRITERION FAILURE The inspectors documented a self-revealing Green finding associated with the failure to follow procedures when performing power ascension testing on the digital feedwater (DFW) system. Specifically, contrary to procedure CPS 2894.01, "Digital FWLC [feedwater level control system] Modifications Test - Power Ascension," Section 9.1, the licensee did not declare a Level 1 criterion failure when unacceptable oscillations were noted during a transition in the power ascension test. This resulted in the licensee declaring the test successful and returning the system to service without taking the appropriate corrective actions to address the oscillations. This contributed to the subsequent scram caused by reactor water level oscillations.
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| The failure to follow procedures when performing power ascension testing on the digital feedwater system was a performance deficiency. The performance deficiency was more than minor because it was associated with the design control attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations and is therefore a finding. Using IMC 0609, Attachment 4 "Initial Characterization of Findings," and Appendix A "The Significance Determination Process for Findings at Power," issued June 19, 2012, the finding was determined to be of very low safety significance (Green) because it did not cause a reactor trip with a coincident loss of mitigating equipment. The inspectors determined this finding affected the conservative bias aspect of the of human performance cross-cutting area described as being present when the organization uses decision making practices that emphasize prudent choices over those that are simply allowable. Specifically, the licensee used non-conservative assumptions when determining whether the condition identified during the power ascension test was allowable (H.14).
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| This finding does not involve enforcement action because no violation of regulatory requirements was identified.
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| Inspection Report# : 2014008 (pdf)
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| Significance: Jun 30, 2014 Identified By: NRC Item Type: FIN Finding ELECTRO HYDRAULIC CONTROL SYSTEM LEAK RESULTS IN MANUAL SCRAM The inspectors documented a self-revealing, Green finding associated with a failure to provide adequate work instructions to perform repairs to the shutoff valve for 1TGCV4 main turbine control valve. Specifically, contrary to station procedure MA-AA-716-010, Maintenance Planning, Revision 21, the work instructions generated to install the shutoff valve failed to reference the appropriate cap screw size, lubricate the cap screws and install lock washers on the cap screws used to attach the shut off valve to the control valve. This allowed the cap screws to loosen and ultimately fail due to fatigue resulting in a leak of electro hydraulic control fluid of sufficient rate to require a manual scram of Unit 1 on April 26, 2013. The valve was replaced and successfully tested and the unit was restarted. The licensee documented this issue in the corrective action program (CAP) as Issue Report (IR) 01506929.
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| The performance deficiency was more than minor because it was associated with the procedure quality attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations and is therefore a finding. Using Manual Chapter 0609, Attachment 4 Initial Characterization of Findings, and Appendix A The Significance Determination Process for Findings at Power, issued June 19, 2012, the finding was screened against the initiating events cornerstone and determined to be of very low safety significance (Green) because the finding did not cause a reactor trip with a coincident loss of mitigating equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The inspectors determined that no cross cutting aspect will be assigned to this performance deficiency since it occurred in 2008 and is not indicative of current plant performance.
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| Page 3 of 13
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| 3Q/2014 Inspection Findings - Clinton Inspection Report# : 2014003 (pdf)
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| Significance: Jun 30, 2014 Identified By: NRC Item Type: FIN Finding FAILURE TO IMPLEMENT ENGINEERING CHANGE RESULTS IN MANUAL REACTOR SCRAM The inspectors documented a self-revealing, Green finding associated with a failure to implement engineering change (EC) 380150 Upgrade Feed Water Level Control and Turbine Speed. Specifically, contrary to station procedure CC-AA-256, Process for Managing Plant Modifications Involving Microprocessor Technology, Revision 2, the licensee failed to identify, evaluate and mitigate software component critical parameters in the engineering change that installed the digital feed water system. This resulted in nonlinear reactor water level oscillations when transferring from the motor driven feed pump to the turbine driven feed pump that required the reactor operator to manually scram the reactor prior to reaching the level 8 automatic scram set point. The licensee documented this issue in the corrective action program as IR 1596987.
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| The performance deficiency was more than minor because it was associated with the design control attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations and is therefore a finding. Using Manual Chapter 0609, Attachment 4 Initial Characterization of Findings, and Appendix A The Significance Determination Process for Findings at Power, issued June 19, 2012, the finding was screened against the initiating events cornerstone and determined to be of very low safety significance (Green) because the finding did not cause a reactor trip with a coincident loss of mitigating equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The inspectors determined this finding affected the cross cutting area of human performance in the aspect of documentation where the organization creates and maintains complete, accurate and up-to date documentation. Specifically, the contractors failed to create complete documentation to be use by the licensee when evaluating the critical parameters.
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| Inspection Report# : 2014003 (pdf)
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| Significance: Feb 14, 2014 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct Identified Combustibles The inspectors identified a finding of very low safety significance and associated NCV of License Condition 2.F for the failure to remove an identified combustible. Specifically, the failure to remove a piece of wood located directly under a safety-related cable tray for a period in excess of three years was a failure to take corrective action as required by the licensees Quality Assurance Program. The licensee entered the issue into their Corrective Action Program and removed the piece of wood by the end of the inspection.
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| The finding was determined to be more than minor because the combustible material was located directly beneath a safety-related cable tray and, as such, represented a credible fire scenario. The finding was determined to be of very low safety significance (i.e., Green) because the impact of the fire would be largely limited to one train/division of equipment important to safety. The inspectors determined that the finding has a cross-cutting aspect in the area of human performance because the licensee did not ensure sufficient resources were available to support nuclear safety.
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| Specifically, the failure to remove the identified combustible was due to a lack of resources to schedule and accomplish removing the material.
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| Inspection Report# : 2014007 (pdf)
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| Page 4 of 13
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| 3Q/2014 Inspection Findings - Clinton Mitigating Systems Significance: Sep 30, 2014 Identified By: NRC Item Type: NCV NonCited Violation EXCEEDED TECHNICAL SPECIFICATION ALLOWED OUTAGE TIME FOR ELECTRICAL POWER SYSTEMS DUE TO AUXILIARY EQUIPMENT OUT OF SERVICE The inspectors identified a non-cited violation of Technical Specification 3.8.4, "DC Sources - Operating" and Technical Specification 3.8.9, "Distribution Systems - Operating" for failing to enter the technical specifications and complete the associated actions prior to the completion time when auxiliary equipment required to support electrical power system safety function was out of service. Specifically, the licensee removed the division 1 safety related portion of the switchgear cooling system from service to perform maintenance and failed to enter the applicable technical specifications that the was required to support system safety function. The licensee documented this issue in the corrective action program as Issue Report (IR) 01674754.
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| The failure to enter the technical specifications and complete the associated actions prior to the completion time when auxiliary equipment required to support electrical power system safety function was out of service was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences and is therefore a finding. Using Manual Chapter 0609, Appendix A, "The Siginificance Determination Process (SDP) for Findings At-Power," issued June 19, 2012, Exhibit 2 for the Mitigating Systems Cornerstone. The inspectors answered "Yes" to the screening question under the Mitigating Systems Cornerstone "Does the finding represent and actual loss of function of at least a single train for > its Tech Spec Allowed Outage Time OR two separate safety systems out-of-service for > its Tech Spec Allowed Outage Time?,' since the finding represented an actual loss of function of at least a single Train for > its Tech Spec Allowed Outage time. Therefore, a detailed risk evaluation was performed using IMC 0609, Appendix A. The Senior Reactor Analysts (SRAs) evaluated the finding using the Clinton Standardized Plant Analysis Risk (SPAR) model version 8.17, Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) version 8.1.0. For switchgear cooling, independent redundant cooling trains are provided for each of the three divisional switchgear areas with one train being non-safety related and the other safety related. In order to characterize the risk significance, the SRAs assumed that during a loss of offsite power (LOOP) event, the non-safety related switchgear cooling train that is normally in operation would become unavailable. The safety-related cooling train, should it be undergoing maintenance, would be unavailable as well. The exposure time for this issue was taken to be 235 hours based on the licensee documentation. Post-processing rules were used to credit an additional 4.0 hours of time to recover offsite power (to allow recovery of the non-safety cooling train) in core damage sequences when the safety-related cooling train for Division 1 equipment was undergoing maintenance during a LOOP. The SRAs also gave credit in the SPAR Model for local operator action to provide alternate switchgear room cooling during a LOOP. The licensee produced Alarm Response Procedure CPS 5050.03, Rev 30c, which directed operators to Procedure CPS 3412.01, "Essential Switchgear Heat Removal (VX)...,"
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| Revision 15. These procedures directed operators to locally open doors, set up protable blowers, or lower electrical loads to help cool the room as necessary. The SRAs used the SPAR-H Human Reliability Analysis Method (NUREG/CR-6883) to estimate the human error probability for identifying and executing the local actions. the performance drivers were "time" (extra time) and "stress" (high) for diagnosis. The performance drivers were "stress" (high) and "ergonomics" (poor) for action. The resultant human error probability using these assumptions was 0.022. Using the above information, the ?CDF during the exposure time is 1.7E-08/yr. The dominant sequences were station blackout sequences, with initial success of RCIC and HPCS, but later failure of those systems and decay heat removal and all injection due to failure to vent containment and its subsequent failure. Based on the detailed risk evaluation, this finding is best characterized as a finding of very low safety-significance (Green.) The inspectors Page 5 of 13
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| 3Q/2014 Inspection Findings - Clinton determined this finding affected the cross-cutting area of human performance in the aspect of avoid complacency where individuals recognize and plan for mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Specifically, the licensee has removed the division 1 or 2 safety related switchgear cooling system fans or condensing units from service numerous times and failed to consider the components inoperable under technical specification definition for operable. (IMC 0310 H.12)
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| Inspection Report# : 2014004 (pdf)
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| Significance: Sep 30, 2014 Identified By: NRC Item Type: NCV NonCited Violation PROGRAMMATIC FAILURE TO COMPLETE OPERABILITY AND FUNCTIONALITY DETERMINATIONS The inspectors identified a non-cited violation of 10 CFR 50 Appendix B, Criterion V, Instructions, Procedures and Drawings, "Procedures," for the failure to accomplish station procedure OP-AA-108-115, "Operability Determinations" Revision 14. Specifically, on multiple occasions operations personnel failed to complete or documented incomplete operability or functionality of safety related or related to safety equipment used at the site.
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| The licensee documented this issue in the corrective action program as Issue Report (IR) 01693256.
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| The failure to complete or provided incomplete operability or functionality determinations used to determine the operability or functionality of safety related or related to safety equipment used at the site is a performance deficiency.
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| The performance deficiency was determined to be more than minor because if left uncorrected, the performance deficiency has the potential to lead to a more significant safety concern and is therefore a finding. Specifically, if operations personnel continue to fail to complete or provide incomplete operability or functionality determination the station could have safety or safety related equipment inoperable without taking appropriate actions for the equipment being inoperable (e.g. entering appropriate technical specification limited condition for operation). Using Manual Chapter 0609, Attachment 4 "Initial Characterization of Findings," and Appendix A "The Significance Determination Process for Findings at Power" the finding was screened against the mitigatins systems cornerstone and determined to be of very low safety significance (Green) because the finding was/did not: 1) a deficiency affecting the design or qualification of a mitigating structure, system or component, 2) represent a loss of system and/or function, 3) represent an actual loss of function of a single train for greater than its technical specification allowed outage time, 4) represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant for greater than 24 hours and 5) did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding or serve weather event. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of Training, where the organization provides training and ensures knowledge transfer to maintain a knowledgeable, technically competent workforce and instill nuclear safety values. Specifically, personnel performing the reviews believed existing training provided sufficient knowledge without the use of additional resources material and current training to operators does not covet this activity. (IMC 0319 H.9)
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| Inspection Report# : 2014004 (pdf)
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| Significance: Sep 30, 2014 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ESTABLISH SURVEILLANCE PRODEDURE FOR REACTOR CORE ISOLATION COOLING PUMP DUE TO UNACCEPTABLE PRECONDITIONING The inspectors determined that the failure to establish a surveillance procedure to test the RCIC system due to unacceptable preconditioning is a performance deficiency. The performance deficiency was more than minor because it was associated with the procedure quality attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability and capability to response to initiating events to prevent Page 6 of 13
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| 3Q/2014 Inspection Findings - Clinton undesirable consequences and is therefore a finding. Using Manual Chapter 0609, Attachment 0609.04 "Initial Characterization of Findings," and Appendix A "The Significance Determination Process for Findings at Power" the finding was screened against the mitigating systems cornerstone and determined to be of very low safety significance (Green) because the finding was/did not: 1) a deficiency affecting the design or qualification of a mitigating structure, system or component, 2) represent a loss of system and/or function, 3) represent an actual loss of function of a single train for greater than its technical specification allowed outage time, 4) represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant for greater than 24 hours and 5) did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding or severe weather event. The inspectors determined this finding affected the cross-cutting area of problem identification and resolution in the aspect of operating experience where the organization systematically and effectively collects, evaluates and implements relevant internal and external operating experience in a timely manner.
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| Specifically, the licensee considered the impact of the operating experience for surveillance testing, but did not consider its impact during normal plant operation. (IMC 0310 P.5)
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| Inspection Report# : 2014004 (pdf)
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| Significance: Jun 30, 2014 Identified By: NRC Item Type: NCV NonCited Violation FOREIGN MATERIAL IN RELAY PREVENTS EMERGENCY DIESEL GENERATOR OUTPUT BREAKER FROM CLOSING The inspectors documented a self-revealing, Green non-cited violation of Clinton Power Station Technical Specification 5.4.1, Procedures, for a failure to prevent foreign material from entering a relay associated with the Division 1 Diesel Generator. Specifically, contrary to station procedure CPS 8501.05, CV-2 Relay Inspection and Calibration with Doble Test Equipment, Revision 4, the licensee failed to verify that relay 227-DGIKA, CV-2 AB phase was clean and free of all foreign material. The foreign material prevented the relay from operating and satisfying the permissive logic required to close the Division 1 Diesel Generator output breaker resulting in having to declare the Diesel Generator inoperable. The relay was replaced and successfully tested and the licensee documented this issue in the corrective action program as IR 01600935.
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| The finding was more than minor because it was associated with the procedure quality attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences and is therefore a finding.
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| Using Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012, Exhibit 2 for the Mitigating Systems Cornerstone, the inspectors answered Yes to the screening question under the Mitigating Systems Cornerstone Does the finding represent an actual loss of function of at least a single Train for > its Tech Spec Allowed Outage Time OR two separate safety systems out-of- service for >
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| its Tech Spec Allowed Outage Time?," since the finding represented an actual loss of function of at least a single Train for > its Tech Spec Allowed Outage Time of 14 days. Therefore, a detailed risk evaluation was performed using IMC 0609, Appendix A. The Senior Reactor Analysts (SRAs) evaluated the finding using the Clinton Standardized Plant Analysis Risk (SPAR) model version 8.17, Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) version 8.1.0 and concluded that the risk increase to the plant due to this finding is very low (Green). The inspectors determined this finding affected the cross cutting area of human performance in the aspect of work management where the organization implements a process of planning, controlling and executing work activities such that nuclear safety is the overriding priority. Specifically, the licensees implementation of their foreign material exclusion process for this maintenance activity lacked sufficient planning, controls and execution to prevent foreign material from entering a risk significant piece of safety related equipment.
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| Inspection Report# : 2014003 (pdf)
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| Page 7 of 13
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| 3Q/2014 Inspection Findings - Clinton Significance: Jun 30, 2014 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO DEVELOP ADEQUATE PROCEDURES FOR PRE-PLANNING AND PERFORMING MAINTENANCE AFFECTING SAFETY-RELATED EQUIPMENT The inspectors documented a self-revealing, Green non-cited violation (NCV) of Clinton Power Station Technical Specification 5.4.1, Procedures for a failure to develop adequate procedures for properly pre-planning and performing maintenance affecting the performance of safety-related equipment which resulted in the subsequent failure of the Division 3 Diesel Room Ventilation damper hydramotor on August 15, 2013. Specifically, during pre-scheduled performance testing of the Division 3 (High Pressure Core Spray System) Emergency Diesel Generator Room Ventilation Damper hydramotor, the ventilation supply air intake damper, 1VD01YC, failed to open as a result of Damper Hydramotor 1TZVD003A experiencing an age-related degradation failure. This was due in part to the licensees failure to properly pre-plan and perform the appropriate preventive maintenance for the hydramotor due to inadequate procedures. Procedure WC-AA-113, Predefine Database Revisions, Revision 2, did not provide adequate instructions appropriate to the circumstances to properly pre-plan and perform maintenance affecting the performance of safety-related equipment. This resulted in a loss of safety function of the HPCS Diesel Generator and its supported High Pressure Core Spray system because of the low confidence that diesel room temperature would be maintained to support the diesel during an event when it would be required to perform its function. The licensee subsequently replaced the hydramotor, tested the new hydramotor successfully and restored the diesel ventilation system to operable. They documented this issue in the corrective action program as IR 1546973 and IR 1547294.
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| The performance deficiency was more than minor because it was associated with the equipment performance attribute of the mitigating systems cornerstone attribute and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences and is therefore a finding. Using Manual Chapter 0609, Appendix A, The SDP for Findings At-Power, issued June 19, 2012, Exhibit 2 for the Mitigating Systems Cornerstone. The inspectors answered Yes to the screening question under the Mitigating Screening Cornerstone Does the finding represent a loss of system and/or function? since the finding resulted in a loss of safety function. Therefore, a detailed risk evaluation was performed using IMC 0609, Appendix A. The SRAs evaluated the finding using the Clinton SPAR model version 8.17, SAPHIRE version 8.1.0 and concluded that the risk increase to the plant due to this finding is very low (Green). The inspectors determined that no cross-cutting aspect will be assigned to this performance deficiency since it occurred in 2005 and is not indicative of current plant performance Inspection Report# : 2014003 (pdf)
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| Significance: Dec 31, 2013 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IMPLEMENT REQUIREMENTS OF STATION SCAFFOLD INSTALLATION PROCEDURE.
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| Inspectors identified a NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures and Drawings for the failure to follow station procedure MA AA-796-024, Scaffold Installation, Inspection, and Removal, Revision 8, to obtain engineering approval for seismic scaffolds not complying with specific requirements of approved station procedures during the C1R14 outage. Specifically, seismic scaffolds identified during walkdowns by the inspectors did not meet procedural requirements for required clearances from or tie off to safety-related components and did not have the required engineering evaluation and approval for acceptability. The licensee documented this issue in the corrective action program (CAP) as Issue Report (IR) 01574003 and completed the required engineering review and approval.
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| The inspectors determined that the licensees failure to follow the station procedure for scaffold installation, Page 8 of 13
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| 3Q/2014 Inspection Findings - Clinton inspection, and removal was a performance deficiency. The performance deficiency is more than minor because it was associated with the protection against external factors attribute of the Mitigating Systems (MS) cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Using IMC 0609, Attachment 4 Initial Characterization of Findings, and Appendix G Shutdown Operations Significance Determination Process, the finding was screened against Attachment 1, Checklist 8 and found to be of very low safety significance (Green) because the finding did not: 1) increase the likelihood of a loss of reactor coolant system (RCS) inventory, 2) degrade the licensees ability to terminate a leak path or add RCS inventory when needed, 3) significantly degrade the licensees ability to recover decay heat removal once it is lost, 4) result in one or less safety relief valves being available to establish a heat removal path to the suppression pool with the vessel head on. The finding was determined to have a cross-cutting aspect in the area of human performance, associated with the resources component, in that the licensee ensures that personnel, equipment, procedures and other resources are available and adequate to assure nuclear safety. Specifically, the licensee failed to ensure that the scaffold coordinator and superintendents had the required training to assure nuclear safety while erecting seismic scaffolds. [H.2(b)]
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| Inspection Report# : 2013005 (pdf)
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| Significance: Dec 31, 2013 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ASSESS AND MANAGE RISK ASSOCIATED WITH THE PERFORMANCE OF SURVEILLANCE TESTING ON AVERAGE POWER RANGE MONITORS Inspectors reviewed a self-revealing NCV of 10 CFR 50.65(a)(4) for failing to manage risk when the Division 4 Nuclear System Protection System (NSPS) inverter unexpectedly transferred from its normal direct current (DC) power source to its alternate alternating current (AC) power source during the Average Power Range Monitor (APRM) D surveillance test. Specifically, the installed operational barrier failed to protect a fuse block when a test cable connector was inadvertently dropped. This caused a momentary electrical short and resulted in the inverter to transfer power sources. The licensee documented this issue in the CAP as IR 01476647 and performed (1) a stand-down with instrument maintenance craftsmen to discuss the event and lessons learned, (2) changes to the licensees risk/hazards assessment process to include a checklist designed to aid in challenging jobsite conditions, (3) conduct of paired observations by maintenance department managers on use of the checklist, and (4) a case study with the maintenance shops using this event to highlight determining risk perception and robust protective barriers.
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| The inspectors determined that the licensees failure to adequately manage the risk associated with performance of surveillance testing for APRM D was a performance deficiency. The performance deficiency is more than minor because it was associated with the configuration control attribute of the MS cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The performance deficiency involved the licensees assessment and management of risk associated with performing maintenance in accordance with 10 CFR 50.65(a)(4); therefore the inspectors used IMC 0609, Attachment 4 Initial Characterization of Findings, and Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, and determined that a detailed risk evaluation would be required since the issue represented an actual loss of safety function of a system. The Region III Senior Reactor Analyst (SRA) completed a detailed risk evaluation using the NRCs Standardized Plant Analysis Risk (SPAR) model for Clinton Power Station (CPS), Version 8.17 and SAPHIRE Version 8.09 to calculate an Incremental Core Damage Probability Deficit (ICDPD) for the unevaluated condition. The SRA ran the SPAR model conservatively assuming that High Pressure Core Spray System (HPCS) was unavailable during the 6-hour time. The result was an ICDPD of less than 2E-08/year. In accordance with IMC 0609, Appendix K, because the ICDPD was not greater than 1E 06/year, the finding was determined to be of very low safety significance (i.e., Green). The finding was determined to have a cross cutting aspect in the area of human performance, associated with the work practices component, in that personnel work practices are used commensurate with the risk of the assigned task, such that work activities are performed safely. Specifically, the technicians did not perform adequate self or peer checks after Page 9 of 13
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| 3Q/2014 Inspection Findings - Clinton installation of the barrier to ensure the barrier would provide protection from shorting. [H.4(a)]
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| Inspection Report# : 2013005 (pdf)
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| Significance: Dec 19, 2013 Identified By: Self-Revealing Item Type: NCV NonCited Violation Insulation Resistance Testing for Unit Substation Transformers Was Incorrectly Performed A finding of very low safety significance (Green) and associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed from an event that resulted in a reactor scram. Specifically, during troubleshooting of the Unit Substation A transformer failure on December 08, 2013, it was identified that the licensee incorrectly measured the resistance between the transformer windings instead of the winding and ground. The licensee entered this concern into its Corrective Action Program as AR 01594794, and satisfactory re-measured the insulation resistance for the un-faulted transformer 1AP11E.
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| The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of equipment performance and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding screened as very low safety significance (Green), because the inspectors answered NO to all Mitigating Systems Screening questions in Exhibit 2 of Appendix A of IMC 0609. The finding was determined to have a cross-cutting aspect in the area of human performance, associated with the work control component, in that the licensee failed to ensure supervisory and management oversight of work activities, including contractors, such that nuclear safety is supported. H.4(c).
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| Inspection Report# : 2013009 (pdf)
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| Significance: Dec 19, 2013 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Acceptance Criteria in the Insulation Resistance Test Procedure The inspectors identified a finding of very low safety significance (Green) and associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to have adequate acceptance criteria in testing procedure. Specifically, the minimum acceptable insulation resistance for transformers as specified in Procedure CPS 8440.01 did not meet the minimum vendor recommended values in accordance with the vendor manual. The licensee entered this concern into its Corrective Action Program as IR 01596730 and IR 01598375.
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| The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective of ensuring capability and reliability of systems that respond to initiating events to prevent undesirable consequences. The finding screened as very low safety significance (Green), because the inspectors answered NO to all Mitigating Systems Screening questions in Exhibit 2 of Appendix A of IMC 0609. The inspectors identified the finding had a cross-cutting aspect in the area of problem identification and resolution, associated with the corrective action program component because the licensee failed to ensure issues potentially impacting nuclear safety are promptly identified. (P.1(a))
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| Inspection Report# : 2013009 (pdf)
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| Barrier Integrity Page 10 of 13
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| 3Q/2014 Inspection Findings - Clinton Significance: Dec 31, 2013 Identified By: NRC Item Type: FIN Finding FAILURE TO IDENTIFY EMBEDDED OPERATOR CHALLENGE Inspectors identified a finding of very low safety significance associated with the licensees failure to identify an embedded operator challenge. Specifically, the licensee proceduralized compensatory actions which were necessary in order to maintain a negative pressure (-0.25 in. H2O) inside the fuel building when opening the inner railroad bay door. The licensee documented this issue in the CAP as IR 1589104 and subsequently screened this issue as an operator challenge.
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| The inspectors determined that the licensees failure to identify an embedded operator challenge was a performance deficiency. This finding was more than minor significance because it was associated with the Barrier Integrity Cornerstone attribute of structure, system and component (SSC) and barrier performance, and adversely affected the cornerstone objective to provide reasonable assurance that the physical design barrier of secondary containment protects the public from radionuclide releases caused by accidents or events. This finding is of very low safety significance due to answering no to all questions under the Barrier Integrity Cornerstone column of IMC 0609, , Phase 1 - Initial Screening and Characterization of Findings. The inspectors concluded that this finding affected the cross-cutting aspect of problem identification and resolution. Specifically, the licensee failed to implement its CAP with a low threshold for identifying issues and did not identify this challenge to operators completely, accurately, and in a timely manner commensurate with its safety significance. [P.1(a)]
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| Inspection Report# : 2013005 (pdf)
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| Emergency Preparedness Significance: Sep 30, 2014 Identified By: NRC Item Type: NCV NonCited Violation INCOMPLETE EVACUATION TIME ESTIMATE SUBMITTALS The inspectors determined that Exelon's failure to submit a complete updated ETE for the Clinton Power Station by December 22, 2012 was a performance deficiency. Specifically, the ETE is an input into the development of protective action strategies prior to an accident and to the protective action recommendation decision making process during an accident. Inadequate ETEs have the potential to reduce the effectiveness of public protective actions implemented by the OROs. The performance deficiency was more than minor because it was associated with the procedure quality attribute of the emergency preparedness cornerstone and adversely affected the cornerstone objective to ensure that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency and is therefore a finding. Using IMC 0609, attachment 0609.04 "Initial Characterization of Findings," and Appendix B, "Emergency Preparedenss (EP) Significance Determination Process (SDP)," the finding was screened by the inspectors and determined to be of very low safety significance (Green) based upon the following. The performance deficiency was associated with planning standard 10 CFR 50.47 (b)(10)," Green Finding column, provides the following exasmples "ETEs and updates to the ETEs were not provided to responsible OROs," and "The current public protective action strategies documented in emergency preparedness implementing procedures (EPIPs) are not consistent with the current ETE." The inspectors concluded that the incomplete updated ETE delayed the NRC's approval of the Clinton Power Station ETE, therefore the ETE was not provided to the site OROs nor was it used to inform the site EPIPs as required by 10 CFR 50.47(b)(10), and Section IV, Paragraph 4 of Appendix E to 10 CFR Part 50. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of documentation where the organization creates and maintains complete, Page 11 of 13
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| 3Q/2014 Inspection Findings - Clinton accurate and up-to-date documentation. Specifically, the Emergency Preparedness organization did not develop the Clinton Power Station ETE as required by the new regulation introduced by the NRC's EP Rule. (IMC 0310 H.7)
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| Inspection Report# : 2014004 (pdf)
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| Occupational Radiation Safety Significance: Dec 31, 2013 Identified By: NRC Item Type: FIN Finding FAILURE TO MAINTAIN RADIATION EXPOSURE ALARA DURING 1R13.
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| Inspectors reviewed a self-revealing finding due to the licensee having unplanned and unintended occupational collective radiation dose because of deficiencies in the licensees Radiological Work Planning and Work Execution Program. Specifically, the licensee failed to properly incorporate as-low-as-reasonably-achievable strategies and insights while planning and executing work activity during the C1R13 refueling outage. During the In-Service Inspection (ISI) examinations performed inside the bio-shield, the dose overage was 28.410 person-rem (68 percent higher than initial estimate). This result was caused by poor radiological planning and work execution of these tasks.
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| The licensee entered this issue into their CAP as IR 01593794 and incorporated the lesson learned into the outage planning.
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| The inspectors determined that the failure to appropriately plan and coordinate outage activities, together with the failure to properly incorporate ALARA strategies or insights while planning and executing ISI examinations inside the bio-shield during the C1R13 refueling outage was a performance deficiency. The finding was more than minor because it was associated with the program and process attribute of the Occupational Radiation Safety Cornerstone.
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| This issue affected the cornerstone objective of ensuring the adequate protection of worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. The finding is also very similar to IMC 0612, Appendix E, Examples of Minor Issues, Example 6.i. This example provides guidance that an issue is not minor if the actual collective dose exceeded 5 person-rem and exceeded the planned, intended dose by more than 50 percent. The inspectors determined that this finding was of very low safety significance because CPSs 3-year rolling average collective was less than the 240 person-rem/unit referenced within IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process. This finding did not have a cross cutting aspect due to not being reflective of current performance as exemplified by improvements in the recently completed C1R14 outage.
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| Inspection Report# : 2013005 (pdf)
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| Public Radiation Safety Security Although the Security Cornerstone is included in the Reactor Oversight Process assessment program, the Commission has decided that specific information related to findings and performance indicators pertaining to the Security Cornerstone will not be publicly available to ensure that security information is not provided to a possible adversary.
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| Other than the fact that a finding or performance indicator is Green or Greater-Than-Green, security related Page 12 of 13
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| 3Q/2014 Inspection Findings - Clinton information will not be displayed on the public web page. Therefore, the cover letters to security inspection reports may be viewed.
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| Miscellaneous Last modified : November 26, 2014 Page 13 of 13
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| 4Q/2014 Inspection Findings - Clinton Clinton 4Q/2014 Plant Inspection Findings Initiating Events Significance: Dec 31, 2014 Identified By: NRC Item Type: NCV Non-Cited Violation STATION PROCEDURES FAILED TO PROVIDE CONTROLS FOR MATERIAL NEAR TRANSFORMERS The inspectors identified a non-cited violation associated with a failure to provide controls for material near the station transformers. Specifically, station procedure CPS 4302.01, "Tornado/High Winds", Revision 21b does not include guidelines or examples of the types of materials to control as potential missiles in high velicity winds or tornadoes, and does not include triggers to perform walkdowns when high winds are predicted, prior to off-normal entry, to control material adjacent to the offsite power transformers that could result in the loss of offsite power. The licensee entered this issue into the corrective action program as action request (AR) 2388608.
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| The failure to provide guidelines or examples of the types of materials to control as potential missiles in high velicity winds or tornadoes and provide triggers to perform walkdowns when high winds are predicted was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations and is therefore a finding. Using Manual Chapter 0609, Attachment 4 "Initial Characterization of Findings," and Appendix A "The Significance Determination Process for Findings at Power", issued June 19, 2012, the finding was screened against the initiating events cornerstone and determined to be of very low safety significance (Green) because the finding did not involve the complete or partial loss of a support system that contributes to the likelihood of, or caused, an initiating event and did not affected mitigation equipment.
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| The inspectors determined this finding affected the cross-cutting area of problem identification and resolution in the aspect of operating experience where the organization systematically and effectively collects, evaluates, and implements relevant internal and external operating experience in a timely manner. Specifically, the licensee opearting experience program failed to ensure evaluation and implementation of interal operating experience in a timely manner after previous identification in the corrective action progrma. (IMC 0310 P.5)
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| Inspection Report# : 2014005 (pdf)
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| Significance: Sep 30, 2014 Identified By: NRC Item Type: NCV Non-Cited Violation FAILURE TO UPDATE THE FINAL SAFETY ANALYSIS REPORT (FSAR) - SD STRUCTURAL INTEGRITY FUNCTION The inspectors identified a Severity Level IV Non-Cited Violation of title 10 Code of Federal Regulations (CFR) 50.71(e), 'Periodic Update of the Final Safety Analysis Report' and an associated Green finding for the licensee's failure to update the Final Safety Analysis Report with a description of the basis for the steam dryer structural integrity submitted to the NRC in support of an extended power uprate license amendment. Specifically, the licensee did not update Section 3.9.5.1.1.9. "Steam Dryers," of the FSAR to include analysis and inspections of the steam dryer Page 1 of 12
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| 4Q/2014 Inspection Findings - Clinton each refueling outage that provided the basis for steam dryer structural integrity. Consequently, the licensee had not completed an inspection of the steam dryer during the most recent refueling outage. The licensee entered this issue into the corrective action program as issue report IR 02223135 and initiated actions to evaluate the Final Safety Analysis Report for revision to include description of the structural integrity function of the steam dryer.
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| The inspectors determined that the licensee's failure to update the Final Safety Analysis Report with the basis for steam dryer structural integrity submitted to the NRC was a performance deficiency. the performance deficiency was determined to be more than minor in accordance with Inspection Manual Chapter 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," dated September 7, 2012, because, if left uncorrected, the performance deficiency would have the potential to lead a more significant safety concern and is therefore a finding. Failure to update the Final Safety Analysis Report with the basis for steam dryer structural integrity could result in a failure to maintain the structural integrity of the steam dryer. Specifically, insuffient steam dryer inspections could result in failure to detect structurally significant cracking and result in a steam dryer failure which generates debris that adversely affect the function of safety-related compoments (e.g. MSIVs). Additionally, the failure to update the Final Safety Analysis Report with the basis for steam dryer structural integrity was more than minor because it was associated with the Initiating Events Cornerstone attribute of Equipment Performance and adversely affected the Cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions.
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| Violations of 10 CFR 50.71(e) are dispositioned using the traditional enforcement process because they are considered to be violations that potentially impede or impact the regulatory process. This violation was also associated with a finding that has been evaluated by the significance determination process (SDP) and communicated with SDP color reflective of the safety impact of the deficient licensee performance. The SDP, however, does not specifically consider regulatory process impact. Thus, although related to a common regulatory concern, it is necessary to address the violation and finding using different processes to correctly reflect both the regulatory importance of the violation and the safety significance of the associated finding.
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| Using Manual Chapter 0609, Attachment 4 "Initial characterization of Findings," and Appendix A "The Significance Determination Process for findings at Power" the finding was screened against the initiating events cornerstone and determined to be of very low safety significance (Green) because the finding did not cause a reactor trip and the loss of mitigating equipment relied upon to transition from the onset of the trip to a stable. The performance deficiency associated with this finding did not reflect current licensee performance; therefore, no cross cutting aspect was identified with this finding.
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| Additionally, in accordance with Section 6.1.d.3 of the NRC Enforcement Policy, this violation was categorized as Severity Level IV because the licensee's failure to update the FSAR as required by 10 CFR 50.71(e) had not yet resulted in any unacceptable change to the facility or procedures.
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| Inspection Report# : 2014004 (pdf)
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| Significance: Sep 30, 2014 Identified By: NRC Item Type: NCV Non-Cited Violation MODIFICATION TO STEAM DRYER TIE BARS 28 AND 30 WITHOUT A 10 CFR 50.59 SAFETY EVALUATION The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.59(d)(1), "Changes, Test, and Experiments" for the licensee's failure to perform a written evaluation, which provided the bases for the determination that a change did not require a license amendment. Specifically, the licensee made a change pursuant to 10 CFR 50.59 (c) with the installation of 1/2 inch holes adjacent to welds attaching tie bars 28 and 30 to the steam dryer vane assembly and did not provide a written evaluation to provide a basis for the determination that this change would not result in a more than minimal increase in the likelihood of occurrence of a malfunction of an system structure or component important to safety (e.g. MSIVs). The licensee entered this finding into the corrective action program as issue report IR 02223135 and identified an action to secure a detailed assessment of these degraded tie bar locations Page 2 of 12
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| 4Q/2014 Inspection Findings - Clinton from the steam dryer vendor. The licensee also consulted with the steam dryer vendor and made a qualitative assessment that the additional unflawed and unaltered portion of the fillet welds present at the end of the tie bar 28 and 30 locations provided a reasonable basis to conclude that these tie bars would not fail and affect the operability of safety-related components.
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| The inspectors determined that the failure to provide a written evaluation, which provided the basis for the determination that a change did not require a license amendment, was a performance deficiency. Specifically, the licensee failed to provide a basis for not applying for a license amendment associated with increased likelihood of a SD failure that impacts safety-related equipment due to reduced structural support available at tie bars 28 and 30. The performance deficiency was determined to be more than minor in accordance with IMC 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," dated September 7, 2012, because it was associated with the Initiating Events cornerstone attribute of equipment performance and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. In addition, the associated violation was determined to be more than minor because the inspectors could not reasonable determine if the changes to the SD at tie bars 28 and 30 would have required NRC prior approval.
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| Violations of 10 CFR 50.59 are dispositioned using the traditional enforcement process instead of the SDP because they are considered to be violations that potentially impede or impact the regulatory process. However, if possible, the underlying technical issue is evaluated under the SDP to determine the severity of the violation. In this case, the inspectors used Manual Chapter 0609, Attachment 4 "Initial characterization of Findings," and Appendix A "The Significance Determination Process for findings a Power" the finding was screened against the initiating events cornerstone and determined to be of very low safety significance (Green) because the finding did not cause a reactor trip and the loss of mitigating equipment relied upon to transition from the onset of the trip to a stable. The performance deficiency associated with this finding did not reflect current licensee performance; therefore, no cross cutting aspect was identified with this finding.
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| In accordance with Section 6.1.d.2 of the NRC Enforcement Policy, this violation was categorized as Severity Level IV because the resulting changes were evaluated by the SDP as having very low safety significance.
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| Inspection Report# : 2014004 (pdf)
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| Significance: Jul 11, 2014 Identified By: NRC Item Type: FIN Finding FAILURE TO IDENTIFY A LEVEL 1 TEST CRITERION FAILURE The inspectors documented a self-revealing Green finding associated with the failure to follow procedures when performing power ascension testing on the digital feedwater (DFW) system. Specifically, contrary to procedure CPS 2894.01, "Digital FWLC [feedwater level control system] Modifications Test - Power Ascension," Section 9.1, the licensee did not declare a Level 1 criterion failure when unacceptable oscillations were noted during a transition in the power ascension test. This resulted in the licensee declaring the test successful and returning the system to service without taking the appropriate corrective actions to address the oscillations. This contributed to the subsequent scram caused by reactor water level oscillations.
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| The failure to follow procedures when performing power ascension testing on the digital feedwater system was a performance deficiency. The performance deficiency was more than minor because it was associated with the design control attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations and is therefore a finding. Using IMC 0609, Attachment 4 "Initial Characterization of Findings," and Appendix A "The Significance Determination Process for Findings at Power," issued June 19, 2012, the finding was determined to be of very low safety significance (Green) because it did not cause a reactor trip with a coincident loss Page 3 of 12
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| 4Q/2014 Inspection Findings - Clinton of mitigating equipment. The inspectors determined this finding affected the conservative bias aspect of the of human performance cross-cutting area described as being present when the organization uses decision making practices that emphasize prudent choices over those that are simply allowable. Specifically, the licensee used non-conservative assumptions when determining whether the condition identified during the power ascension test was allowable (H.14).
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| This finding does not involve enforcement action because no violation of regulatory requirements was identified.
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| Inspection Report# : 2014008 (pdf)
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| Significance: Jun 30, 2014 Identified By: NRC Item Type: FIN Finding ELECTRO HYDRAULIC CONTROL SYSTEM LEAK RESULTS IN MANUAL SCRAM The inspectors documented a self-revealing, Green finding associated with a failure to provide adequate work instructions to perform repairs to the shutoff valve for 1TGCV4 main turbine control valve. Specifically, contrary to station procedure MA-AA-716-010, Maintenance Planning, Revision 21, the work instructions generated to install the shutoff valve failed to reference the appropriate cap screw size, lubricate the cap screws and install lock washers on the cap screws used to attach the shut off valve to the control valve. This allowed the cap screws to loosen and ultimately fail due to fatigue resulting in a leak of electro hydraulic control fluid of sufficient rate to require a manual scram of Unit 1 on April 26, 2013. The valve was replaced and successfully tested and the unit was restarted. The licensee documented this issue in the corrective action program (CAP) as Issue Report (IR) 01506929.
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| The performance deficiency was more than minor because it was associated with the procedure quality attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations and is therefore a finding. Using Manual Chapter 0609, Attachment 4 Initial Characterization of Findings, and Appendix A The Significance Determination Process for Findings at Power, issued June 19, 2012, the finding was screened against the initiating events cornerstone and determined to be of very low safety significance (Green) because the finding did not cause a reactor trip with a coincident loss of mitigating equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The inspectors determined that no cross cutting aspect will be assigned to this performance deficiency since it occurred in 2008 and is not indicative of current plant performance.
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| Inspection Report# : 2014003 (pdf)
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| Significance: Jun 30, 2014 Identified By: NRC Item Type: FIN Finding FAILURE TO IMPLEMENT ENGINEERING CHANGE RESULTS IN MANUAL REACTOR SCRAM The inspectors documented a self-revealing, Green finding associated with a failure to implement engineering change (EC) 380150 Upgrade Feed Water Level Control and Turbine Speed. Specifically, contrary to station procedure CC-AA-256, Process for Managing Plant Modifications Involving Microprocessor Technology, Revision 2, the licensee failed to identify, evaluate and mitigate software component critical parameters in the engineering change that installed the digital feed water system. This resulted in nonlinear reactor water level oscillations when transferring from the motor driven feed pump to the turbine driven feed pump that required the reactor operator to manually scram the reactor prior to reaching the level 8 automatic scram set point. The licensee documented this issue in the corrective action program as IR 1596987.
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| The performance deficiency was more than minor because it was associated with the design control attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations and is Page 4 of 12
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| 4Q/2014 Inspection Findings - Clinton therefore a finding. Using Manual Chapter 0609, Attachment 4 Initial Characterization of Findings, and Appendix A The Significance Determination Process for Findings at Power, issued June 19, 2012, the finding was screened against the initiating events cornerstone and determined to be of very low safety significance (Green) because the finding did not cause a reactor trip with a coincident loss of mitigating equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The inspectors determined this finding affected the cross cutting area of human performance in the aspect of documentation where the organization creates and maintains complete, accurate and up-to date documentation. Specifically, the contractors failed to create complete documentation to be use by the licensee when evaluating the critical parameters.
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| Inspection Report# : 2014003 (pdf)
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| Significance: Feb 14, 2014 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Correct Identified Combustibles The inspectors identified a finding of very low safety significance and associated NCV of License Condition 2.F for the failure to remove an identified combustible. Specifically, the failure to remove a piece of wood located directly under a safety-related cable tray for a period in excess of three years was a failure to take corrective action as required by the licensees Quality Assurance Program. The licensee entered the issue into their Corrective Action Program and removed the piece of wood by the end of the inspection.
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| The finding was determined to be more than minor because the combustible material was located directly beneath a safety-related cable tray and, as such, represented a credible fire scenario. The finding was determined to be of very low safety significance (i.e., Green) because the impact of the fire would be largely limited to one train/division of equipment important to safety. The inspectors determined that the finding has a cross-cutting aspect in the area of human performance because the licensee did not ensure sufficient resources were available to support nuclear safety.
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| Specifically, the failure to remove the identified combustible was due to a lack of resources to schedule and accomplish removing the material.
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| Inspection Report# : 2014007 (pdf)
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| Mitigating Systems Significance: Dec 31, 2014 Identified By: NRC Item Type: NCV Non-Cited Violation FAILURE TO TRANSLATE SEISMIC DESIGN REQUIREMENTS INTO APPLICABLE PROCEDURES The inspectors identified a green non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for the failure to adequately translate seismic requirements from a design calculation into applicable procedures. Specifically the licensee failed to incorporate the seismic requirements for the Division III 4.16 KV switchgear as described in calculation IP-Q-0391 Seismic Qualification of 480V ABB Unit Sub Switchgears, Div I & II Westinghouse Switchgears and Div III GE 4.16KV Switchgears, into procedure CPS 1014.11 6900/4160/480V Switchgear/Circuit Breaker Operability Program, resulting in the licensee incorrectly declaring Division III switchgear operable when in a seismically unanalyzed condition. The licensee entered this issue into their corrective action program as AR 2386676.
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| Page 5 of 12
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| 4Q/2014 Inspection Findings - Clinton The inspectors determined that the failure to adequately incorporate the seismic requirements of the design calculation into the applicable procedure was a performance deficiency. The finding is more than minor because it is associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 4, External Events Screening Questions, dated June 19, 2012, the inspectors answered Yes to question 1 of External Events screening questions, because the finding could potentially degrade one train of the emergency power system. Thus the inspectors consulted the regional senior reactor analyst (SRA).
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| Based on the Detailed Risk Evaluation, the inspectors determined that the finding was of very low safety significance (Green). The inspectors determined that there was no cross-cutting aspect associated with this finding because the cause of the performance deficiency occurred more than fifteen years ago, and was not representative of current licensee performance.
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| Inspection Report# : 2014005 (pdf)
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| Significance: Dec 31, 2014 Identified By: Self-Revealing Item Type: NCV Non-Cited Violation FAILURE TO PROVIDE PROCEDURE INSTRUCTION RESULTS IN EXCEEDING TECHNICAL SPECIFICATION HEAT UP RATE DURING PLANT START UP The inspectors are documenting a self-revealing non-cited violation of Technical Specification 5.4., Procedures, for the licensees failure to establish instructions in station procedure CPS 9059.01, Reactor Coolant System Leakage Test, Revision 9b. Specifically, the licensee failed to provide instructions to ensure that the main steam piping between the reactor vessel and the inboard main steam isolation valves were completely drained of water at the completion of testing. The licensee entered this issue into the corrective action program as action request AR 01590671.
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| The inspectors determined that the licensees failure to establish instructions to ensure that the main steam piping between the reactor vessel and the inboard main steam isolation valves were completely drained of water prior to starting up the reactor was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences, and is therefore a finding.
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| Using Manual Chapter 0609, Attachment 4 Initial Characterization of Findings, and Appendix A The Significance Determination Process for Findings at Power the finding was screened against the mitigating systems cornerstone and determined to be of very low safety significance (Green) because the finding was/did not: 1) a deficiency affecting the design or qualification of a mitigating structure, system or component, 2) represent a loss of system and/or function, 3) represent an actual loss of function of a single train for greater than its technical specification allowed outage time, 4) represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant for greater than 24 hours and 5) did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding or severe weather event.
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| The inspectors determined this finding affected the cross cutting area of human performance in the aspect of work management where the organization implements a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. Specifically, the licensee failed to have a plan or provide a control method to ensure the main steam piping was drained prior to commencing reactor start up. (IMC 0301 H.5)
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| Page 6 of 12
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| 4Q/2014 Inspection Findings - Clinton Inspection Report# : 2014005 (pdf)
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| Significance: Sep 30, 2014 Identified By: NRC Item Type: NCV Non-Cited Violation EXCEEDED TECHNICAL SPECIFICATION ALLOWED OUTAGE TIME FOR ELECTRICAL POWER SYSTEMS DUE TO AUXILIARY EQUIPMENT OUT OF SERVICE The inspectors identified a non-cited violation of Technical Specification 3.8.4, "DC Sources - Operating" and Technical Specification 3.8.9, "Distribution Systems - Operating" for failing to enter the technical specifications and complete the associated actions prior to the completion time when auxiliary equipment required to support electrical power system safety function was out of service. Specifically, the licensee removed the division 1 safety related portion of the switchgear cooling system from service to perform maintenance and failed to enter the applicable technical specifications that the was required to support system safety function. The licensee documented this issue in the corrective action program as Issue Report (IR) 01674754.
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| The failure to enter the technical specifications and complete the associated actions prior to the completion time when auxiliary equipment required to support electrical power system safety function was out of service was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences and is therefore a finding. Using Manual Chapter 0609, Appendix A, "The Siginificance Determination Process (SDP) for Findings At-Power," issued June 19, 2012, Exhibit 2 for the Mitigating Systems Cornerstone. The inspectors answered "Yes" to the screening question under the Mitigating Systems Cornerstone "Does the finding represent and actual loss of function of at least a single train for > its Tech Spec Allowed Outage Time OR two separate safety systems out-of-service for > its Tech Spec Allowed Outage Time?,' since the finding represented an actual loss of function of at least a single Train for > its Tech Spec Allowed Outage time. Therefore, a detailed risk evaluation was performed using IMC 0609, Appendix A. The Senior Reactor Analysts (SRAs) evaluated the finding using the Clinton Standardized Plant Analysis Risk (SPAR) model version 8.17, Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) version 8.1.0. For switchgear cooling, independent redundant cooling trains are provided for each of the three divisional switchgear areas with one train being non-safety related and the other safety related. In order to characterize the risk significance, the SRAs assumed that during a loss of offsite power (LOOP) event, the non-safety related switchgear cooling train that is normally in operation would become unavailable. The safety-related cooling train, should it be undergoing maintenance, would be unavailable as well. The exposure time for this issue was taken to be 235 hours based on the licensee documentation. Post-processing rules were used to credit an additional 4.0 hours of time to recover offsite power (to allow recovery of the non-safety cooling train) in core damage sequences when the safety-related cooling train for Division 1 equipment was undergoing maintenance during a LOOP. The SRAs also gave credit in the SPAR Model for local operator action to provide alternate switchgear room cooling during a LOOP. The licensee produced Alarm Response Procedure CPS 5050.03, Rev 30c, which directed operators to Procedure CPS 3412.01, "Essential Switchgear Heat Removal (VX)...,"
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| Revision 15. These procedures directed operators to locally open doors, set up protable blowers, or lower electrical loads to help cool the room as necessary. The SRAs used the SPAR-H Human Reliability Analysis Method (NUREG/CR-6883) to estimate the human error probability for identifying and executing the local actions. the performance drivers were "time" (extra time) and "stress" (high) for diagnosis. The performance drivers were "stress" (high) and "ergonomics" (poor) for action. The resultant human error probability using these assumptions was 0.022. Using the above information, the ?CDF during the exposure time is 1.7E-08/yr. The dominant sequences were station blackout sequences, with initial success of RCIC and HPCS, but later failure of those systems and decay heat removal and all injection due to failure to vent containment and its subsequent failure. Based on the detailed risk evaluation, this finding is best characterized as a finding of very low safety-significance (Green.) The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of avoid complacency where individuals recognize and plan for mistakes, latent issues, and inherent risk, even while expecting successful Page 7 of 12
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| 4Q/2014 Inspection Findings - Clinton outcomes. Specifically, the licensee has removed the division 1 or 2 safety related switchgear cooling system fans or condensing units from service numerous times and failed to consider the components inoperable under technical specification definition for operable. (IMC 0310 H.12)
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| Inspection Report# : 2014004 (pdf)
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| Significance: Sep 30, 2014 Identified By: NRC Item Type: NCV Non-Cited Violation PROGRAMMATIC FAILURE TO COMPLETE OPERABILITY AND FUNCTIONALITY DETERMINATIONS The inspectors identified a non-cited violation of 10 CFR 50 Appendix B, Criterion V, Instructions, Procedures and Drawings, "Procedures," for the failure to accomplish station procedure OP-AA-108-115, "Operability Determinations" Revision 14. Specifically, on multiple occasions operations personnel failed to complete or documented incomplete operability or functionality of safety related or related to safety equipment used at the site.
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| The licensee documented this issue in the corrective action program as Issue Report (IR) 01693256.
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| The failure to complete or provided incomplete operability or functionality determinations used to determine the operability or functionality of safety related or related to safety equipment used at the site is a performance deficiency.
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| The performance deficiency was determined to be more than minor because if left uncorrected, the performance deficiency has the potential to lead to a more significant safety concern and is therefore a finding. Specifically, if operations personnel continue to fail to complete or provide incomplete operability or functionality determination the station could have safety or safety related equipment inoperable without taking appropriate actions for the equipment being inoperable (e.g. entering appropriate technical specification limited condition for operation). Using Manual Chapter 0609, Attachment 4 "Initial Characterization of Findings," and Appendix A "The Significance Determination Process for Findings at Power" the finding was screened against the mitigatins systems cornerstone and determined to be of very low safety significance (Green) because the finding was/did not: 1) a deficiency affecting the design or qualification of a mitigating structure, system or component, 2) represent a loss of system and/or function, 3) represent an actual loss of function of a single train for greater than its technical specification allowed outage time, 4) represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant for greater than 24 hours and 5) did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding or serve weather event. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of Training, where the organization provides training and ensures knowledge transfer to maintain a knowledgeable, technically competent workforce and instill nuclear safety values. Specifically, personnel performing the reviews believed existing training provided sufficient knowledge without the use of additional resources material and current training to operators does not covet this activity. (IMC 0319 H.9)
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| Inspection Report# : 2014004 (pdf)
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| Significance: Sep 30, 2014 Identified By: NRC Item Type: NCV Non-Cited Violation FAILURE TO ESTABLISH SURVEILLANCE PRODEDURE FOR REACTOR CORE ISOLATION COOLING PUMP DUE TO UNACCEPTABLE PRECONDITIONING The inspectors determined that the failure to establish a surveillance procedure to test the RCIC system due to unacceptable preconditioning is a performance deficiency. The performance deficiency was more than minor because it was associated with the procedure quality attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability and capability to response to initiating events to prevent undesirable consequences and is therefore a finding. Using Manual Chapter 0609, Attachment 0609.04 "Initial Characterization of Findings," and Appendix A "The Significance Determination Process for Findings at Power" the Page 8 of 12
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| 4Q/2014 Inspection Findings - Clinton finding was screened against the mitigating systems cornerstone and determined to be of very low safety significance (Green) because the finding was/did not: 1) a deficiency affecting the design or qualification of a mitigating structure, system or component, 2) represent a loss of system and/or function, 3) represent an actual loss of function of a single train for greater than its technical specification allowed outage time, 4) represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant for greater than 24 hours and 5) did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding or severe weather event. The inspectors determined this finding affected the cross-cutting area of problem identification and resolution in the aspect of operating experience where the organization systematically and effectively collects, evaluates and implements relevant internal and external operating experience in a timely manner.
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| Specifically, the licensee considered the impact of the operating experience for surveillance testing, but did not consider its impact during normal plant operation. (IMC 0310 P.5)
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| Inspection Report# : 2014004 (pdf)
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| Significance: Jun 30, 2014 Identified By: NRC Item Type: NCV Non-Cited Violation FOREIGN MATERIAL IN RELAY PREVENTS EMERGENCY DIESEL GENERATOR OUTPUT BREAKER FROM CLOSING The inspectors documented a self-revealing, Green non-cited violation of Clinton Power Station Technical Specification 5.4.1, Procedures, for a failure to prevent foreign material from entering a relay associated with the Division 1 Diesel Generator. Specifically, contrary to station procedure CPS 8501.05, CV-2 Relay Inspection and Calibration with Doble Test Equipment, Revision 4, the licensee failed to verify that relay 227-DGIKA, CV-2 AB phase was clean and free of all foreign material. The foreign material prevented the relay from operating and satisfying the permissive logic required to close the Division 1 Diesel Generator output breaker resulting in having to declare the Diesel Generator inoperable. The relay was replaced and successfully tested and the licensee documented this issue in the corrective action program as IR 01600935.
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| The finding was more than minor because it was associated with the procedure quality attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences and is therefore a finding.
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| Using Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012, Exhibit 2 for the Mitigating Systems Cornerstone, the inspectors answered Yes to the screening question under the Mitigating Systems Cornerstone Does the finding represent an actual loss of function of at least a single Train for > its Tech Spec Allowed Outage Time OR two separate safety systems out-of- service for >
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| its Tech Spec Allowed Outage Time?," since the finding represented an actual loss of function of at least a single Train for > its Tech Spec Allowed Outage Time of 14 days. Therefore, a detailed risk evaluation was performed using IMC 0609, Appendix A. The Senior Reactor Analysts (SRAs) evaluated the finding using the Clinton Standardized Plant Analysis Risk (SPAR) model version 8.17, Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) version 8.1.0 and concluded that the risk increase to the plant due to this finding is very low (Green). The inspectors determined this finding affected the cross cutting area of human performance in the aspect of work management where the organization implements a process of planning, controlling and executing work activities such that nuclear safety is the overriding priority. Specifically, the licensees implementation of their foreign material exclusion process for this maintenance activity lacked sufficient planning, controls and execution to prevent foreign material from entering a risk significant piece of safety related equipment.
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| Inspection Report# : 2014003 (pdf)
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| Significance: Jun 30, 2014 Identified By: NRC Page 9 of 12
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| 4Q/2014 Inspection Findings - Clinton Item Type: NCV Non-Cited Violation FAILURE TO DEVELOP ADEQUATE PROCEDURES FOR PRE-PLANNING AND PERFORMING MAINTENANCE AFFECTING SAFETY-RELATED EQUIPMENT The inspectors documented a self-revealing, Green non-cited violation (NCV) of Clinton Power Station Technical Specification 5.4.1, Procedures for a failure to develop adequate procedures for properly pre-planning and performing maintenance affecting the performance of safety-related equipment which resulted in the subsequent failure of the Division 3 Diesel Room Ventilation damper hydramotor on August 15, 2013. Specifically, during pre-scheduled performance testing of the Division 3 (High Pressure Core Spray System) Emergency Diesel Generator Room Ventilation Damper hydramotor, the ventilation supply air intake damper, 1VD01YC, failed to open as a result of Damper Hydramotor 1TZVD003A experiencing an age-related degradation failure. This was due in part to the licensees failure to properly pre-plan and perform the appropriate preventive maintenance for the hydramotor due to inadequate procedures. Procedure WC-AA-113, Predefine Database Revisions, Revision 2, did not provide adequate instructions appropriate to the circumstances to properly pre-plan and perform maintenance affecting the performance of safety-related equipment. This resulted in a loss of safety function of the HPCS Diesel Generator and its supported High Pressure Core Spray system because of the low confidence that diesel room temperature would be maintained to support the diesel during an event when it would be required to perform its function. The licensee subsequently replaced the hydramotor, tested the new hydramotor successfully and restored the diesel ventilation system to operable. They documented this issue in the corrective action program as IR 1546973 and IR 1547294.
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| The performance deficiency was more than minor because it was associated with the equipment performance attribute of the mitigating systems cornerstone attribute and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences and is therefore a finding. Using Manual Chapter 0609, Appendix A, The SDP for Findings At-Power, issued June 19, 2012, Exhibit 2 for the Mitigating Systems Cornerstone. The inspectors answered Yes to the screening question under the Mitigating Screening Cornerstone Does the finding represent a loss of system and/or function? since the finding resulted in a loss of safety function. Therefore, a detailed risk evaluation was performed using IMC 0609, Appendix A. The SRAs evaluated the finding using the Clinton SPAR model version 8.17, SAPHIRE version 8.1.0 and concluded that the risk increase to the plant due to this finding is very low (Green). The inspectors determined that no cross-cutting aspect will be assigned to this performance deficiency since it occurred in 2005 and is not indicative of current plant performance Inspection Report# : 2014003 (pdf)
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| Barrier Integrity Significance: Dec 31, 2014 Identified By: NRC Item Type: NCV Non-Cited Violation FAILURE TO UPDATE THE UPDATED SAFETY ANALYSIS REPORT - 1VR08C FUNCTION The inspectors identified a Severity Level IV non-cited violation of Title 10 Code of Federal Regulations (CFR) 50.71 (e), Periodic Update of the USAR and an associated Green finding for the licensees failure to update the USAR with the correct description of the function of 1VR08C. Specifically the licensee did not update Section 9.4.5.5 of the USAR to include the correct function of 1VR08C as described in a commitment made to the NRC in letter U-600850.
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| Consequently the licensee performed a 50.59 evaluation for abandoning a portion of the system that did not consider the correct function of the component. The licensee entered this issue into their corrective action program as AR 1692665.
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| Page 10 of 12
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| 4Q/2014 Inspection Findings - Clinton The inspectors determined that the failure to update the USAR with the correct function of 1VR08C was a performance deficiency. The performance deficiency was determined to be more than minor in accordance with Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, because, if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern and is therefore a finding. Specifically, failure to update the USAR with the correct safety related function of VR08C could result in the licensee making operability and functionality determinations based on incorrect assumptions. Additionally, the failure to update the USAR with the correct function of the fan was more than minor because it was associated with the Barrier Integrity Cornerstone attribute of design control, plant modifications and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events.
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| Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings at Power, the finding was screened against the Barrier Integrity cornerstone and determined to be of very low safety significance (Green) because the finding does not represent an actual open pathway in the physical integrity of reactor containment, containment isolation system, and heat removal components and it did not involve an actual reduction in function of hydrogen igniters in reactor containment. The performance deficiency associated with this finding did not reflect current licensee performance; therefore, no cross cutting aspect was identified with this finding.
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| Additionally, in accordance with Section 6.1.d.3 of the NRC Enforcement Policy, this violation was categorized as Severity Level IV because the licensees failure to update the USAR as required by 10 CFR 50.71(e) had not yet resulted in any unacceptable change to the facility or procedures.
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| Inspection Report# : 2014005 (pdf)
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| Emergency Preparedness Significance: Sep 30, 2014 Identified By: NRC Item Type: NCV Non-Cited Violation INCOMPLETE EVACUATION TIME ESTIMATE SUBMITTALS The inspectors determined that Exelon's failure to submit a complete updated ETE for the Clinton Power Station by December 22, 2012 was a performance deficiency. Specifically, the ETE is an input into the development of protective action strategies prior to an accident and to the protective action recommendation decision making process during an accident. Inadequate ETEs have the potential to reduce the effectiveness of public protective actions implemented by the OROs. The performance deficiency was more than minor because it was associated with the procedure quality attribute of the emergency preparedness cornerstone and adversely affected the cornerstone objective to ensure that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency and is therefore a finding. Using IMC 0609, attachment 0609.04 "Initial Characterization of Findings," and Appendix B, "Emergency Preparedenss (EP) Significance Determination Process (SDP)," the finding was screened by the inspectors and determined to be of very low safety significance (Green) based upon the following. The performance deficiency was associated with planning standard 10 CFR 50.47 (b)(10)," Green Finding column, provides the following exasmples "ETEs and updates to the ETEs were not provided to responsible OROs," and "The current public protective action strategies documented in emergency preparedness implementing procedures (EPIPs) are not consistent with the current ETE." The inspectors concluded that the incomplete updated ETE delayed the NRC's approval of the Clinton Power Station ETE, therefore the ETE was not provided to the site OROs nor was it used to inform the site EPIPs as required by 10 CFR 50.47(b)(10), and Section Page 11 of 12
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| 4Q/2014 Inspection Findings - Clinton IV, Paragraph 4 of Appendix E to 10 CFR Part 50. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of documentation where the organization creates and maintains complete, accurate and up-to-date documentation. Specifically, the Emergency Preparedness organization did not develop the Clinton Power Station ETE as required by the new regulation introduced by the NRC's EP Rule. (IMC 0310 H.7)
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| Inspection Report# : 2014004 (pdf)
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| Occupational Radiation Safety Public Radiation Safety Security Although the Security Cornerstone is included in the Reactor Oversight Process assessment program, the Commission has decided that specific information related to findings and performance indicators pertaining to the Security Cornerstone will not be publicly available to ensure that security information is not provided to a possible adversary.
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| Other than the fact that a finding or performance indicator is Green or Greater-Than-Green, security related information will not be displayed on the public web page. Therefore, the cover letters to security inspection reports may be viewed.
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| Miscellaneous Last modified : April 01, 2015 Page 12 of 12
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| 1Q/2015 Inspection Findings - Clinton Clinton 1Q/2015 Plant Inspection Findings Initiating Events Significance: Dec 31, 2014 Identified By: NRC Item Type: NCV Non-Cited Violation STATION PROCEDURES FAILED TO PROVIDE CONTROLS FOR MATERIAL NEAR TRANSFORMERS The inspectors identified a non-cited violation associated with a failure to provide controls for material near the station transformers. Specifically, station procedure CPS 4302.01, "Tornado/High Winds", Revision 21b does not include guidelines or examples of the types of materials to control as potential missiles in high velicity winds or tornadoes, and does not include triggers to perform walkdowns when high winds are predicted, prior to off-normal entry, to control material adjacent to the offsite power transformers that could result in the loss of offsite power. The licensee entered this issue into the corrective action program as action request (AR) 2388608.
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| The failure to provide guidelines or examples of the types of materials to control as potential missiles in high velicity winds or tornadoes and provide triggers to perform walkdowns when high winds are predicted was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations and is therefore a finding. Using Manual Chapter 0609, Attachment 4 "Initial Characterization of Findings," and Appendix A "The Significance Determination Process for Findings at Power", issued June 19, 2012, the finding was screened against the initiating events cornerstone and determined to be of very low safety significance (Green) because the finding did not involve the complete or partial loss of a support system that contributes to the likelihood of, or caused, an initiating event and did not affected mitigation equipment.
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| The inspectors determined this finding affected the cross-cutting area of problem identification and resolution in the aspect of operating experience where the organization systematically and effectively collects, evaluates, and implements relevant internal and external operating experience in a timely manner. Specifically, the licensee opearting experience program failed to ensure evaluation and implementation of interal operating experience in a timely manner after previous identification in the corrective action progrma. (IMC 0310 P.5)
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| Inspection Report# : 2014005 (pdf)
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| Significance: Sep 30, 2014 Identified By: NRC Item Type: NCV Non-Cited Violation FAILURE TO UPDATE THE FINAL SAFETY ANALYSIS REPORT (FSAR) - SD STRUCTURAL INTEGRITY FUNCTION The inspectors identified a Severity Level IV Non-Cited Violation of title 10 Code of Federal Regulations (CFR) 50.71(e), 'Periodic Update of the Final Safety Analysis Report' and an associated Green finding for the licensee's failure to update the Final Safety Analysis Report with a description of the basis for the steam dryer structural integrity submitted to the NRC in support of an extended power uprate license amendment. Specifically, the licensee did not update Section 3.9.5.1.1.9. "Steam Dryers," of the FSAR to include analysis and inspections of the steam dryer Page 1 of 14
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| 1Q/2015 Inspection Findings - Clinton each refueling outage that provided the basis for steam dryer structural integrity. Consequently, the licensee had not completed an inspection of the steam dryer during the most recent refueling outage. The licensee entered this issue into the corrective action program as issue report IR 02223135 and initiated actions to evaluate the Final Safety Analysis Report for revision to include description of the structural integrity function of the steam dryer.
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| The inspectors determined that the licensee's failure to update the Final Safety Analysis Report with the basis for steam dryer structural integrity submitted to the NRC was a performance deficiency. the performance deficiency was determined to be more than minor in accordance with Inspection Manual Chapter 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," dated September 7, 2012, because, if left uncorrected, the performance deficiency would have the potential to lead a more significant safety concern and is therefore a finding. Failure to update the Final Safety Analysis Report with the basis for steam dryer structural integrity could result in a failure to maintain the structural integrity of the steam dryer. Specifically, insuffient steam dryer inspections could result in failure to detect structurally significant cracking and result in a steam dryer failure which generates debris that adversely affect the function of safety-related compoments (e.g. MSIVs). Additionally, the failure to update the Final Safety Analysis Report with the basis for steam dryer structural integrity was more than minor because it was associated with the Initiating Events Cornerstone attribute of Equipment Performance and adversely affected the Cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions.
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| Violations of 10 CFR 50.71(e) are dispositioned using the traditional enforcement process because they are considered to be violations that potentially impede or impact the regulatory process. This violation was also associated with a finding that has been evaluated by the significance determination process (SDP) and communicated with SDP color reflective of the safety impact of the deficient licensee performance. The SDP, however, does not specifically consider regulatory process impact. Thus, although related to a common regulatory concern, it is necessary to address the violation and finding using different processes to correctly reflect both the regulatory importance of the violation and the safety significance of the associated finding.
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| Using Manual Chapter 0609, Attachment 4 "Initial characterization of Findings," and Appendix A "The Significance Determination Process for findings at Power" the finding was screened against the initiating events cornerstone and determined to be of very low safety significance (Green) because the finding did not cause a reactor trip and the loss of mitigating equipment relied upon to transition from the onset of the trip to a stable. The performance deficiency associated with this finding did not reflect current licensee performance; therefore, no cross cutting aspect was identified with this finding.
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| Additionally, in accordance with Section 6.1.d.3 of the NRC Enforcement Policy, this violation was categorized as Severity Level IV because the licensee's failure to update the FSAR as required by 10 CFR 50.71(e) had not yet resulted in any unacceptable change to the facility or procedures.
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| Inspection Report# : 2014004 (pdf)
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| Significance: Sep 30, 2014 Identified By: NRC Item Type: NCV Non-Cited Violation MODIFICATION TO STEAM DRYER TIE BARS 28 AND 30 WITHOUT A 10 CFR 50.59 SAFETY EVALUATION The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.59(d)(1), "Changes, Test, and Experiments" for the licensee's failure to perform a written evaluation, which provided the bases for the determination that a change did not require a license amendment. Specifically, the licensee made a change pursuant to 10 CFR 50.59 (c) with the installation of 1/2 inch holes adjacent to welds attaching tie bars 28 and 30 to the steam dryer vane assembly and did not provide a written evaluation to provide a basis for the determination that this change would not result in a more than minimal increase in the likelihood of occurrence of a malfunction of an system structure or component important to safety (e.g. MSIVs). The licensee entered this finding into the corrective action program as issue report IR 02223135 and identified an action to secure a detailed assessment of these degraded tie bar locations Page 2 of 14
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| 1Q/2015 Inspection Findings - Clinton from the steam dryer vendor. The licensee also consulted with the steam dryer vendor and made a qualitative assessment that the additional unflawed and unaltered portion of the fillet welds present at the end of the tie bar 28 and 30 locations provided a reasonable basis to conclude that these tie bars would not fail and affect the operability of safety-related components.
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| The inspectors determined that the failure to provide a written evaluation, which provided the basis for the determination that a change did not require a license amendment, was a performance deficiency. Specifically, the licensee failed to provide a basis for not applying for a license amendment associated with increased likelihood of a SD failure that impacts safety-related equipment due to reduced structural support available at tie bars 28 and 30. The performance deficiency was determined to be more than minor in accordance with IMC 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," dated September 7, 2012, because it was associated with the Initiating Events cornerstone attribute of equipment performance and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. In addition, the associated violation was determined to be more than minor because the inspectors could not reasonable determine if the changes to the SD at tie bars 28 and 30 would have required NRC prior approval.
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| Violations of 10 CFR 50.59 are dispositioned using the traditional enforcement process instead of the SDP because they are considered to be violations that potentially impede or impact the regulatory process. However, if possible, the underlying technical issue is evaluated under the SDP to determine the severity of the violation. In this case, the inspectors used Manual Chapter 0609, Attachment 4 "Initial characterization of Findings," and Appendix A "The Significance Determination Process for findings a Power" the finding was screened against the initiating events cornerstone and determined to be of very low safety significance (Green) because the finding did not cause a reactor trip and the loss of mitigating equipment relied upon to transition from the onset of the trip to a stable. The performance deficiency associated with this finding did not reflect current licensee performance; therefore, no cross cutting aspect was identified with this finding.
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| In accordance with Section 6.1.d.2 of the NRC Enforcement Policy, this violation was categorized as Severity Level IV because the resulting changes were evaluated by the SDP as having very low safety significance.
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| Inspection Report# : 2014004 (pdf)
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| Significance: Jul 11, 2014 Identified By: NRC Item Type: FIN Finding FAILURE TO IDENTIFY A LEVEL 1 TEST CRITERION FAILURE The inspectors documented a self-revealing Green finding associated with the failure to follow procedures when performing power ascension testing on the digital feedwater (DFW) system. Specifically, contrary to procedure CPS 2894.01, "Digital FWLC [feedwater level control system] Modifications Test - Power Ascension," Section 9.1, the licensee did not declare a Level 1 criterion failure when unacceptable oscillations were noted during a transition in the power ascension test. This resulted in the licensee declaring the test successful and returning the system to service without taking the appropriate corrective actions to address the oscillations. This contributed to the subsequent scram caused by reactor water level oscillations.
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| The failure to follow procedures when performing power ascension testing on the digital feedwater system was a performance deficiency. The performance deficiency was more than minor because it was associated with the design control attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations and is therefore a finding. Using IMC 0609, Attachment 4 "Initial Characterization of Findings," and Appendix A "The Significance Determination Process for Findings at Power," issued June 19, 2012, the finding was determined to be of very low safety significance (Green) because it did not cause a reactor trip with a coincident loss Page 3 of 14
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| 1Q/2015 Inspection Findings - Clinton of mitigating equipment. The inspectors determined this finding affected the conservative bias aspect of the of human performance cross-cutting area described as being present when the organization uses decision making practices that emphasize prudent choices over those that are simply allowable. Specifically, the licensee used non-conservative assumptions when determining whether the condition identified during the power ascension test was allowable (H.14).
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| This finding does not involve enforcement action because no violation of regulatory requirements was identified.
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| Inspection Report# : 2014008 (pdf)
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| Significance: Jun 30, 2014 Identified By: NRC Item Type: FIN Finding ELECTRO HYDRAULIC CONTROL SYSTEM LEAK RESULTS IN MANUAL SCRAM The inspectors documented a self-revealing, Green finding associated with a failure to provide adequate work instructions to perform repairs to the shutoff valve for 1TGCV4 main turbine control valve. Specifically, contrary to station procedure MA-AA-716-010, Maintenance Planning, Revision 21, the work instructions generated to install the shutoff valve failed to reference the appropriate cap screw size, lubricate the cap screws and install lock washers on the cap screws used to attach the shut off valve to the control valve. This allowed the cap screws to loosen and ultimately fail due to fatigue resulting in a leak of electro hydraulic control fluid of sufficient rate to require a manual scram of Unit 1 on April 26, 2013. The valve was replaced and successfully tested and the unit was restarted. The licensee documented this issue in the corrective action program (CAP) as Issue Report (IR) 01506929.
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| The performance deficiency was more than minor because it was associated with the procedure quality attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations and is therefore a finding. Using Manual Chapter 0609, Attachment 4 Initial Characterization of Findings, and Appendix A The Significance Determination Process for Findings at Power, issued June 19, 2012, the finding was screened against the initiating events cornerstone and determined to be of very low safety significance (Green) because the finding did not cause a reactor trip with a coincident loss of mitigating equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The inspectors determined that no cross cutting aspect will be assigned to this performance deficiency since it occurred in 2008 and is not indicative of current plant performance.
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| Inspection Report# : 2014003 (pdf)
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| Significance: Jun 30, 2014 Identified By: NRC Item Type: FIN Finding FAILURE TO IMPLEMENT ENGINEERING CHANGE RESULTS IN MANUAL REACTOR SCRAM The inspectors documented a self-revealing, Green finding associated with a failure to implement engineering change (EC) 380150 Upgrade Feed Water Level Control and Turbine Speed. Specifically, contrary to station procedure CC-AA-256, Process for Managing Plant Modifications Involving Microprocessor Technology, Revision 2, the licensee failed to identify, evaluate and mitigate software component critical parameters in the engineering change that installed the digital feed water system. This resulted in nonlinear reactor water level oscillations when transferring from the motor driven feed pump to the turbine driven feed pump that required the reactor operator to manually scram the reactor prior to reaching the level 8 automatic scram set point. The licensee documented this issue in the corrective action program as IR 1596987.
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| The performance deficiency was more than minor because it was associated with the design control attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations and is Page 4 of 14
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| 1Q/2015 Inspection Findings - Clinton therefore a finding. Using Manual Chapter 0609, Attachment 4 Initial Characterization of Findings, and Appendix A The Significance Determination Process for Findings at Power, issued June 19, 2012, the finding was screened against the initiating events cornerstone and determined to be of very low safety significance (Green) because the finding did not cause a reactor trip with a coincident loss of mitigating equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The inspectors determined this finding affected the cross cutting area of human performance in the aspect of documentation where the organization creates and maintains complete, accurate and up-to date documentation. Specifically, the contractors failed to create complete documentation to be use by the licensee when evaluating the critical parameters.
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| Inspection Report# : 2014003 (pdf)
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| Mitigating Systems Significance: Mar 31, 2015 Identified By: NRC Item Type: FIN Finding FAILURE TO PERFORM ADEQUATE CHANNEL CALIBRATION ON SEISMIC INSTRUMENTATION The inspectors identified a Green Finding associated with the licensee's failure to perform an adequate channel calibration to determine the functionality of the stations seismic monitoring equipmen used for evaluaing earthquakes.
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| Specifically, station procedure CPS 9437.21, "Trix Time-History Accelerometer Channel Calibration," Revision 39c, did not include steps to ensure that battery backup power was provided to operate the equipment on a loss of the normal power source as part of the operability requirements. The licensee documented the issue in the corrective action program as action request (AR) 02454630. As a corrective action the licensee planned to correct procedure CPS 9437.21 to verify proper battery operation.
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| The licensee's failure to perform an adequate channel calibration to determine the functionality of the stations seismic monitoring equipment used for evaluating earthquakes was a performance deficiency. Specifically, station procedures did not include steps to ensure that battery backup power was provided to operate the equipment on a loss of the normal power source. The performance deficiency was more than minor because it adversely impacted the protection against external factors attribute of the Mitigating Systems cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual chapter 0609, Attachment 4 "Initial Characterization of Findings," and Appendix A "The Significance Determination Process for Findings at Power," issued June 19, 2012, the inspectors answered "yes" to the Mitigating Systems cornerstone question, "Does the finding involve the ... degradation of equipment ... specificallty designed to mitigate a seismic ...
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| initiating event ..." Therefore, the inspectors addressed the questions in Exhibit 4, "External Event Screening Questions." The inspectors answered "no" to the two questions in Exhibit 4. Specifically, 1) if completely failed the seismic monitor would not cause an initiating event or degrade multi-trains or risk-significant systems; and 2) the finding does not involve the total loss of any safety function. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of conservative bias where individuals use decision making-practices that emphasize prudent choices over those that are simply allowable and a proposed action is determined to be a safe in order to proceed, rather than unsafe in order to stop. Specifically, the licensee documented the issue of the voltage being high out of specification and instead of performing additional corrective actions to determine if leaving the voltage out of specification was appropriate marked the step as not applicable and proceeded with the rest of the procedure. (Section 1R15)
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| Inspection Report# : 2015001 (pdf)
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| Page 5 of 14
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| 1Q/2015 Inspection Findings - Clinton Significance: Mar 31, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation UNQUALIFIED SAFETY-RELATED CABLES USED IN A SUBMERGED ENVIRONMENT The inspectors identified a finding and an associated non-citied violation of 10 CFR 50 Appendix B, Criterion III, "Design Control," for the failure to maintain safety-related cables for the SX system in an environment for which they were designed. Specifically, the licensee failed to maintain SX safety-related cables in an environment for which they were designed when the cables were allowed to be submerged in water inside cable vaults. The licensee documented this issue in their corrective action program (CAP) as action request (AR) 02474543. Corrective actions included draining the cable vaults so that the cables were no longer submerged.
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| The licensee's failure to maintain safety-related cables for the SX system in an environment for which they were designed was a performance deficiency. The performance deficiency was determined to be more than minor because the finding was associated with the Mitigating Systems Cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licnesee failed to maintain SX safety-related cables in an environment for which they were designed when the cables were allowed to be submerged in water inside cable vaults. Using IMC 0609, "Significance Determination Process," Attachment 0609.04, "Initial Characterization of Findings," issued on June 19, 2012. Specifically, the inspectors used IMC 0609 Appendix A "SDP for Findings At-Power," issued June 19, 2012, Exhibit 2, "Mitigating Systems Screening Questions" to screen the finding. The finding screened as of very low safety significance (Green) because the inspectors answered yes to the question "does the SSC maintain its operability or functionality." Specifically, the SX system submerged cables did not cause the SX system to be inoperable or nonfunctional. The inspectors determined this finding affected the cross-cutting area of problem identification and resolution in the aspect of resolution, where the organization takes effective corrective actions to address issues in a timely manner commensurate with their safety significance. Specifically, the licensee failed to implement effective corrective actions to address an adverse trend of water in cable vaults which led to (SX) safety-related cables being submerged in water.
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| Inspection Report# : 2015001 (pdf)
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| Significance: Mar 31, 2015 Identified By: NRC Item Type: AV Apparent Violation FAILURE OF THE DIVISION 3 SHUTDOWN SERVICE WATER PUMP DUE TO AN INADEQUATE BUSHING DESIGN A self-revealed finding, preliminarily determined to be of low to moderate safety significance (White) and an associated AV of 10 CFR 50 Appendix B, Criterion III, Design Control, was identified for the failure to verify the suitability of the replacement pump design for the Division 3 Shutdown Service Water system. Spceifically, the design of the suction bell bushing for the replacement pump was inadequate to pass sufficient cooling water flow to the pump internals without being affected by mud and silt from the lake water. This finding was self-revealed on September 16, 2014, during a surveillance test to ensure operability of the Division 3 shutdown cooling water pump after the pump failed to start due to a damaged bushing rendering the pump inoperable. This finding does not represent an immediate safety concern because the licensee replaced the pump in September of 2014 with a pump of similar design and provided adequate documentation that assures the pump will remain operable until a different design for the bushing that failed can be installed by June of 2016.
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| The inspectors determined that the licensee's failure to verify the suitability of the design for the Division 3 Shutdown Service Water pump was a performance deficiency warranting a significance evaluation. The inspectors determined that the finding was more than minor in accordance with IMC 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," dated September 7, 2012, because it was associated with the Mitigating Systems Cornerstone Page 6 of 14
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| 1Q/2015 Inspection Findings - Clinton attributes of design control and equipment performance and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. A Significance and Enforcement Review Panel (SERP), using IMC 0609, Appendix A, "Significance Determination Process For Findings At-Power," dated June 19, 2012, preliminarily determined the finding to be of low to moderate safety significance (White). The performance deficiency associated with this finding did not reflect current licensee performance; therefore, no cross cutting aspect was identified with this finding.
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| Inspection Report# : 2015001 (pdf)
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| Significance: Mar 20, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation Inadequate 50.59 Evaluation for Switchgear in Seismically Unanalyzed Conditions (Section 1R17.1b.)
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| Severity Level IV Green. The inspectors identified a finding of very-low safety significance, and an associated Non-Cited Violation of Title 10, Code of Federal Regulations Part 50, Section 59, Changes, Tests and Experiments, (effective January 1, 1997) for a procedure change dated May 2, 1997, where the licensee allowed safety-related switchgear to operate for a limited period of time during plant operation in equipment configurations that were seismically unanalyzed. Specifically, for Safety Evaluation Log 97 060, CPS [Clinton Power Station]
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| Procedure No. 1014.11, Revision 0, the licensee failed to include a written safety evaluation which provided the bases that concluded for all switchgear configurations that a seismically unanalyzed condition does not involve an unreviewed safety question, and the possibility for a malfunction of a different type than any evaluated previously in the Safety Analysis Report may be created. The licensee entered the issue into their Corrective Action Program as Action Request 02471583, NRC Mod 50.59 Inspection Safety Eval 97 060 for CPS 1014.11, dated March 20, 2015.
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| The performance deficiency was determined to be more than minor because the finding was associated with the Mitigating Systems cornerstone attribute of protection against external factors and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, switchgear in a seismically unanalyzed condition when relied upon to perform a safety function did not ensure the availability, reliability, or capability of the associated Mitigating Systems to respond to an initiating event such as an earthquake. The inspectors determined that the underlying technical issue was of very-low safety significance (Green) using a detailed risk evaluation. The inspectors did not identify a cross-cutting aspect associated with the finding because the finding was not representative of current performance.
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| Inspection Report# : 2015008 (pdf)
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| Significance: Dec 31, 2014 Identified By: NRC Item Type: NCV Non-Cited Violation FAILURE TO TRANSLATE SEISMIC DESIGN REQUIREMENTS INTO APPLICABLE PROCEDURES The inspectors identified a green non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for the failure to adequately translate seismic requirements from a design calculation into applicable procedures. Specifically the licensee failed to incorporate the seismic requirements for the Division III 4.16 KV switchgear as described in calculation IP-Q-0391 Seismic Qualification of 480V ABB Unit Sub Switchgears, Div I & II Westinghouse Switchgears and Div III GE 4.16KV Switchgears, into procedure CPS 1014.11 6900/4160/480V Switchgear/Circuit Breaker Operability Program, resulting in the licensee incorrectly declaring Division III switchgear operable when in a seismically unanalyzed condition. The licensee entered this issue into their corrective action program as AR 2386676.
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| The inspectors determined that the failure to adequately incorporate the seismic requirements of the design calculation Page 7 of 14
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| 1Q/2015 Inspection Findings - Clinton into the applicable procedure was a performance deficiency. The finding is more than minor because it is associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 4, External Events Screening Questions, dated June 19, 2012, the inspectors answered Yes to question 1 of External Events screening questions, because the finding could potentially degrade one train of the emergency power system. Thus the inspectors consulted the regional senior reactor analyst (SRA).
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| Based on the Detailed Risk Evaluation, the inspectors determined that the finding was of very low safety significance (Green). The inspectors determined that there was no cross-cutting aspect associated with this finding because the cause of the performance deficiency occurred more than fifteen years ago, and was not representative of current licensee performance.
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| Inspection Report# : 2014005 (pdf)
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| Significance: Dec 31, 2014 Identified By: Self-Revealing Item Type: NCV Non-Cited Violation FAILURE TO PROVIDE PROCEDURE INSTRUCTION RESULTS IN EXCEEDING TECHNICAL SPECIFICATION HEAT UP RATE DURING PLANT START UP The inspectors are documenting a self-revealing non-cited violation of Technical Specification 5.4., Procedures, for the licensees failure to establish instructions in station procedure CPS 9059.01, Reactor Coolant System Leakage Test, Revision 9b. Specifically, the licensee failed to provide instructions to ensure that the main steam piping between the reactor vessel and the inboard main steam isolation valves were completely drained of water at the completion of testing. The licensee entered this issue into the corrective action program as action request AR 01590671.
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| The inspectors determined that the licensees failure to establish instructions to ensure that the main steam piping between the reactor vessel and the inboard main steam isolation valves were completely drained of water prior to starting up the reactor was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences, and is therefore a finding.
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| Using Manual Chapter 0609, Attachment 4 Initial Characterization of Findings, and Appendix A The Significance Determination Process for Findings at Power the finding was screened against the mitigating systems cornerstone and determined to be of very low safety significance (Green) because the finding was/did not: 1) a deficiency affecting the design or qualification of a mitigating structure, system or component, 2) represent a loss of system and/or function, 3) represent an actual loss of function of a single train for greater than its technical specification allowed outage time, 4) represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant for greater than 24 hours and 5) did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding or severe weather event.
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| The inspectors determined this finding affected the cross cutting area of human performance in the aspect of work management where the organization implements a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. Specifically, the licensee failed to have a plan or provide a control method to ensure the main steam piping was drained prior to commencing reactor start up. (IMC 0301 H.5)
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| Inspection Report# : 2014005 (pdf)
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| Page 8 of 14
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| 1Q/2015 Inspection Findings - Clinton Significance: Sep 30, 2014 Identified By: NRC Item Type: NCV Non-Cited Violation EXCEEDED TECHNICAL SPECIFICATION ALLOWED OUTAGE TIME FOR ELECTRICAL POWER SYSTEMS DUE TO AUXILIARY EQUIPMENT OUT OF SERVICE The inspectors identified a non-cited violation of Technical Specification 3.8.4, "DC Sources - Operating" and Technical Specification 3.8.9, "Distribution Systems - Operating" for failing to enter the technical specifications and complete the associated actions prior to the completion time when auxiliary equipment required to support electrical power system safety function was out of service. Specifically, the licensee removed the division 1 safety related portion of the switchgear cooling system from service to perform maintenance and failed to enter the applicable technical specifications that the was required to support system safety function. The licensee documented this issue in the corrective action program as Issue Report (IR) 01674754.
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| The failure to enter the technical specifications and complete the associated actions prior to the completion time when auxiliary equipment required to support electrical power system safety function was out of service was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences and is therefore a finding. Using Manual Chapter 0609, Appendix A, "The Siginificance Determination Process (SDP) for Findings At-Power," issued June 19, 2012, Exhibit 2 for the Mitigating Systems Cornerstone. The inspectors answered "Yes" to the screening question under the Mitigating Systems Cornerstone "Does the finding represent and actual loss of function of at least a single train for > its Tech Spec Allowed Outage Time OR two separate safety systems out-of-service for > its Tech Spec Allowed Outage Time?,' since the finding represented an actual loss of function of at least a single Train for > its Tech Spec Allowed Outage time. Therefore, a detailed risk evaluation was performed using IMC 0609, Appendix A. The Senior Reactor Analysts (SRAs) evaluated the finding using the Clinton Standardized Plant Analysis Risk (SPAR) model version 8.17, Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) version 8.1.0. For switchgear cooling, independent redundant cooling trains are provided for each of the three divisional switchgear areas with one train being non-safety related and the other safety related. In order to characterize the risk significance, the SRAs assumed that during a loss of offsite power (LOOP) event, the non-safety related switchgear cooling train that is normally in operation would become unavailable. The safety-related cooling train, should it be undergoing maintenance, would be unavailable as well. The exposure time for this issue was taken to be 235 hours based on the licensee documentation. Post-processing rules were used to credit an additional 4.0 hours of time to recover offsite power (to allow recovery of the non-safety cooling train) in core damage sequences when the safety-related cooling train for Division 1 equipment was undergoing maintenance during a LOOP. The SRAs also gave credit in the SPAR Model for local operator action to provide alternate switchgear room cooling during a LOOP. The licensee produced Alarm Response Procedure CPS 5050.03, Rev 30c, which directed operators to Procedure CPS 3412.01, "Essential Switchgear Heat Removal (VX)...,"
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| Revision 15. These procedures directed operators to locally open doors, set up protable blowers, or lower electrical loads to help cool the room as necessary. The SRAs used the SPAR-H Human Reliability Analysis Method (NUREG/CR-6883) to estimate the human error probability for identifying and executing the local actions. the performance drivers were "time" (extra time) and "stress" (high) for diagnosis. The performance drivers were "stress" (high) and "ergonomics" (poor) for action. The resultant human error probability using these assumptions was 0.022. Using the above information, the ?CDF during the exposure time is 1.7E-08/yr. The dominant sequences were station blackout sequences, with initial success of RCIC and HPCS, but later failure of those systems and decay heat removal and all injection due to failure to vent containment and its subsequent failure. Based on the detailed risk evaluation, this finding is best characterized as a finding of very low safety-significance (Green.) The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of avoid complacency where individuals recognize and plan for mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Specifically, the licensee has removed the division 1 or 2 safety related switchgear cooling system fans or condensing units from service numerous times and failed to consider the components inoperable under technical Page 9 of 14
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| 1Q/2015 Inspection Findings - Clinton specification definition for operable. (IMC 0310 H.12)
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| Inspection Report# : 2014004 (pdf)
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| Significance: Sep 30, 2014 Identified By: NRC Item Type: NCV Non-Cited Violation PROGRAMMATIC FAILURE TO COMPLETE OPERABILITY AND FUNCTIONALITY DETERMINATIONS The inspectors identified a non-cited violation of 10 CFR 50 Appendix B, Criterion V, Instructions, Procedures and Drawings, "Procedures," for the failure to accomplish station procedure OP-AA-108-115, "Operability Determinations" Revision 14. Specifically, on multiple occasions operations personnel failed to complete or documented incomplete operability or functionality of safety related or related to safety equipment used at the site.
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| The licensee documented this issue in the corrective action program as Issue Report (IR) 01693256.
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| The failure to complete or provided incomplete operability or functionality determinations used to determine the operability or functionality of safety related or related to safety equipment used at the site is a performance deficiency.
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| The performance deficiency was determined to be more than minor because if left uncorrected, the performance deficiency has the potential to lead to a more significant safety concern and is therefore a finding. Specifically, if operations personnel continue to fail to complete or provide incomplete operability or functionality determination the station could have safety or safety related equipment inoperable without taking appropriate actions for the equipment being inoperable (e.g. entering appropriate technical specification limited condition for operation). Using Manual Chapter 0609, Attachment 4 "Initial Characterization of Findings," and Appendix A "The Significance Determination Process for Findings at Power" the finding was screened against the mitigatins systems cornerstone and determined to be of very low safety significance (Green) because the finding was/did not: 1) a deficiency affecting the design or qualification of a mitigating structure, system or component, 2) represent a loss of system and/or function, 3) represent an actual loss of function of a single train for greater than its technical specification allowed outage time, 4) represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant for greater than 24 hours and 5) did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding or serve weather event. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of Training, where the organization provides training and ensures knowledge transfer to maintain a knowledgeable, technically competent workforce and instill nuclear safety values. Specifically, personnel performing the reviews believed existing training provided sufficient knowledge without the use of additional resources material and current training to operators does not covet this activity. (IMC 0319 H.9)
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| Inspection Report# : 2014004 (pdf)
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| Significance: Sep 30, 2014 Identified By: NRC Item Type: NCV Non-Cited Violation FAILURE TO ESTABLISH SURVEILLANCE PRODEDURE FOR REACTOR CORE ISOLATION COOLING PUMP DUE TO UNACCEPTABLE PRECONDITIONING The inspectors determined that the failure to establish a surveillance procedure to test the RCIC system due to unacceptable preconditioning is a performance deficiency. The performance deficiency was more than minor because it was associated with the procedure quality attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability and capability to response to initiating events to prevent undesirable consequences and is therefore a finding. Using Manual Chapter 0609, Attachment 0609.04 "Initial Characterization of Findings," and Appendix A "The Significance Determination Process for Findings at Power" the finding was screened against the mitigating systems cornerstone and determined to be of very low safety significance (Green) because the finding was/did not: 1) a deficiency affecting the design or qualification of a mitigating structure, Page 10 of 14
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| 1Q/2015 Inspection Findings - Clinton system or component, 2) represent a loss of system and/or function, 3) represent an actual loss of function of a single train for greater than its technical specification allowed outage time, 4) represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant for greater than 24 hours and 5) did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding or severe weather event. The inspectors determined this finding affected the cross-cutting area of problem identification and resolution in the aspect of operating experience where the organization systematically and effectively collects, evaluates and implements relevant internal and external operating experience in a timely manner.
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| Specifically, the licensee considered the impact of the operating experience for surveillance testing, but did not consider its impact during normal plant operation. (IMC 0310 P.5)
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| Inspection Report# : 2014004 (pdf)
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| Significance: Jun 30, 2014 Identified By: NRC Item Type: NCV Non-Cited Violation FOREIGN MATERIAL IN RELAY PREVENTS EMERGENCY DIESEL GENERATOR OUTPUT BREAKER FROM CLOSING The inspectors documented a self-revealing, Green non-cited violation of Clinton Power Station Technical Specification 5.4.1, Procedures, for a failure to prevent foreign material from entering a relay associated with the Division 1 Diesel Generator. Specifically, contrary to station procedure CPS 8501.05, CV-2 Relay Inspection and Calibration with Doble Test Equipment, Revision 4, the licensee failed to verify that relay 227-DGIKA, CV-2 AB phase was clean and free of all foreign material. The foreign material prevented the relay from operating and satisfying the permissive logic required to close the Division 1 Diesel Generator output breaker resulting in having to declare the Diesel Generator inoperable. The relay was replaced and successfully tested and the licensee documented this issue in the corrective action program as IR 01600935.
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| The finding was more than minor because it was associated with the procedure quality attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences and is therefore a finding.
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| Using Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012, Exhibit 2 for the Mitigating Systems Cornerstone, the inspectors answered Yes to the screening question under the Mitigating Systems Cornerstone Does the finding represent an actual loss of function of at least a single Train for > its Tech Spec Allowed Outage Time OR two separate safety systems out-of- service for >
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| its Tech Spec Allowed Outage Time?," since the finding represented an actual loss of function of at least a single Train for > its Tech Spec Allowed Outage Time of 14 days. Therefore, a detailed risk evaluation was performed using IMC 0609, Appendix A. The Senior Reactor Analysts (SRAs) evaluated the finding using the Clinton Standardized Plant Analysis Risk (SPAR) model version 8.17, Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) version 8.1.0 and concluded that the risk increase to the plant due to this finding is very low (Green). The inspectors determined this finding affected the cross cutting area of human performance in the aspect of work management where the organization implements a process of planning, controlling and executing work activities such that nuclear safety is the overriding priority. Specifically, the licensees implementation of their foreign material exclusion process for this maintenance activity lacked sufficient planning, controls and execution to prevent foreign material from entering a risk significant piece of safety related equipment.
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| Inspection Report# : 2014003 (pdf)
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| Significance: Jun 30, 2014 Identified By: NRC Item Type: NCV Non-Cited Violation FAILURE TO DEVELOP ADEQUATE PROCEDURES FOR PRE-PLANNING AND PERFORMING Page 11 of 14
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| 1Q/2015 Inspection Findings - Clinton MAINTENANCE AFFECTING SAFETY-RELATED EQUIPMENT The inspectors documented a self-revealing, Green non-cited violation (NCV) of Clinton Power Station Technical Specification 5.4.1, Procedures for a failure to develop adequate procedures for properly pre-planning and performing maintenance affecting the performance of safety-related equipment which resulted in the subsequent failure of the Division 3 Diesel Room Ventilation damper hydramotor on August 15, 2013. Specifically, during pre-scheduled performance testing of the Division 3 (High Pressure Core Spray System) Emergency Diesel Generator Room Ventilation Damper hydramotor, the ventilation supply air intake damper, 1VD01YC, failed to open as a result of Damper Hydramotor 1TZVD003A experiencing an age-related degradation failure. This was due in part to the licensees failure to properly pre-plan and perform the appropriate preventive maintenance for the hydramotor due to inadequate procedures. Procedure WC-AA-113, Predefine Database Revisions, Revision 2, did not provide adequate instructions appropriate to the circumstances to properly pre-plan and perform maintenance affecting the performance of safety-related equipment. This resulted in a loss of safety function of the HPCS Diesel Generator and its supported High Pressure Core Spray system because of the low confidence that diesel room temperature would be maintained to support the diesel during an event when it would be required to perform its function. The licensee subsequently replaced the hydramotor, tested the new hydramotor successfully and restored the diesel ventilation system to operable. They documented this issue in the corrective action program as IR 1546973 and IR 1547294.
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| The performance deficiency was more than minor because it was associated with the equipment performance attribute of the mitigating systems cornerstone attribute and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences and is therefore a finding. Using Manual Chapter 0609, Appendix A, The SDP for Findings At-Power, issued June 19, 2012, Exhibit 2 for the Mitigating Systems Cornerstone. The inspectors answered Yes to the screening question under the Mitigating Screening Cornerstone Does the finding represent a loss of system and/or function? since the finding resulted in a loss of safety function. Therefore, a detailed risk evaluation was performed using IMC 0609, Appendix A. The SRAs evaluated the finding using the Clinton SPAR model version 8.17, SAPHIRE version 8.1.0 and concluded that the risk increase to the plant due to this finding is very low (Green). The inspectors determined that no cross-cutting aspect will be assigned to this performance deficiency since it occurred in 2005 and is not indicative of current plant performance Inspection Report# : 2014003 (pdf)
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| Barrier Integrity Significance: Dec 31, 2014 Identified By: NRC Item Type: NCV Non-Cited Violation FAILURE TO UPDATE THE UPDATED SAFETY ANALYSIS REPORT - 1VR08C FUNCTION The inspectors identified a Severity Level IV non-cited violation of Title 10 Code of Federal Regulations (CFR) 50.71 (e), Periodic Update of the USAR and an associated Green finding for the licensees failure to update the USAR with the correct description of the function of 1VR08C. Specifically the licensee did not update Section 9.4.5.5 of the USAR to include the correct function of 1VR08C as described in a commitment made to the NRC in letter U-600850.
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| Consequently the licensee performed a 50.59 evaluation for abandoning a portion of the system that did not consider the correct function of the component. The licensee entered this issue into their corrective action program as AR 1692665.
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| The inspectors determined that the failure to update the USAR with the correct function of 1VR08C was a performance deficiency. The performance deficiency was determined to be more than minor in accordance with Page 12 of 14
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| 1Q/2015 Inspection Findings - Clinton Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, because, if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern and is therefore a finding. Specifically, failure to update the USAR with the correct safety related function of VR08C could result in the licensee making operability and functionality determinations based on incorrect assumptions. Additionally, the failure to update the USAR with the correct function of the fan was more than minor because it was associated with the Barrier Integrity Cornerstone attribute of design control, plant modifications and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events.
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| Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings at Power, the finding was screened against the Barrier Integrity cornerstone and determined to be of very low safety significance (Green) because the finding does not represent an actual open pathway in the physical integrity of reactor containment, containment isolation system, and heat removal components and it did not involve an actual reduction in function of hydrogen igniters in reactor containment. The performance deficiency associated with this finding did not reflect current licensee performance; therefore, no cross cutting aspect was identified with this finding.
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| Additionally, in accordance with Section 6.1.d.3 of the NRC Enforcement Policy, this violation was categorized as Severity Level IV because the licensees failure to update the USAR as required by 10 CFR 50.71(e) had not yet resulted in any unacceptable change to the facility or procedures.
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| Inspection Report# : 2014005 (pdf)
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| Emergency Preparedness Significance: Sep 30, 2014 Identified By: NRC Item Type: NCV Non-Cited Violation INCOMPLETE EVACUATION TIME ESTIMATE SUBMITTALS The inspectors determined that Exelon's failure to submit a complete updated ETE for the Clinton Power Station by December 22, 2012 was a performance deficiency. Specifically, the ETE is an input into the development of protective action strategies prior to an accident and to the protective action recommendation decision making process during an accident. Inadequate ETEs have the potential to reduce the effectiveness of public protective actions implemented by the OROs. The performance deficiency was more than minor because it was associated with the procedure quality attribute of the emergency preparedness cornerstone and adversely affected the cornerstone objective to ensure that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency and is therefore a finding. Using IMC 0609, attachment 0609.04 "Initial Characterization of Findings," and Appendix B, "Emergency Preparedenss (EP) Significance Determination Process (SDP)," the finding was screened by the inspectors and determined to be of very low safety significance (Green) based upon the following. The performance deficiency was associated with planning standard 10 CFR 50.47 (b)(10)," Green Finding column, provides the following exasmples "ETEs and updates to the ETEs were not provided to responsible OROs," and "The current public protective action strategies documented in emergency preparedness implementing procedures (EPIPs) are not consistent with the current ETE." The inspectors concluded that the incomplete updated ETE delayed the NRC's approval of the Clinton Power Station ETE, therefore the ETE was not provided to the site OROs nor was it used to inform the site EPIPs as required by 10 CFR 50.47(b)(10), and Section IV, Paragraph 4 of Appendix E to 10 CFR Part 50. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of documentation where the organization creates and maintains complete, Page 13 of 14
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| 1Q/2015 Inspection Findings - Clinton accurate and up-to-date documentation. Specifically, the Emergency Preparedness organization did not develop the Clinton Power Station ETE as required by the new regulation introduced by the NRC's EP Rule. (IMC 0310 H.7)
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| Inspection Report# : 2014004 (pdf)
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| Occupational Radiation Safety Public Radiation Safety Security Although the Security Cornerstone is included in the Reactor Oversight Process assessment program, the Commission has decided that specific information related to findings and performance indicators pertaining to the Security Cornerstone will not be publicly available to ensure that security information is not provided to a possible adversary.
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| Other than the fact that a finding or performance indicator is Green or Greater-Than-Green, security related information will not be displayed on the public web page. Therefore, the cover letters to security inspection reports may be viewed.
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| Miscellaneous Last modified : June 16, 2015 Page 14 of 14
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| 2Q/2015 Inspection Findings - Clinton Clinton 2Q/2015 Plant Inspection Findings Initiating Events Significance: Dec 31, 2014 Identified By: NRC Item Type: NCV Non-Cited Violation STATION PROCEDURES FAILED TO PROVIDE CONTROLS FOR MATERIAL NEAR TRANSFORMERS The inspectors identified a non-cited violation associated with a failure to provide controls for material near the station transformers. Specifically, station procedure CPS 4302.01, "Tornado/High Winds", Revision 21b does not include guidelines or examples of the types of materials to control as potential missiles in high velicity winds or tornadoes, and does not include triggers to perform walkdowns when high winds are predicted, prior to off-normal entry, to control material adjacent to the offsite power transformers that could result in the loss of offsite power. The licensee entered this issue into the corrective action program as action request (AR) 2388608.
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| The failure to provide guidelines or examples of the types of materials to control as potential missiles in high velicity winds or tornadoes and provide triggers to perform walkdowns when high winds are predicted was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations and is therefore a finding. Using Manual Chapter 0609, Attachment 4 "Initial Characterization of Findings," and Appendix A "The Significance Determination Process for Findings at Power", issued June 19, 2012, the finding was screened against the initiating events cornerstone and determined to be of very low safety significance (Green) because the finding did not involve the complete or partial loss of a support system that contributes to the likelihood of, or caused, an initiating event and did not affected mitigation equipment.
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| The inspectors determined this finding affected the cross-cutting area of problem identification and resolution in the aspect of operating experience where the organization systematically and effectively collects, evaluates, and implements relevant internal and external operating experience in a timely manner. Specifically, the licensee opearting experience program failed to ensure evaluation and implementation of interal operating experience in a timely manner after previous identification in the corrective action progrma. (IMC 0310 P.5)
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| Inspection Report# : 2014005 (pdf)
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| Significance: Sep 30, 2014 Identified By: NRC Item Type: NCV Non-Cited Violation FAILURE TO UPDATE THE FINAL SAFETY ANALYSIS REPORT (FSAR) - SD STRUCTURAL INTEGRITY FUNCTION The inspectors identified a Severity Level IV Non-Cited Violation of title 10 Code of Federal Regulations (CFR) 50.71(e), 'Periodic Update of the Final Safety Analysis Report' and an associated Green finding for the licensee's failure to update the Final Safety Analysis Report with a description of the basis for the steam dryer structural integrity submitted to the NRC in support of an extended power uprate license amendment. Specifically, the licensee did not update Section 3.9.5.1.1.9. "Steam Dryers," of the FSAR to include analysis and inspections of the steam dryer Page 1 of 12
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| 2Q/2015 Inspection Findings - Clinton each refueling outage that provided the basis for steam dryer structural integrity. Consequently, the licensee had not completed an inspection of the steam dryer during the most recent refueling outage. The licensee entered this issue into the corrective action program as issue report IR 02223135 and initiated actions to evaluate the Final Safety Analysis Report for revision to include description of the structural integrity function of the steam dryer.
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| The inspectors determined that the licensee's failure to update the Final Safety Analysis Report with the basis for steam dryer structural integrity submitted to the NRC was a performance deficiency. the performance deficiency was determined to be more than minor in accordance with Inspection Manual Chapter 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," dated September 7, 2012, because, if left uncorrected, the performance deficiency would have the potential to lead a more significant safety concern and is therefore a finding. Failure to update the Final Safety Analysis Report with the basis for steam dryer structural integrity could result in a failure to maintain the structural integrity of the steam dryer. Specifically, insuffient steam dryer inspections could result in failure to detect structurally significant cracking and result in a steam dryer failure which generates debris that adversely affect the function of safety-related compoments (e.g. MSIVs). Additionally, the failure to update the Final Safety Analysis Report with the basis for steam dryer structural integrity was more than minor because it was associated with the Initiating Events Cornerstone attribute of Equipment Performance and adversely affected the Cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions.
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| Violations of 10 CFR 50.71(e) are dispositioned using the traditional enforcement process because they are considered to be violations that potentially impede or impact the regulatory process. This violation was also associated with a finding that has been evaluated by the significance determination process (SDP) and communicated with SDP color reflective of the safety impact of the deficient licensee performance. The SDP, however, does not specifically consider regulatory process impact. Thus, although related to a common regulatory concern, it is necessary to address the violation and finding using different processes to correctly reflect both the regulatory importance of the violation and the safety significance of the associated finding.
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| Using Manual Chapter 0609, Attachment 4 "Initial characterization of Findings," and Appendix A "The Significance Determination Process for findings at Power" the finding was screened against the initiating events cornerstone and determined to be of very low safety significance (Green) because the finding did not cause a reactor trip and the loss of mitigating equipment relied upon to transition from the onset of the trip to a stable. The performance deficiency associated with this finding did not reflect current licensee performance; therefore, no cross cutting aspect was identified with this finding.
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| Additionally, in accordance with Section 6.1.d.3 of the NRC Enforcement Policy, this violation was categorized as Severity Level IV because the licensee's failure to update the FSAR as required by 10 CFR 50.71(e) had not yet resulted in any unacceptable change to the facility or procedures.
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| Inspection Report# : 2014004 (pdf)
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| Significance: Sep 30, 2014 Identified By: NRC Item Type: NCV Non-Cited Violation MODIFICATION TO STEAM DRYER TIE BARS 28 AND 30 WITHOUT A 10 CFR 50.59 SAFETY EVALUATION The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.59(d)(1), "Changes, Test, and Experiments" for the licensee's failure to perform a written evaluation, which provided the bases for the determination that a change did not require a license amendment. Specifically, the licensee made a change pursuant to 10 CFR 50.59 (c) with the installation of 1/2 inch holes adjacent to welds attaching tie bars 28 and 30 to the steam dryer vane assembly and did not provide a written evaluation to provide a basis for the determination that this change would not result in a more than minimal increase in the likelihood of occurrence of a malfunction of an system structure or component important to safety (e.g. MSIVs). The licensee entered this finding into the corrective action program as issue report IR 02223135 and identified an action to secure a detailed assessment of these degraded tie bar locations Page 2 of 12
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| 2Q/2015 Inspection Findings - Clinton from the steam dryer vendor. The licensee also consulted with the steam dryer vendor and made a qualitative assessment that the additional unflawed and unaltered portion of the fillet welds present at the end of the tie bar 28 and 30 locations provided a reasonable basis to conclude that these tie bars would not fail and affect the operability of safety-related components.
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| The inspectors determined that the failure to provide a written evaluation, which provided the basis for the determination that a change did not require a license amendment, was a performance deficiency. Specifically, the licensee failed to provide a basis for not applying for a license amendment associated with increased likelihood of a SD failure that impacts safety-related equipment due to reduced structural support available at tie bars 28 and 30. The performance deficiency was determined to be more than minor in accordance with IMC 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," dated September 7, 2012, because it was associated with the Initiating Events cornerstone attribute of equipment performance and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. In addition, the associated violation was determined to be more than minor because the inspectors could not reasonable determine if the changes to the SD at tie bars 28 and 30 would have required NRC prior approval.
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| Violations of 10 CFR 50.59 are dispositioned using the traditional enforcement process instead of the SDP because they are considered to be violations that potentially impede or impact the regulatory process. However, if possible, the underlying technical issue is evaluated under the SDP to determine the severity of the violation. In this case, the inspectors used Manual Chapter 0609, Attachment 4 "Initial characterization of Findings," and Appendix A "The Significance Determination Process for findings a Power" the finding was screened against the initiating events cornerstone and determined to be of very low safety significance (Green) because the finding did not cause a reactor trip and the loss of mitigating equipment relied upon to transition from the onset of the trip to a stable. The performance deficiency associated with this finding did not reflect current licensee performance; therefore, no cross cutting aspect was identified with this finding.
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| In accordance with Section 6.1.d.2 of the NRC Enforcement Policy, this violation was categorized as Severity Level IV because the resulting changes were evaluated by the SDP as having very low safety significance.
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| Inspection Report# : 2014004 (pdf)
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| Significance: Jul 11, 2014 Identified By: NRC Item Type: FIN Finding FAILURE TO IDENTIFY A LEVEL 1 TEST CRITERION FAILURE The inspectors documented a self-revealing Green finding associated with the failure to follow procedures when performing power ascension testing on the digital feedwater (DFW) system. Specifically, contrary to procedure CPS 2894.01, "Digital FWLC [feedwater level control system] Modifications Test - Power Ascension," Section 9.1, the licensee did not declare a Level 1 criterion failure when unacceptable oscillations were noted during a transition in the power ascension test. This resulted in the licensee declaring the test successful and returning the system to service without taking the appropriate corrective actions to address the oscillations. This contributed to the subsequent scram caused by reactor water level oscillations.
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| The failure to follow procedures when performing power ascension testing on the digital feedwater system was a performance deficiency. The performance deficiency was more than minor because it was associated with the design control attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations and is therefore a finding. Using IMC 0609, Attachment 4 "Initial Characterization of Findings," and Appendix A "The Significance Determination Process for Findings at Power," issued June 19, 2012, the finding was determined to be of very low safety significance (Green) because it did not cause a reactor trip with a coincident loss Page 3 of 12
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| 2Q/2015 Inspection Findings - Clinton of mitigating equipment. The inspectors determined this finding affected the conservative bias aspect of the of human performance cross-cutting area described as being present when the organization uses decision making practices that emphasize prudent choices over those that are simply allowable. Specifically, the licensee used non-conservative assumptions when determining whether the condition identified during the power ascension test was allowable (H.14).
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| This finding does not involve enforcement action because no violation of regulatory requirements was identified.
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| Inspection Report# : 2014008 (pdf)
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| Mitigating Systems Significance: Mar 31, 2015 Identified By: NRC Item Type: FIN Finding FAILURE TO PERFORM ADEQUATE CHANNEL CALIBRATION ON SEISMIC INSTRUMENTATION The inspectors identified a Green Finding associated with the licensee's failure to perform an adequate channel calibration to determine the functionality of the stations seismic monitoring equipmen used for evaluaing earthquakes.
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| Specifically, station procedure CPS 9437.21, "Trix Time-History Accelerometer Channel Calibration," Revision 39c, did not include steps to ensure that battery backup power was provided to operate the equipment on a loss of the normal power source as part of the operability requirements. The licensee documented the issue in the corrective action program as action request (AR) 02454630. As a corrective action the licensee planned to correct procedure CPS 9437.21 to verify proper battery operation.
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| The licensee's failure to perform an adequate channel calibration to determine the functionality of the stations seismic monitoring equipment used for evaluating earthquakes was a performance deficiency. Specifically, station procedures did not include steps to ensure that battery backup power was provided to operate the equipment on a loss of the normal power source. The performance deficiency was more than minor because it adversely impacted the protection against external factors attribute of the Mitigating Systems cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual chapter 0609, Attachment 4 "Initial Characterization of Findings," and Appendix A "The Significance Determination Process for Findings at Power," issued June 19, 2012, the inspectors answered "yes" to the Mitigating Systems cornerstone question, "Does the finding involve the ... degradation of equipment ... specificallty designed to mitigate a seismic ...
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| initiating event ..." Therefore, the inspectors addressed the questions in Exhibit 4, "External Event Screening Questions." The inspectors answered "no" to the two questions in Exhibit 4. Specifically, 1) if completely failed the seismic monitor would not cause an initiating event or degrade multi-trains or risk-significant systems; and 2) the finding does not involve the total loss of any safety function. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of conservative bias where individuals use decision making-practices that emphasize prudent choices over those that are simply allowable and a proposed action is determined to be a safe in order to proceed, rather than unsafe in order to stop. Specifically, the licensee documented the issue of the voltage being high out of specification and instead of performing additional corrective actions to determine if leaving the voltage out of specification was appropriate marked the step as not applicable and proceeded with the rest of the procedure. (Section 1R15)
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| Inspection Report# : 2015001 (pdf)
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| Significance: Mar 31, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation UNQUALIFIED SAFETY-RELATED CABLES USED IN A SUBMERGED ENVIRONMENT Page 4 of 12
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| 2Q/2015 Inspection Findings - Clinton The inspectors identified a finding and an associated non-citied violation of 10 CFR 50 Appendix B, Criterion III, "Design Control," for the failure to maintain safety-related cables for the SX system in an environment for which they were designed. Specifically, the licensee failed to maintain SX safety-related cables in an environment for which they were designed when the cables were allowed to be submerged in water inside cable vaults. The licensee documented this issue in their corrective action program (CAP) as action request (AR) 02474543. Corrective actions included draining the cable vaults so that the cables were no longer submerged.
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| The licensee's failure to maintain safety-related cables for the SX system in an environment for which they were designed was a performance deficiency. The performance deficiency was determined to be more than minor because the finding was associated with the Mitigating Systems Cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licnesee failed to maintain SX safety-related cables in an environment for which they were designed when the cables were allowed to be submerged in water inside cable vaults. Using IMC 0609, "Significance Determination Process," Attachment 0609.04, "Initial Characterization of Findings," issued on June 19, 2012. Specifically, the inspectors used IMC 0609 Appendix A "SDP for Findings At-Power," issued June 19, 2012, Exhibit 2, "Mitigating Systems Screening Questions" to screen the finding. The finding screened as of very low safety significance (Green) because the inspectors answered yes to the question "does the SSC maintain its operability or functionality." Specifically, the SX system submerged cables did not cause the SX system to be inoperable or nonfunctional. The inspectors determined this finding affected the cross-cutting area of problem identification and resolution in the aspect of resolution, where the organization takes effective corrective actions to address issues in a timely manner commensurate with their safety significance. Specifically, the licensee failed to implement effective corrective actions to address an adverse trend of water in cable vaults which led to (SX) safety-related cables being submerged in water.
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| Inspection Report# : 2015001 (pdf)
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| Significance: Mar 31, 2015 Identified By: NRC Item Type: AV Apparent Violation FAILURE OF THE DIVISION 3 SHUTDOWN SERVICE WATER PUMP DUE TO AN INADEQUATE BUSHING DESIGN A self-revealed finding, preliminarily determined to be of low to moderate safety significance (White) and an associated AV of 10 CFR 50 Appendix B, Criterion III, Design Control, was identified for the failure to verify the suitability of the replacement pump design for the Division 3 Shutdown Service Water system. Spceifically, the design of the suction bell bushing for the replacement pump was inadequate to pass sufficient cooling water flow to the pump internals without being affected by mud and silt from the lake water. This finding was self-revealed on September 16, 2014, during a surveillance test to ensure operability of the Division 3 shutdown cooling water pump after the pump failed to start due to a damaged bushing rendering the pump inoperable. This finding does not represent an immediate safety concern because the licensee replaced the pump in September of 2014 with a pump of similar design and provided adequate documentation that assures the pump will remain operable until a different design for the bushing that failed can be installed by June of 2016.
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| The inspectors determined that the licensee's failure to verify the suitability of the design for the Division 3 Shutdown Service Water pump was a performance deficiency warranting a significance evaluation. The inspectors determined that the finding was more than minor in accordance with IMC 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," dated September 7, 2012, because it was associated with the Mitigating Systems Cornerstone attributes of design control and equipment performance and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. A Significance and Enforcement Review Panel (SERP), using IMC 0609, Appendix A, "Significance Determination Process For Findings At-Power," dated June 19, 2012, preliminarily determined the finding to be of low to moderate safety significance (White). The performance deficiency associated with this finding did not reflect Page 5 of 12
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| 2Q/2015 Inspection Findings - Clinton current licensee performance; therefore, no cross cutting aspect was identified with this finding.
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| Inspection Report# : 2015001 (pdf)
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| Significance: Mar 20, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation Inadequate 50.59 Evaluation for Switchgear in Seismically Unanalyzed Conditions (Section 1R17.1b.)
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| Severity Level IV Green. The inspectors identified a finding of very-low safety significance, and an associated Non-Cited Violation of Title 10, Code of Federal Regulations Part 50, Section 59, Changes, Tests and Experiments, (effective January 1, 1997) for a procedure change dated May 2, 1997, where the licensee allowed safety-related switchgear to operate for a limited period of time during plant operation in equipment configurations that were seismically unanalyzed. Specifically, for Safety Evaluation Log 97 060, CPS [Clinton Power Station]
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| Procedure No. 1014.11, Revision 0, the licensee failed to include a written safety evaluation which provided the bases that concluded for all switchgear configurations that a seismically unanalyzed condition does not involve an unreviewed safety question, and the possibility for a malfunction of a different type than any evaluated previously in the Safety Analysis Report may be created. The licensee entered the issue into their Corrective Action Program as Action Request 02471583, NRC Mod 50.59 Inspection Safety Eval 97 060 for CPS 1014.11, dated March 20, 2015.
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| The performance deficiency was determined to be more than minor because the finding was associated with the Mitigating Systems cornerstone attribute of protection against external factors and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, switchgear in a seismically unanalyzed condition when relied upon to perform a safety function did not ensure the availability, reliability, or capability of the associated Mitigating Systems to respond to an initiating event such as an earthquake. The inspectors determined that the underlying technical issue was of very-low safety significance (Green) using a detailed risk evaluation. The inspectors did not identify a cross-cutting aspect associated with the finding because the finding was not representative of current performance.
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| Inspection Report# : 2015008 (pdf)
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| Significance: Dec 31, 2014 Identified By: NRC Item Type: NCV Non-Cited Violation FAILURE TO TRANSLATE SEISMIC DESIGN REQUIREMENTS INTO APPLICABLE PROCEDURES The inspectors identified a green non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for the failure to adequately translate seismic requirements from a design calculation into applicable procedures. Specifically the licensee failed to incorporate the seismic requirements for the Division III 4.16 KV switchgear as described in calculation IP-Q-0391 Seismic Qualification of 480V ABB Unit Sub Switchgears, Div I & II Westinghouse Switchgears and Div III GE 4.16KV Switchgears, into procedure CPS 1014.11 6900/4160/480V Switchgear/Circuit Breaker Operability Program, resulting in the licensee incorrectly declaring Division III switchgear operable when in a seismically unanalyzed condition. The licensee entered this issue into their corrective action program as AR 2386676.
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| The inspectors determined that the failure to adequately incorporate the seismic requirements of the design calculation into the applicable procedure was a performance deficiency. The finding is more than minor because it is associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Appendix A, The Significance Determination Page 6 of 12
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| 2Q/2015 Inspection Findings - Clinton Process for Findings At-Power, Exhibit 4, External Events Screening Questions, dated June 19, 2012, the inspectors answered Yes to question 1 of External Events screening questions, because the finding could potentially degrade one train of the emergency power system. Thus the inspectors consulted the regional senior reactor analyst (SRA).
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| Based on the Detailed Risk Evaluation, the inspectors determined that the finding was of very low safety significance (Green). The inspectors determined that there was no cross-cutting aspect associated with this finding because the cause of the performance deficiency occurred more than fifteen years ago, and was not representative of current licensee performance.
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| Inspection Report# : 2014005 (pdf)
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| Significance: Dec 31, 2014 Identified By: Self-Revealing Item Type: NCV Non-Cited Violation FAILURE TO PROVIDE PROCEDURE INSTRUCTION RESULTS IN EXCEEDING TECHNICAL SPECIFICATION HEAT UP RATE DURING PLANT START UP The inspectors are documenting a self-revealing non-cited violation of Technical Specification 5.4., Procedures, for the licensees failure to establish instructions in station procedure CPS 9059.01, Reactor Coolant System Leakage Test, Revision 9b. Specifically, the licensee failed to provide instructions to ensure that the main steam piping between the reactor vessel and the inboard main steam isolation valves were completely drained of water at the completion of testing. The licensee entered this issue into the corrective action program as action request AR 01590671.
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| The inspectors determined that the licensees failure to establish instructions to ensure that the main steam piping between the reactor vessel and the inboard main steam isolation valves were completely drained of water prior to starting up the reactor was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences, and is therefore a finding.
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| Using Manual Chapter 0609, Attachment 4 Initial Characterization of Findings, and Appendix A The Significance Determination Process for Findings at Power the finding was screened against the mitigating systems cornerstone and determined to be of very low safety significance (Green) because the finding was/did not: 1) a deficiency affecting the design or qualification of a mitigating structure, system or component, 2) represent a loss of system and/or function, 3) represent an actual loss of function of a single train for greater than its technical specification allowed outage time, 4) represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant for greater than 24 hours and 5) did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding or severe weather event.
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| The inspectors determined this finding affected the cross cutting area of human performance in the aspect of work management where the organization implements a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. Specifically, the licensee failed to have a plan or provide a control method to ensure the main steam piping was drained prior to commencing reactor start up. (IMC 0301 H.5)
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| Inspection Report# : 2014005 (pdf)
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| Significance: Sep 30, 2014 Identified By: NRC Item Type: NCV Non-Cited Violation Page 7 of 12
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| 2Q/2015 Inspection Findings - Clinton EXCEEDED TECHNICAL SPECIFICATION ALLOWED OUTAGE TIME FOR ELECTRICAL POWER SYSTEMS DUE TO AUXILIARY EQUIPMENT OUT OF SERVICE The inspectors identified a non-cited violation of Technical Specification 3.8.4, "DC Sources - Operating" and Technical Specification 3.8.9, "Distribution Systems - Operating" for failing to enter the technical specifications and complete the associated actions prior to the completion time when auxiliary equipment required to support electrical power system safety function was out of service. Specifically, the licensee removed the division 1 safety related portion of the switchgear cooling system from service to perform maintenance and failed to enter the applicable technical specifications that the was required to support system safety function. The licensee documented this issue in the corrective action program as Issue Report (IR) 01674754.
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| The failure to enter the technical specifications and complete the associated actions prior to the completion time when auxiliary equipment required to support electrical power system safety function was out of service was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences and is therefore a finding. Using Manual Chapter 0609, Appendix A, "The Siginificance Determination Process (SDP) for Findings At-Power," issued June 19, 2012, Exhibit 2 for the Mitigating Systems Cornerstone. The inspectors answered "Yes" to the screening question under the Mitigating Systems Cornerstone "Does the finding represent and actual loss of function of at least a single train for > its Tech Spec Allowed Outage Time OR two separate safety systems out-of-service for > its Tech Spec Allowed Outage Time?,' since the finding represented an actual loss of function of at least a single Train for > its Tech Spec Allowed Outage time. Therefore, a detailed risk evaluation was performed using IMC 0609, Appendix A. The Senior Reactor Analysts (SRAs) evaluated the finding using the Clinton Standardized Plant Analysis Risk (SPAR) model version 8.17, Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) version 8.1.0. For switchgear cooling, independent redundant cooling trains are provided for each of the three divisional switchgear areas with one train being non-safety related and the other safety related. In order to characterize the risk significance, the SRAs assumed that during a loss of offsite power (LOOP) event, the non-safety related switchgear cooling train that is normally in operation would become unavailable. The safety-related cooling train, should it be undergoing maintenance, would be unavailable as well. The exposure time for this issue was taken to be 235 hours based on the licensee documentation. Post-processing rules were used to credit an additional 4.0 hours of time to recover offsite power (to allow recovery of the non-safety cooling train) in core damage sequences when the safety-related cooling train for Division 1 equipment was undergoing maintenance during a LOOP. The SRAs also gave credit in the SPAR Model for local operator action to provide alternate switchgear room cooling during a LOOP. The licensee produced Alarm Response Procedure CPS 5050.03, Rev 30c, which directed operators to Procedure CPS 3412.01, "Essential Switchgear Heat Removal (VX)...,"
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| Revision 15. These procedures directed operators to locally open doors, set up protable blowers, or lower electrical loads to help cool the room as necessary. The SRAs used the SPAR-H Human Reliability Analysis Method (NUREG/CR-6883) to estimate the human error probability for identifying and executing the local actions. the performance drivers were "time" (extra time) and "stress" (high) for diagnosis. The performance drivers were "stress" (high) and "ergonomics" (poor) for action. The resultant human error probability using these assumptions was 0.022. Using the above information, the ?CDF during the exposure time is 1.7E-08/yr. The dominant sequences were station blackout sequences, with initial success of RCIC and HPCS, but later failure of those systems and decay heat removal and all injection due to failure to vent containment and its subsequent failure. Based on the detailed risk evaluation, this finding is best characterized as a finding of very low safety-significance (Green.) The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of avoid complacency where individuals recognize and plan for mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Specifically, the licensee has removed the division 1 or 2 safety related switchgear cooling system fans or condensing units from service numerous times and failed to consider the components inoperable under technical specification definition for operable. (IMC 0310 H.12)
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| Inspection Report# : 2014004 (pdf)
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| Page 8 of 12
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| 2Q/2015 Inspection Findings - Clinton Significance: Sep 30, 2014 Identified By: NRC Item Type: NCV Non-Cited Violation PROGRAMMATIC FAILURE TO COMPLETE OPERABILITY AND FUNCTIONALITY DETERMINATIONS The inspectors identified a non-cited violation of 10 CFR 50 Appendix B, Criterion V, Instructions, Procedures and Drawings, "Procedures," for the failure to accomplish station procedure OP-AA-108-115, "Operability Determinations" Revision 14. Specifically, on multiple occasions operations personnel failed to complete or documented incomplete operability or functionality of safety related or related to safety equipment used at the site.
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| The licensee documented this issue in the corrective action program as Issue Report (IR) 01693256.
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| The failure to complete or provided incomplete operability or functionality determinations used to determine the operability or functionality of safety related or related to safety equipment used at the site is a performance deficiency.
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| The performance deficiency was determined to be more than minor because if left uncorrected, the performance deficiency has the potential to lead to a more significant safety concern and is therefore a finding. Specifically, if operations personnel continue to fail to complete or provide incomplete operability or functionality determination the station could have safety or safety related equipment inoperable without taking appropriate actions for the equipment being inoperable (e.g. entering appropriate technical specification limited condition for operation). Using Manual Chapter 0609, Attachment 4 "Initial Characterization of Findings," and Appendix A "The Significance Determination Process for Findings at Power" the finding was screened against the mitigatins systems cornerstone and determined to be of very low safety significance (Green) because the finding was/did not: 1) a deficiency affecting the design or qualification of a mitigating structure, system or component, 2) represent a loss of system and/or function, 3) represent an actual loss of function of a single train for greater than its technical specification allowed outage time, 4) represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant for greater than 24 hours and 5) did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding or serve weather event. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of Training, where the organization provides training and ensures knowledge transfer to maintain a knowledgeable, technically competent workforce and instill nuclear safety values. Specifically, personnel performing the reviews believed existing training provided sufficient knowledge without the use of additional resources material and current training to operators does not covet this activity. (IMC 0319 H.9)
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| Inspection Report# : 2014004 (pdf)
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| Significance: Sep 30, 2014 Identified By: NRC Item Type: NCV Non-Cited Violation FAILURE TO ESTABLISH SURVEILLANCE PRODEDURE FOR REACTOR CORE ISOLATION COOLING PUMP DUE TO UNACCEPTABLE PRECONDITIONING The inspectors determined that the failure to establish a surveillance procedure to test the RCIC system due to unacceptable preconditioning is a performance deficiency. The performance deficiency was more than minor because it was associated with the procedure quality attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability and capability to response to initiating events to prevent undesirable consequences and is therefore a finding. Using Manual Chapter 0609, Attachment 0609.04 "Initial Characterization of Findings," and Appendix A "The Significance Determination Process for Findings at Power" the finding was screened against the mitigating systems cornerstone and determined to be of very low safety significance (Green) because the finding was/did not: 1) a deficiency affecting the design or qualification of a mitigating structure, system or component, 2) represent a loss of system and/or function, 3) represent an actual loss of function of a single train for greater than its technical specification allowed outage time, 4) represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant for greater than 24 hours Page 9 of 12
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| 2Q/2015 Inspection Findings - Clinton and 5) did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding or severe weather event. The inspectors determined this finding affected the cross-cutting area of problem identification and resolution in the aspect of operating experience where the organization systematically and effectively collects, evaluates and implements relevant internal and external operating experience in a timely manner.
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| Specifically, the licensee considered the impact of the operating experience for surveillance testing, but did not consider its impact during normal plant operation. (IMC 0310 P.5)
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| Inspection Report# : 2014004 (pdf)
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| Barrier Integrity Significance: Dec 31, 2014 Identified By: NRC Item Type: NCV Non-Cited Violation FAILURE TO UPDATE THE UPDATED SAFETY ANALYSIS REPORT - 1VR08C FUNCTION The inspectors identified a Severity Level IV non-cited violation of Title 10 Code of Federal Regulations (CFR) 50.71 (e), Periodic Update of the USAR and an associated Green finding for the licensees failure to update the USAR with the correct description of the function of 1VR08C. Specifically the licensee did not update Section 9.4.5.5 of the USAR to include the correct function of 1VR08C as described in a commitment made to the NRC in letter U-600850.
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| Consequently the licensee performed a 50.59 evaluation for abandoning a portion of the system that did not consider the correct function of the component. The licensee entered this issue into their corrective action program as AR 1692665.
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| The inspectors determined that the failure to update the USAR with the correct function of 1VR08C was a performance deficiency. The performance deficiency was determined to be more than minor in accordance with Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, because, if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern and is therefore a finding. Specifically, failure to update the USAR with the correct safety related function of VR08C could result in the licensee making operability and functionality determinations based on incorrect assumptions. Additionally, the failure to update the USAR with the correct function of the fan was more than minor because it was associated with the Barrier Integrity Cornerstone attribute of design control, plant modifications and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events.
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| Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings at Power, the finding was screened against the Barrier Integrity cornerstone and determined to be of very low safety significance (Green) because the finding does not represent an actual open pathway in the physical integrity of reactor containment, containment isolation system, and heat removal components and it did not involve an actual reduction in function of hydrogen igniters in reactor containment. The performance deficiency associated with this finding did not reflect current licensee performance; therefore, no cross cutting aspect was identified with this finding.
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| Additionally, in accordance with Section 6.1.d.3 of the NRC Enforcement Policy, this violation was categorized as Severity Level IV because the licensees failure to update the USAR as required by 10 CFR 50.71(e) had not yet resulted in any unacceptable change to the facility or procedures.
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| Inspection Report# : 2014005 (pdf)
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| Page 10 of 12
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| 2Q/2015 Inspection Findings - Clinton Emergency Preparedness Significance: Sep 30, 2014 Identified By: NRC Item Type: NCV Non-Cited Violation INCOMPLETE EVACUATION TIME ESTIMATE SUBMITTALS The inspectors determined that Exelon's failure to submit a complete updated ETE for the Clinton Power Station by December 22, 2012 was a performance deficiency. Specifically, the ETE is an input into the development of protective action strategies prior to an accident and to the protective action recommendation decision making process during an accident. Inadequate ETEs have the potential to reduce the effectiveness of public protective actions implemented by the OROs. The performance deficiency was more than minor because it was associated with the procedure quality attribute of the emergency preparedness cornerstone and adversely affected the cornerstone objective to ensure that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency and is therefore a finding. Using IMC 0609, attachment 0609.04 "Initial Characterization of Findings," and Appendix B, "Emergency Preparedenss (EP) Significance Determination Process (SDP)," the finding was screened by the inspectors and determined to be of very low safety significance (Green) based upon the following. The performance deficiency was associated with planning standard 10 CFR 50.47 (b)(10)," Green Finding column, provides the following exasmples "ETEs and updates to the ETEs were not provided to responsible OROs," and "The current public protective action strategies documented in emergency preparedness implementing procedures (EPIPs) are not consistent with the current ETE." The inspectors concluded that the incomplete updated ETE delayed the NRC's approval of the Clinton Power Station ETE, therefore the ETE was not provided to the site OROs nor was it used to inform the site EPIPs as required by 10 CFR 50.47(b)(10), and Section IV, Paragraph 4 of Appendix E to 10 CFR Part 50. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of documentation where the organization creates and maintains complete, accurate and up-to-date documentation. Specifically, the Emergency Preparedness organization did not develop the Clinton Power Station ETE as required by the new regulation introduced by the NRC's EP Rule. (IMC 0310 H.7)
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| Inspection Report# : 2014004 (pdf)
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| Occupational Radiation Safety Public Radiation Safety Security Although the Security Cornerstone is included in the Reactor Oversight Process assessment program, the Commission has decided that specific information related to findings and performance indicators pertaining to the Security Cornerstone will not be publicly available to ensure that security information is not provided to a possible adversary.
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| Other than the fact that a finding or performance indicator is Green or Greater-Than-Green, security related information will not be displayed on the public web page. Therefore, the cover letters to security inspection reports may be viewed.
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| Page 11 of 12
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| 2Q/2015 Inspection Findings - Clinton Miscellaneous Last modified : August 07, 2015 Page 12 of 12
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| 3Q/2015 Inspection Findings - Clinton Clinton 3Q/2015 Plant Inspection Findings Initiating Events Significance: Sep 30, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation FAILURE TO FOLLOW PROCEDURE LEAVES CONTROL ROOM CABINET DOORS UNATTENDED IN SEISMICALLY UNANALYSED CONDITION The inspectors identified a finding of very low safety significance (Green) and an associated NCV of 10 CFR50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to maintain control room doors in a seismically analyzed condition, in accordance with station procedure CPS 1014.11, 6900/4160/480v Switchgear/Circuit Breaker Operability Program, Revision 5a. Specifically, on several occasions the licensee failed to maintain control room cabinet doors in seismically qualified positions, while performing maintenance or trouble shooting activities, by leaving the doors open and unattended. The licensee documented the issue in the Corrective Action Program (CAP) as action request (AR) 02518477. The licensee has revised the station procedure to ensure control room cabinet doors either remain latched closed or are completely removed when unattended and has issued a standing order to ensure the requirements are reinforced.
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| The performance deficiency was more than minor in accordance with IMC 0612, "Power Reactor Inspection Reports,"
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| Appendix B, "Issue Screening," dated September 7, 2012, because, it was associated with the configuration control performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations and is therefore a finding. Specifically, leaving the control doors in a seismically unanalyzed condition could challenge critical safety functions during a seismic event. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process (SDP) for Findings at Power, issued June 19, 2012, the finding was screened against the Initiating Events cornerstone and determined to be of very low safety significance (Green) because the finding did not result in exceeding the reactor coolant system leak rate for a small loss of coolant accident (LOCA), cause a reactor trip, involve the complete or partial loss of a support system that contributes to the likelihood of, or caused, an initiating event and did not affect mitigation equipment. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of resources where leaders ensure that personnel, equipment, procedures and other resources are available and adequate to support nuclear safety. Specifically, the licensee failed to ensure the personnel performing maintenance and troubleshooting had adequate documentation in written work instructions to maintain control room cabinets in seismically analyzed conditions.
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| Inspection Report# : 2015003 (pdf)
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| Significance: Jun 30, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation POST MAINTENANCE TEST FAILED TO DEMONSTRATE REQUIRED FLOW THROUGH RCIC ROOM COOLER A self-revealed finding of very low safety significance and an associated Non-Cited Violation of 10 CFR 50 Page 1 of 13
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| 3Q/2015 Inspection Findings - Clinton Appendix B, Criterion XI, Test Control, was documented by the inspectors for the failure to perform adequate post maintenance testing that would assure that the Reactor Core Isolation Cooling (RCIC) room cooler would perform its intended function when restored to service following maintenance. Specifically, the licensee declared the room cooler operable with insufficient cooling flow through the cooler. The licensee documented the issue in the licensees corrective action program (CAP) as action request (AR) 02447013. The licensee operated the RCIC Room Cooler outlet valve from its throttled position to fully open to flush the seat and the upstream piping and positioned the valve to maintain the required flow to restore the cooler to an operable condition.
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| The failure to perform adequate post maintenance testing that would assure that the RCIC Room cooler would perform its intended function when restored to service following maintenance is a performance deficiency. The performance deficiency was determined to be more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, because it was associated with the procedure quality attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences and is, therefore, a finding. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings at Power, dated June 19, 2012, the finding was screened against the Mitigating Systems cornerstone and determined to need a detailed risk evaluation because the finding represents the loss of a system and/or function. The Region III Senior Reactor Analysts (SRAs) evaluated the finding using the Clinton Station Standardized Plant Analysis Risk (SPAR) Model Version 8.17, Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) Version 8.1.2. The SRAs reviewed the licensees Apparent Cause Investigation Report IR 2447013. The exposure time was assumed to be 150.5 hours based on information in that report. The SRAs modeled the condition using failure of the RCIC pump as a surrogate for failure of the RCIC room cooler. The basic event representing the RCIC pump failure-to-run was set to True for the 150.5 hour duration. The result was a ?CDF of 9.98E-08/yr. The dominant sequence was a station blackout initiating event; failure of high pressure core spray; failure of reactor core isolation cooling; and failure to recover offsite or emergency AC power within 30-minutes. Based on the detailed risk evaluation, the finding is best characterized as a finding of very low safety-significance (Green). The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of work management where the organization implements a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. The work process includes the identification and management of risk commensurate to the work and the need for coordination with different groups of job activities. Specifically, the licensee failed to plan and execute adequate post maintenance testing that would have ensured the satisfactory operation of the RCIC Room cooler following planned maintenance.
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| [H.5]
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| Inspection Report# : 2015002 (pdf)
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| Significance: Jun 30, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation FAILURE TO TRANSLATE SUFFICIENT GLAND STRESS TO PACKING GLAND NUTS RESULTED IN VALVE PACKING FAILURE AND PLANT SHUTDOWN A finding of very low safety significance and an associated Non-Cited Violation of 10 CFR50, Appendix B, Criterion III, Design Control, was self-revealed on January 19, 2015, when a steam leak developed from the RCIC system inboard steam isolation valve (1E51F0063) stem packing. Specifically, the licensee failed to identify and implement a torque value for the gland packing nuts for the RCIC system inboard steam isolation valve 1E51F0063 to overcome service induced consolidation and prevent packing leakage. This resulted in a plant down power to 83 percent and subsequent plant shutdown due to increasing unidentified reactor coolant system leakage. The licensee documented the issue in the licensees CAP as AR 02439437. The licensee repacked the valve utilizing the station procedure CPS 8120.37, Valve Packing Installation, and applicable SealPro data sheet. A four ring set of A.P. Services graphite packing was installed with a new live load assembly sized to a new torque value of 59 ft-lbs. and the valve packing was tested to verify no leakage.
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| Page 2 of 13
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| 3Q/2015 Inspection Findings - Clinton The inspectors determined that the failure to apply sufficient packing gland torque to overcome service induced consolidation and prevent packing leakage on the RCIC system inboard steam isolation valve was a performance deficiency. The performance deficiency was more than minor in accordance with IMC 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," dated September 7, 2012, because, it was associated with the equipment performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations and is therefore a finding. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings at Power, issued June 19, 2012, the finding was screened against the Initiating Events cornerstone and determined to be of very low safety significance (Green) because the finding did not result in exceeding the RCS leak rate for a small LOCA, cause a reactor trip, involve the complete or partial loss of a support system that contributes to the likelihood of, or caused, an initiating event and did not affect mitigation equipment. The inspectors determined that no cross-cutting aspect would be associated with this finding since the performance deficiency occurred in 2010 and was not representative of current licensee performance in the of valve packing.
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| Inspection Report# : 2015002 (pdf)
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| Significance: Jun 30, 2015 Identified By: NRC Item Type: FIN Finding FAILURE TO EVALUATE THE OPERATIONAL IMPACT OF THE TDRFP LOCKOUT SWITCH POSITION A self-revealed finding was identified for failure to evaluate the consequences of an adverse condition, in accordance with the operational decision making process. Specifically, contrary to station procedure OP-AA-106-101-1006 Operational Decision Making Process, Revision 14, the licensee failed to adequately implement the procedure to ensure the consequences of leaving the switch in the lockout position were evaluated, which resulted in the loss of the manual trip function for the A turbine driven reactor feed pump (TDRFP). The licensee documented the issue in the licensees CAP as action request (AR) 02440052. The licensee repaired the ground condition and returned the switch to its normal position. The licensee also revised the surveillance procedure to document the limitations associated with putting the emergency governor trip test and lockout switch in the lockout position.
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| The inspectors determined that the failure to adequately implement the procedure to ensure the consequences of leaving the switch in the lockout position were evaluated, which resulted in the loss of the manual trip function for the A TDRFP, was a performance deficiency. Specifically, by not evaluating leaving the emergency governor trip test and lockout switch in the lockout position, the licensee lost the ability to manually trip the A TDRFP, which challenged the operators during the reactor shutdown, and nearly resulted in a Level 8 reactor SCRAM. The performance deficiency was more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, because if left uncorrected the performance deficiency had the potential to lead to a more significant safety concern. The performance deficiency was also associated with the configuration control attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown and power operations and is therefore a finding. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings at Power, issued June 19, 2012, the finding was screened against the Initiating Events cornerstone and determined to be of very low safety significance (Green) because the finding did not cause a reactor trip or the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of resources where leaders ensure that personnel, equipment, procedures, and other resources are available and adequate to support nuclear safety. Specifically, the surveillance procedure for the RFPT emergency governor and trip mechanism test Section 2.1.1 stated if an actual signal was generated during testing, the lockout valve would de-energize to allow the trip mechanism to operate and trip the RFPT, which led to the understanding that the trip functions were unaffected by the switch position. (H.1)
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| Page 3 of 13
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| 3Q/2015 Inspection Findings - Clinton Inspection Report# : 2015002 (pdf)
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| Significance: Dec 31, 2014 Identified By: NRC Item Type: NCV Non-Cited Violation STATION PROCEDURES FAILED TO PROVIDE CONTROLS FOR MATERIAL NEAR TRANSFORMERS The inspectors identified a non-cited violation associated with a failure to provide controls for material near the station transformers. Specifically, station procedure CPS 4302.01, "Tornado/High Winds", Revision 21b does not include guidelines or examples of the types of materials to control as potential missiles in high velicity winds or tornadoes, and does not include triggers to perform walkdowns when high winds are predicted, prior to off-normal entry, to control material adjacent to the offsite power transformers that could result in the loss of offsite power. The licensee entered this issue into the corrective action program as action request (AR) 2388608.
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| The failure to provide guidelines or examples of the types of materials to control as potential missiles in high velicity winds or tornadoes and provide triggers to perform walkdowns when high winds are predicted was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations and is therefore a finding. Using Manual Chapter 0609, Attachment 4 "Initial Characterization of Findings," and Appendix A "The Significance Determination Process for Findings at Power", issued June 19, 2012, the finding was screened against the initiating events cornerstone and determined to be of very low safety significance (Green) because the finding did not involve the complete or partial loss of a support system that contributes to the likelihood of, or caused, an initiating event and did not affected mitigation equipment.
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| The inspectors determined this finding affected the cross-cutting area of problem identification and resolution in the aspect of operating experience where the organization systematically and effectively collects, evaluates, and implements relevant internal and external operating experience in a timely manner. Specifically, the licensee opearting experience program failed to ensure evaluation and implementation of interal operating experience in a timely manner after previous identification in the corrective action progrma. (IMC 0310 P.5)
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| Inspection Report# : 2014005 (pdf)
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| Mitigating Systems Significance: Sep 30, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation FAILURE TO IMPLEMENT AND COMPLY WITH TRANSIENT EQUIPMENT/MATERIALS PROGRAM The inspectors identified a green finding and an associated NCV of 10 CFR 50, Appendix B, Criterion V Instructions, Procedures, and Drawings for the licensees failure to implement and comply with station procedure CPS 1019.05, Transient Equipment/Materials, Revision 23, to ensure that transient equipment and materials are controlled so there is no impact to safe operation of plant equipment. Specifically, on numerous occasions the inspectors identified equipment and materials improperly staged, improperly secured or in areas without engineering evaluations. The licensee documented the issue in the CAP as action requests (AR) 02507167 and AR 02529227. In each occasion identified by the inspectors the licensee subsequently removed the items identified to restore Page 4 of 13
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| 3Q/2015 Inspection Findings - Clinton compliance with the station procedures.
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| The inspectors determined the licensees failure to implement and comply with station procedures to ensure that transient equipment and materials are controlled so there is no impact to safe operation of plant equipment was a performance deficiency. The performance deficiency was determined to be more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Screening, dated September 7, 2012, because if left uncorrected it had the potential to lead to a more significant safety concern.
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| Specifically, transient equipment and material in proximity of safety related components has the potential of impacting these components during a seismic event, potentially rendering them unable to fulfill their safety function.
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| The performance deficiency is also associated with the protection against external factors attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability and capability of systems that response to initiating events to prevent undesirable consequences, and is therefore a finding.
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| Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process (SDP) for Findings at Power, issued June 19, 2012, the finding was screened against the Mitigating Systems cornerstone and determined to be of very low safety significance (Green) because the finding did not represent a loss of system or function, it did not represent an actual loss of function of at least a single train for >
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| its TS allowed outage time and it did not represent an actual loss of one or more not TS trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of field presence where leaders are commonly seen in the work areas of the plant observing, coaching, reinforcing standards and expectation. Deviations from standards and expectations are corrected promptly. Specifically, after various examples of material placement being an issue, the licensee didnt perform in field observations, caching and reinforcement of standards and expectations in the identified areas.
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| Inspection Report# : 2015003 (pdf)
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| Significance: Mar 31, 2015 Identified By: NRC Item Type: FIN Finding FAILURE TO PERFORM ADEQUATE CHANNEL CALIBRATION ON SEISMIC INSTRUMENTATION The inspectors identified a Green Finding associated with the licensee's failure to perform an adequate channel calibration to determine the functionality of the stations seismic monitoring equipmen used for evaluaing earthquakes.
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| Specifically, station procedure CPS 9437.21, "Trix Time-History Accelerometer Channel Calibration," Revision 39c, did not include steps to ensure that battery backup power was provided to operate the equipment on a loss of the normal power source as part of the operability requirements. The licensee documented the issue in the corrective action program as action request (AR) 02454630. As a corrective action the licensee planned to correct procedure CPS 9437.21 to verify proper battery operation.
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| The licensee's failure to perform an adequate channel calibration to determine the functionality of the stations seismic monitoring equipment used for evaluating earthquakes was a performance deficiency. Specifically, station procedures did not include steps to ensure that battery backup power was provided to operate the equipment on a loss of the normal power source. The performance deficiency was more than minor because it adversely impacted the protection against external factors attribute of the Mitigating Systems cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual chapter 0609, Attachment 4 "Initial Characterization of Findings," and Appendix A "The Significance Determination Process for Findings at Power," issued June 19, 2012, the inspectors answered "yes" to the Mitigating Systems cornerstone question, "Does the finding involve the ... degradation of equipment ... specificallty designed to mitigate a seismic ...
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| initiating event ..." Therefore, the inspectors addressed the questions in Exhibit 4, "External Event Screening Questions." The inspectors answered "no" to the two questions in Exhibit 4. Specifically, 1) if completely failed the seismic monitor would not cause an initiating event or degrade multi-trains or risk-significant systems; and 2) the Page 5 of 13
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| 3Q/2015 Inspection Findings - Clinton finding does not involve the total loss of any safety function. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of conservative bias where individuals use decision making-practices that emphasize prudent choices over those that are simply allowable and a proposed action is determined to be a safe in order to proceed, rather than unsafe in order to stop. Specifically, the licensee documented the issue of the voltage being high out of specification and instead of performing additional corrective actions to determine if leaving the voltage out of specification was appropriate marked the step as not applicable and proceeded with the rest of the procedure. (Section 1R15)
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| Inspection Report# : 2015001 (pdf)
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| Significance: Mar 31, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation UNQUALIFIED SAFETY-RELATED CABLES USED IN A SUBMERGED ENVIRONMENT The inspectors identified a finding and an associated non-citied violation of 10 CFR 50 Appendix B, Criterion III, "Design Control," for the failure to maintain safety-related cables for the SX system in an environment for which they were designed. Specifically, the licensee failed to maintain SX safety-related cables in an environment for which they were designed when the cables were allowed to be submerged in water inside cable vaults. The licensee documented this issue in their corrective action program (CAP) as action request (AR) 02474543. Corrective actions included draining the cable vaults so that the cables were no longer submerged.
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| The licensee's failure to maintain safety-related cables for the SX system in an environment for which they were designed was a performance deficiency. The performance deficiency was determined to be more than minor because the finding was associated with the Mitigating Systems Cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licnesee failed to maintain SX safety-related cables in an environment for which they were designed when the cables were allowed to be submerged in water inside cable vaults. Using IMC 0609, "Significance Determination Process," Attachment 0609.04, "Initial Characterization of Findings," issued on June 19, 2012. Specifically, the inspectors used IMC 0609 Appendix A "SDP for Findings At-Power," issued June 19, 2012, Exhibit 2, "Mitigating Systems Screening Questions" to screen the finding. The finding screened as of very low safety significance (Green) because the inspectors answered yes to the question "does the SSC maintain its operability or functionality." Specifically, the SX system submerged cables did not cause the SX system to be inoperable or nonfunctional. The inspectors determined this finding affected the cross-cutting area of problem identification and resolution in the aspect of resolution, where the organization takes effective corrective actions to address issues in a timely manner commensurate with their safety significance. Specifically, the licensee failed to implement effective corrective actions to address an adverse trend of water in cable vaults which led to (SX) safety-related cables being submerged in water.
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| Inspection Report# : 2015001 (pdf)
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| Significance: Mar 31, 2015 Identified By: NRC Item Type: AV Apparent Violation FAILURE OF THE DIVISION 3 SHUTDOWN SERVICE WATER PUMP DUE TO AN INADEQUATE BUSHING DESIGN A self-revealed finding, preliminarily determined to be of low to moderate safety significance (White) and an associated AV of 10 CFR 50 Appendix B, Criterion III, Design Control, was identified for the failure to verify the suitability of the replacement pump design for the Division 3 Shutdown Service Water system. Spceifically, the design of the suction bell bushing for the replacement pump was inadequate to pass sufficient cooling water flow to the pump internals without being affected by mud and silt from the lake water. This finding was self-revealed on September 16, 2014, during a surveillance test to ensure operability of the Division 3 shutdown cooling water pump Page 6 of 13
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| 3Q/2015 Inspection Findings - Clinton after the pump failed to start due to a damaged bushing rendering the pump inoperable. This finding does not represent an immediate safety concern because the licensee replaced the pump in September of 2014 with a pump of similar design and provided adequate documentation that assures the pump will remain operable until a different design for the bushing that failed can be installed by June of 2016.
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| The inspectors determined that the licensee's failure to verify the suitability of the design for the Division 3 Shutdown Service Water pump was a performance deficiency warranting a significance evaluation. The inspectors determined that the finding was more than minor in accordance with IMC 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," dated September 7, 2012, because it was associated with the Mitigating Systems Cornerstone attributes of design control and equipment performance and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. A Significance and Enforcement Review Panel (SERP), using IMC 0609, Appendix A, "Significance Determination Process For Findings At-Power," dated June 19, 2012, preliminarily determined the finding to be of low to moderate safety significance (White). The performance deficiency associated with this finding did not reflect current licensee performance; therefore, no cross cutting aspect was identified with this finding.
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| Inspection Report# : 2015001 (pdf)
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| Significance: Mar 20, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation Inadequate 50.59 Evaluation for Switchgear in Seismically Unanalyzed Conditions (Section 1R17.1b.)
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| Severity Level IV Green. The inspectors identified a finding of very-low safety significance, and an associated Non-Cited Violation of Title 10, Code of Federal Regulations Part 50, Section 59, Changes, Tests and Experiments, (effective January 1, 1997) for a procedure change dated May 2, 1997, where the licensee allowed safety-related switchgear to operate for a limited period of time during plant operation in equipment configurations that were seismically unanalyzed. Specifically, for Safety Evaluation Log 97 060, CPS [Clinton Power Station]
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| Procedure No. 1014.11, Revision 0, the licensee failed to include a written safety evaluation which provided the bases that concluded for all switchgear configurations that a seismically unanalyzed condition does not involve an unreviewed safety question, and the possibility for a malfunction of a different type than any evaluated previously in the Safety Analysis Report may be created. The licensee entered the issue into their Corrective Action Program as Action Request 02471583, NRC Mod 50.59 Inspection Safety Eval 97 060 for CPS 1014.11, dated March 20, 2015.
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| The performance deficiency was determined to be more than minor because the finding was associated with the Mitigating Systems cornerstone attribute of protection against external factors and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, switchgear in a seismically unanalyzed condition when relied upon to perform a safety function did not ensure the availability, reliability, or capability of the associated Mitigating Systems to respond to an initiating event such as an earthquake. The inspectors determined that the underlying technical issue was of very-low safety significance (Green) using a detailed risk evaluation. The inspectors did not identify a cross-cutting aspect associated with the finding because the finding was not representative of current performance.
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| Inspection Report# : 2015008 (pdf)
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| Significance: Dec 31, 2014 Identified By: NRC Item Type: NCV Non-Cited Violation FAILURE TO TRANSLATE SEISMIC DESIGN REQUIREMENTS INTO APPLICABLE PROCEDURES The inspectors identified a green non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for the Page 7 of 13
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| 3Q/2015 Inspection Findings - Clinton failure to adequately translate seismic requirements from a design calculation into applicable procedures. Specifically the licensee failed to incorporate the seismic requirements for the Division III 4.16 KV switchgear as described in calculation IP-Q-0391 Seismic Qualification of 480V ABB Unit Sub Switchgears, Div I & II Westinghouse Switchgears and Div III GE 4.16KV Switchgears, into procedure CPS 1014.11 6900/4160/480V Switchgear/Circuit Breaker Operability Program, resulting in the licensee incorrectly declaring Division III switchgear operable when in a seismically unanalyzed condition. The licensee entered this issue into their corrective action program as AR 2386676.
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| The inspectors determined that the failure to adequately incorporate the seismic requirements of the design calculation into the applicable procedure was a performance deficiency. The finding is more than minor because it is associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 4, External Events Screening Questions, dated June 19, 2012, the inspectors answered Yes to question 1 of External Events screening questions, because the finding could potentially degrade one train of the emergency power system. Thus the inspectors consulted the regional senior reactor analyst (SRA).
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| Based on the Detailed Risk Evaluation, the inspectors determined that the finding was of very low safety significance (Green). The inspectors determined that there was no cross-cutting aspect associated with this finding because the cause of the performance deficiency occurred more than fifteen years ago, and was not representative of current licensee performance.
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| Inspection Report# : 2014005 (pdf)
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| Significance: Dec 31, 2014 Identified By: Self-Revealing Item Type: NCV Non-Cited Violation FAILURE TO PROVIDE PROCEDURE INSTRUCTION RESULTS IN EXCEEDING TECHNICAL SPECIFICATION HEAT UP RATE DURING PLANT START UP The inspectors are documenting a self-revealing non-cited violation of Technical Specification 5.4., Procedures, for the licensees failure to establish instructions in station procedure CPS 9059.01, Reactor Coolant System Leakage Test, Revision 9b. Specifically, the licensee failed to provide instructions to ensure that the main steam piping between the reactor vessel and the inboard main steam isolation valves were completely drained of water at the completion of testing. The licensee entered this issue into the corrective action program as action request AR 01590671.
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| The inspectors determined that the licensees failure to establish instructions to ensure that the main steam piping between the reactor vessel and the inboard main steam isolation valves were completely drained of water prior to starting up the reactor was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences, and is therefore a finding.
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| Using Manual Chapter 0609, Attachment 4 Initial Characterization of Findings, and Appendix A The Significance Determination Process for Findings at Power the finding was screened against the mitigating systems cornerstone and determined to be of very low safety significance (Green) because the finding was/did not: 1) a deficiency affecting the design or qualification of a mitigating structure, system or component, 2) represent a loss of system and/or function, 3) represent an actual loss of function of a single train for greater than its technical specification Page 8 of 13
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| 3Q/2015 Inspection Findings - Clinton allowed outage time, 4) represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant for greater than 24 hours and 5) did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding or severe weather event.
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| The inspectors determined this finding affected the cross cutting area of human performance in the aspect of work management where the organization implements a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. Specifically, the licensee failed to have a plan or provide a control method to ensure the main steam piping was drained prior to commencing reactor start up. (IMC 0301 H.5)
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| Inspection Report# : 2014005 (pdf)
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| Barrier Integrity Significance: Sep 30, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation FAILURE TO OBTAIN A LICENSE AMENDMENT PRIOR TO MAKING MODIFICATIONS TO SECONDARY CONTAINMENT The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.59(d)(1), Changes, Tests, and Experiments for the licensees failure to provide a written evaluation, which provided the basis for determining that the change to the secondary containment completed on December 18, 2014 did not require a license amendment.
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| Specifically, the licensee made a change pursuant to 10 CFR 50.59(c), to the secondary containment, and eliminated the tornado wind and tornado missile loading condition from the FB Railroad Airlock (the enclosure walls and roof) and associated outer door (1SD1-31) Seismic Category I requirements and did not provide a written evaluation to provide a basis for the determination that this change would not result in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system or component important to safety.
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| The inspectors determined that the licensees failure to provide a written evaluation, which provided the basis for determining that the change to the secondary containment completed on December 18, 2014 did not require a license amendment was a performance deficiency. Specifically, the licensee made a change pursuant to 10 CFR 50.59(c) to the secondary containment and eliminated the tornado wind and tornado missile loading condition from the FB Railroad Airlock (the enclosure walls and roof) and associated outer door and did not provide a written evaluation to provide a basis for the determination that this change would not result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety. The performance deficiency was determined to be more than minor in accordance with IMC 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," dated September 7, 2012, because it was associated with the design control attribute of the barrier integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system and containment) protect the public from radionuclide releases caused by accidents or events. In addition, the associated violation was determined to be more than minor because the inspectors could not reasonably determine if the changes to secondary containment would have required NRC prior approval. The licensee documented the issue in the CAP as action request (AR) 02534694. The licensee is complying with technical specifications anytime the inner railroad bay door is opened by entering the applicable action statements, evaluating weather conditions and impact to plant risk and establishing the necessary mitigating actions required prior to opening the door. Violations of 10 CFR 50.59 are dispositioned using the traditional enforcement process instead of the SDP because they are considered to be violations that potentially impede or impact the regulatory process. However, if possible, the underlying technical issue is evaluated under the SDP to determine the severity of the violation. In this case, the inspectors used IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings at Power, issued June 19, 2012, Page 9 of 13
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| 3Q/2015 Inspection Findings - Clinton the finding was screened against the barrier integrity cornerstone and determined to be of very low safety significance (Green) because the finding did not represent a degradation only of the radiological barrier function for the Standby Gas Treatment (SBGT) system nor did it represent a degradation of the function of the control room against smoke or toxic atmosphere. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of procedure adherence where individuals follow processes, procedures and work instructions. Specifically, the licensee failed to follow the 50.59 regulatory process as defined in station procedure LS-AA-104-1000, 50.59 Resource Manual, Revision 9.
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| Inspection Report# : 2015003 (pdf)
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| Significance: Sep 30, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation FAILURE TO ENTER APPROPRIATE TS ACTION STATEMENT FOR INOPERABLE RADIATION MONITORS DURING OPDRV ACTIVITIES The inspectors identified a green finding and associated NCV of T.S. 3.3.6.1 Primary Containment and Drywell Isolation Instrumentation and 3.3.6.2 Secondary Containment Isolation Instrumentation, for the failure to enter the appropriate action statement and take the associated actions related to inoperable containment radiation monitor instrumentation during operations with the potential to drain the reactor vessel. Specifically, with the containment ventilation dampers closed, the containment radiation monitor instrumentation would not be able to perform its safety function of sending a containment isolation signal for elevated containment radiation levels as required during OPDRVS. At the time of discovery the licensee had already concluded OPDRV activities and was therefore no longer in a mode of applicability. The licensee documented the issue in the CAP as action request (AR) 2566708. When this issue was identified the maintenance on the VR/VQ system was complete and no OPDRVs were in progress, therefor the T.S. noncompliance was no longer in effect.
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| The inspectors determined that the failure to enter T.S. 3.3.6.1 and 3.3.6.2 when the radiation monitor instrumentation was not able to perform its safety function during an OPDRV, was a performance deficiency. Specifically, the licensee failed to recognize that when the containment ventilation dampers were closed, the radiation monitors could not detect the radiation levels in primary containment and therefore could not fulfill their safety function of sending containment isolation signals in the case of elevated radiation levels in containment. The performance deficiency was more than minor in accordance with IMC 0612, Power Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, because, it was associated with the SSC and Barrier Performance attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events, and is therefore a finding.
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| Specifically, the automatic containment isolation signal function of the radiation monitors was impacted when the containment ventilation dampers were closed during OPDRV operations. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings, dated May 9, 2014, the finding was screened against the Barrier Integrity cornerstone and determined to need a detailed risk evaluation because the finding represents a degradation of the ability to close or isolate the containment. Using Appendix G Exhibit 4, Barrier Integrity Screening Questions, the Senior Reactor Analyst (SRA) determined that the finding degraded the ability to close or isolate the containment per Section B, Containment Barrier, Question 6. Therefore, the evaluation was continued using IMC 0609 Appendix H, Containment Integrity Significance Determination Process. The SRA determined this to be a Type B finding, because it was related to a degraded condition that had implications for containment integrity without affecting the likelihood of core damage. The SRA used Section 6.2 of Appendix H, Approach for Assessing Type B Findings at Shutdown. Based on information from the inspectors, during all OPDRV time windows, the reactor water level was confirmed to be greater than the minimum level required for movement of irradiated fuel assemblies (i.e., greater than 228 above the flange). This plant condition meets the definition of Plant Operating State 3 (POS 3) of Appendix H. Therefore, based on the plant being in POS 3 during Page 10 of 13
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| 3Q/2015 Inspection Findings - Clinton the OPDRV time windows, the finding screens as Green based on Step 2.1 of Section 6.2 of Appendix H. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of conservative bias where individuals use decision making practices that emphasize prudent choices over those that are simple allowable. A proposed action is determined to be safe in order to proceed, rather than unsafe in order to stop.
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| Specifically, the licensee relied solely on the successful completion of the surveillance requirements to determine the radiation monitor instrumentation was operable rather than considering the impact the closed dampers would have on their ability to fulfill their safety function.
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| Inspection Report# : 2015003 (pdf)
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| Significance: Dec 31, 2014 Identified By: NRC Item Type: NCV Non-Cited Violation FAILURE TO UPDATE THE UPDATED SAFETY ANALYSIS REPORT - 1VR08C FUNCTION The inspectors identified a Severity Level IV non-cited violation of Title 10 Code of Federal Regulations (CFR) 50.71 (e), Periodic Update of the USAR and an associated Green finding for the licensees failure to update the USAR with the correct description of the function of 1VR08C. Specifically the licensee did not update Section 9.4.5.5 of the USAR to include the correct function of 1VR08C as described in a commitment made to the NRC in letter U-600850.
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| Consequently the licensee performed a 50.59 evaluation for abandoning a portion of the system that did not consider the correct function of the component. The licensee entered this issue into their corrective action program as AR 1692665.
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| The inspectors determined that the failure to update the USAR with the correct function of 1VR08C was a performance deficiency. The performance deficiency was determined to be more than minor in accordance with Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, because, if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern and is therefore a finding. Specifically, failure to update the USAR with the correct safety related function of VR08C could result in the licensee making operability and functionality determinations based on incorrect assumptions. Additionally, the failure to update the USAR with the correct function of the fan was more than minor because it was associated with the Barrier Integrity Cornerstone attribute of design control, plant modifications and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events.
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| Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings at Power, the finding was screened against the Barrier Integrity cornerstone and determined to be of very low safety significance (Green) because the finding does not represent an actual open pathway in the physical integrity of reactor containment, containment isolation system, and heat removal components and it did not involve an actual reduction in function of hydrogen igniters in reactor containment. The performance deficiency associated with this finding did not reflect current licensee performance; therefore, no cross cutting aspect was identified with this finding.
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| Additionally, in accordance with Section 6.1.d.3 of the NRC Enforcement Policy, this violation was categorized as Severity Level IV because the licensees failure to update the USAR as required by 10 CFR 50.71(e) had not yet resulted in any unacceptable change to the facility or procedures.
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| Inspection Report# : 2014005 (pdf)
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| Page 11 of 13
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| 3Q/2015 Inspection Findings - Clinton Emergency Preparedness Occupational Radiation Safety Significance: Jun 30, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation CONTRACT WORKERS NOT MONITORED FOR OCCUPATIONAL RADIATION EXPOSURE The inspectors identified a finding of very-low safety significance and an associated NCV of Technical Specification (TS) 5.4.1, Procedures, for the failure to monitor the radiation dose received by a group of workers as required by station procedure RP-AA-210, Dosimetry Issue, Usage, and Control. Specifically, contractor employees who did not wear individual dosimetry were not monitored by the usage of an Area Badging Program and the workers were not excluded from wearing individual dosimetry by the usage of medical isotopes or external radioactivity being detected, or a previously performed evaluation by RP Supervision. The licensee documented the issue in the licensees CAP as action request AR 02452005. The trailer was relocated to a distance further away from the radioactive material storage area. This reduced the radiation dose rate in the trailer.
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| The inspectors determined that the issue of concern was a performance deficiency because the licensee did not monitor a group of workers using one or more methods as required by procedure, RP-AA-210, Dosimetry Issue, Usage and Control. The licensee did not assign radiation dosimetry to each worker, nor was an Area Badging Program in place. The inspectors determined that the cause of the performance deficiency was reasonably within the licensees ability to foresee and correct and should have been prevented. The issue was not subject to traditional enforcement since the concern did not have a significant safety consequence, did not impact the NRCs ability to perform its regulatory function, and was not willful. The performance deficiency was determined to be of more than minor safety significance in accordance with IMC 0612, Appendix B, Issue Screening, issued September 7, 2012, because it was associated with the program and process attribute of the Occupational Radiation Safety Cornerstone, and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation. Specifically, the licensee could not demonstrate compliance with other sections of 10 CFR Part 20, such as occupational dose limits, and records and reporting of individual monitoring results. The inspectors also reviewed the guidance in IMC 0612, Appendix E, Examples of Minor Issues, and did not find any similar examples.
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| In accordance with IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, issued August 19, 2008, the inspectors determined that the finding had very low safety significance (Green) because the finding: (1) did not involve as-low-as-reasonably-achievable planning and controls; (2) did not involve a radiological overexposure; (3) there was not a substantial potential for an overexposure; and (4) there was no compromised ability to assess dose. This finding has a cross-cutting aspect in the area of Human Performance, Change Management, because the primary cause of the finding was due to inadequate change management.
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| Specifically, licensee supervision incorrectly located the trailer near a posted radiation area without performing an appropriate evaluation to ensure the personnel or area was correctly monitored. [H.3]
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| Inspection Report# : 2015002 (pdf)
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| Public Radiation Safety Page 12 of 13
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| 3Q/2015 Inspection Findings - Clinton Security Although the Security Cornerstone is included in the Reactor Oversight Process assessment program, the Commission has decided that specific information related to findings and performance indicators pertaining to the Security Cornerstone will not be publicly available to ensure that security information is not provided to a possible adversary.
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| Other than the fact that a finding or performance indicator is Green or Greater-Than-Green, security related information will not be displayed on the public web page. Therefore, the cover letters to security inspection reports may be viewed.
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| Miscellaneous Last modified : December 15, 2015 Page 13 of 13
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| 1Q/2016 Inspection Findings - Clinton Clinton 1Q/2016 Plant Inspection Findings Initiating Events Significance: Dec 31, 2015 Identified By: NRC Item Type: FIN Finding Failure to Follow Station Procedures for Plant Activities The inspectors identified a finding of very low safety significance for the failure to ensure that activities were accomplished in accordance with prescribed procedures as required by station procedure HU-AA-104-101 Procedure Use and Adherence. Specifically, the inspectors identified two examples where the licensee failed to adhere to prescribed station procedures when performing activities in the plant. The licensee placed both issues in their corrective action program as AR 02600726 and addressed the nonconformances created by the failure to follow the procedures. The licensee planned to perform an apparent cause evaluation to determine why there was an adverse trend related to procedure adherence.
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| The inspectors determined that the failure to perform activities in accordance with prescribed procedures as required by station procedure HU-AA-104-101, Procedure Use and Adherence, was a performance deficiency. Specifically, the inspectors identified two instances where the licensee failed to follow procedures when performing activities in the plant. The performance deficiency was more than minor in accordance with IMC 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," dated September 7, 2012, because, if left uncorrected the performance deficiency had the potential to lead to a more significant safety concern. Specifically, by not performing activities in accordance with a procedure the licensee could manipulate equipment and challenge the operators, and cause unexpected transients. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process (SDP) for Findings at Power, issued June 19, 2012, the finding was screened against the Initiating Events cornerstone and determined to be of very low safety significance because the finding did not cause a reactor trip or the loss of mitigation equipment and it did not involve the complete or partial loss of a support system that contributes to the likelihood of, or cause, an initiating event. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of challenging the unknown which stated, individuals stop when faced with uncertain conditions. Risks are evaluated and managed before proceeding. Contrary to this, when challenged with unknown conditions, the licensee did not stop and properly evaluate the issues before proceeding, resulting in adverse impacts to station equipment. (H.11)
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| Inspection Report# : 2015004 (pdf)
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| Significance: Sep 30, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation FAILURE TO FOLLOW PROCEDURE LEAVES CONTROL ROOM CABINET DOORS UNATTENDED IN SEISMICALLY UNANALYSED CONDITION The inspectors identified a finding of very low safety significance (Green) and an associated NCV of 10 CFR50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to maintain control room doors in a seismically analyzed condition, in accordance with station procedure CPS 1014.11, 6900/4160/480v Switchgear/Circuit Breaker Operability Program, Revision 5a. Specifically, on several occasions the licensee failed Page 1 of 11
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| 1Q/2016 Inspection Findings - Clinton to maintain control room cabinet doors in seismically qualified positions, while performing maintenance or trouble shooting activities, by leaving the doors open and unattended. The licensee documented the issue in the Corrective Action Program (CAP) as action request (AR) 02518477. The licensee has revised the station procedure to ensure control room cabinet doors either remain latched closed or are completely removed when unattended and has issued a standing order to ensure the requirements are reinforced.
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| The performance deficiency was more than minor in accordance with IMC 0612, "Power Reactor Inspection Reports,"
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| Appendix B, "Issue Screening," dated September 7, 2012, because, it was associated with the configuration control performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations and is therefore a finding. Specifically, leaving the control doors in a seismically unanalyzed condition could challenge critical safety functions during a seismic event. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process (SDP) for Findings at Power, issued June 19, 2012, the finding was screened against the Initiating Events cornerstone and determined to be of very low safety significance (Green) because the finding did not result in exceeding the reactor coolant system leak rate for a small loss of coolant accident (LOCA), cause a reactor trip, involve the complete or partial loss of a support system that contributes to the likelihood of, or caused, an initiating event and did not affect mitigation equipment. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of resources where leaders ensure that personnel, equipment, procedures and other resources are available and adequate to support nuclear safety. Specifically, the licensee failed to ensure the personnel performing maintenance and troubleshooting had adequate documentation in written work instructions to maintain control room cabinets in seismically analyzed conditions.
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| Inspection Report# : 2015003 (pdf)
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| Significance: Jun 30, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation POST MAINTENANCE TEST FAILED TO DEMONSTRATE REQUIRED FLOW THROUGH RCIC ROOM COOLER A self-revealed finding of very low safety significance and an associated Non-Cited Violation of 10 CFR 50 Appendix B, Criterion XI, Test Control, was documented by the inspectors for the failure to perform adequate post maintenance testing that would assure that the Reactor Core Isolation Cooling (RCIC) room cooler would perform its intended function when restored to service following maintenance. Specifically, the licensee declared the room cooler operable with insufficient cooling flow through the cooler. The licensee documented the issue in the licensees corrective action program (CAP) as action request (AR) 02447013. The licensee operated the RCIC Room Cooler outlet valve from its throttled position to fully open to flush the seat and the upstream piping and positioned the valve to maintain the required flow to restore the cooler to an operable condition.
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| The failure to perform adequate post maintenance testing that would assure that the RCIC Room cooler would perform its intended function when restored to service following maintenance is a performance deficiency. The performance deficiency was determined to be more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, because it was associated with the procedure quality attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences and is, therefore, a finding. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings at Power, dated June 19, 2012, the finding was screened against the Mitigating Systems cornerstone and determined to need a detailed risk evaluation because the finding represents the loss of a system and/or function. The Region III Senior Reactor Analysts (SRAs) evaluated the finding using the Clinton Station Standardized Plant Analysis Risk (SPAR) Model Version 8.17, Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) Version 8.1.2. The SRAs reviewed the Page 2 of 11
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| 1Q/2016 Inspection Findings - Clinton licensees Apparent Cause Investigation Report IR 2447013. The exposure time was assumed to be 150.5 hours based on information in that report. The SRAs modeled the condition using failure of the RCIC pump as a surrogate for failure of the RCIC room cooler. The basic event representing the RCIC pump failure-to-run was set to True for the 150.5 hour duration. The result was a ?CDF of 9.98E-08/yr. The dominant sequence was a station blackout initiating event; failure of high pressure core spray; failure of reactor core isolation cooling; and failure to recover offsite or emergency AC power within 30-minutes. Based on the detailed risk evaluation, the finding is best characterized as a finding of very low safety-significance (Green). The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of work management where the organization implements a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. The work process includes the identification and management of risk commensurate to the work and the need for coordination with different groups of job activities. Specifically, the licensee failed to plan and execute adequate post maintenance testing that would have ensured the satisfactory operation of the RCIC Room cooler following planned maintenance.
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| [H.5]
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| Inspection Report# : 2015002 (pdf)
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| Significance: Jun 30, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation FAILURE TO TRANSLATE SUFFICIENT GLAND STRESS TO PACKING GLAND NUTS RESULTED IN VALVE PACKING FAILURE AND PLANT SHUTDOWN A finding of very low safety significance and an associated Non-Cited Violation of 10 CFR50, Appendix B, Criterion III, Design Control, was self-revealed on January 19, 2015, when a steam leak developed from the RCIC system inboard steam isolation valve (1E51F0063) stem packing. Specifically, the licensee failed to identify and implement a torque value for the gland packing nuts for the RCIC system inboard steam isolation valve 1E51F0063 to overcome service induced consolidation and prevent packing leakage. This resulted in a plant down power to 83 percent and subsequent plant shutdown due to increasing unidentified reactor coolant system leakage. The licensee documented the issue in the licensees CAP as AR 02439437. The licensee repacked the valve utilizing the station procedure CPS 8120.37, Valve Packing Installation, and applicable SealPro data sheet. A four ring set of A.P. Services graphite packing was installed with a new live load assembly sized to a new torque value of 59 ft-lbs. and the valve packing was tested to verify no leakage.
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| The inspectors determined that the failure to apply sufficient packing gland torque to overcome service induced consolidation and prevent packing leakage on the RCIC system inboard steam isolation valve was a performance deficiency. The performance deficiency was more than minor in accordance with IMC 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," dated September 7, 2012, because, it was associated with the equipment performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations and is therefore a finding. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings at Power, issued June 19, 2012, the finding was screened against the Initiating Events cornerstone and determined to be of very low safety significance (Green) because the finding did not result in exceeding the RCS leak rate for a small LOCA, cause a reactor trip, involve the complete or partial loss of a support system that contributes to the likelihood of, or caused, an initiating event and did not affect mitigation equipment. The inspectors determined that no cross-cutting aspect would be associated with this finding since the performance deficiency occurred in 2010 and was not representative of current licensee performance in the of valve packing.
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| Inspection Report# : 2015002 (pdf)
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| Significance: Jun 30, 2015 Page 3 of 11
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| 1Q/2016 Inspection Findings - Clinton Identified By: NRC Item Type: FIN Finding FAILURE TO EVALUATE THE OPERATIONAL IMPACT OF THE TDRFP LOCKOUT SWITCH POSITION A self-revealed finding was identified for failure to evaluate the consequences of an adverse condition, in accordance with the operational decision making process. Specifically, contrary to station procedure OP-AA-106-101-1006 Operational Decision Making Process, Revision 14, the licensee failed to adequately implement the procedure to ensure the consequences of leaving the switch in the lockout position were evaluated, which resulted in the loss of the manual trip function for the A turbine driven reactor feed pump (TDRFP). The licensee documented the issue in the licensees CAP as action request (AR) 02440052. The licensee repaired the ground condition and returned the switch to its normal position. The licensee also revised the surveillance procedure to document the limitations associated with putting the emergency governor trip test and lockout switch in the lockout position.
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| The inspectors determined that the failure to adequately implement the procedure to ensure the consequences of leaving the switch in the lockout position were evaluated, which resulted in the loss of the manual trip function for the A TDRFP, was a performance deficiency. Specifically, by not evaluating leaving the emergency governor trip test and lockout switch in the lockout position, the licensee lost the ability to manually trip the A TDRFP, which challenged the operators during the reactor shutdown, and nearly resulted in a Level 8 reactor SCRAM. The performance deficiency was more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, because if left uncorrected the performance deficiency had the potential to lead to a more significant safety concern. The performance deficiency was also associated with the configuration control attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown and power operations and is therefore a finding. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings at Power, issued June 19, 2012, the finding was screened against the Initiating Events cornerstone and determined to be of very low safety significance (Green) because the finding did not cause a reactor trip or the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of resources where leaders ensure that personnel, equipment, procedures, and other resources are available and adequate to support nuclear safety. Specifically, the surveillance procedure for the RFPT emergency governor and trip mechanism test Section 2.1.1 stated if an actual signal was generated during testing, the lockout valve would de-energize to allow the trip mechanism to operate and trip the RFPT, which led to the understanding that the trip functions were unaffected by the switch position. (H.1)
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| Inspection Report# : 2015002 (pdf)
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| Mitigating Systems Significance: Feb 04, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Perform and Adequate Equipment Apparent Cause Evaluation (Section 4OA4)
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| The inspectors identified a finding of very-low safety significance (Green), and an associated Non-Cited Violation of Title 10, Code of Federal Regulations, Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to follow Step 4.3.4 of procedure PI-AA-125, Corrective Action Program Procedure.
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| Specifically, the licensee failed to perform Class B Equipment Apparent Cause Evaluation (EACE) 2381871, 1SX01PC Failed to Start for Testing, in accordance with PI-AA-125-1003, Apparent Cause Evaluation Manual, because they: (1) failed to analyze each causal factor to determine contributing causes as required by Step 4.4.1.2; and Page 4 of 11
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| 1Q/2016 Inspection Findings - Clinton (2) failed to assign an effectiveness review for the EACE as required by Step 4.4.9.1. The licensee entered this finding into their Corrective Action Program and revised their EACE to: (1) include three contributing causes; (2) upgrade a corrective action to a corrective action to prevent recurrence; and (3) assign an effectiveness review to determine the effectiveness of the corrective action to prevent recurrence.
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| The performance deficiency was determined to be more than minor because if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, an effectiveness review is required to provide assurance that the Division 3 SX pump design change is successful in preventing recurrence of pump failure before another pump failure occurs, which would be a more significant safety concern. The finding impacted the Mitigating Systems Cornerstone and screened as having very-low safety significance (Green) because although the finding is a deficiency ultimately affecting the design or qualification of the Division 3 SX pump, the pump still maintains its operability. The inspectors determined this finding had an associated cross-cutting aspect in the area of Human Performance (Conservative Bias) because although a B Apparent Cause Evaluation may have been allowable for investigating the failure of the Division 3 SX pump, had an A Root Cause Analysis been performed, a more rigorous investigation process would have been used to identify contributing causes, assign corrective actions, and identify effectiveness reviews for the failure of the Division 3 SX pump. [H.14] (Section 4OA4.02.03.f)
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| Inspection Report# : 2016008 (pdf)
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| Significance: Dec 31, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Perform Activities Affecting Quality in Accordance with Prescribed Procedures The inspectors identified a finding of very low safety significance and an associated Non-Cited Violation of 10 CFR50, Appendix B, Criterion V, Instructions Procedures and Drawings, for the failure to ensure that activities affecting quality were accomplished in accordance with the appropriate instructions, procedures and drawings.
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| Specifically, the inspectors identified two examples where the licensee failed to perform activities affecting quality in accordance with prescribed procedures. The licensee entered this issue into their corrective action program as action request (AR) 02600726 and planned to perform an apparent cause evaluation to address the trend. Separate action requests were also written and immediate corrective actions were taken for each identified example to address the nonconformances created by the failure to follow procedures.
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| The inspectors determined that the failure to ensure that activities affecting quality were accomplished in accordance with the appropriate instructions, procedures and drawings as required by 10 CFR 50 Appendix B Criterion V, was a performance deficiency. Specifically, the inspectors identified two instances where the licensee failed to follow procedures resulting in impacts to safety related equipment and processes. The performance deficiency was more than minor in accordance with Inspection Manual Chapter (IMC) 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," dated September 7, 2012, because, if left uncorrected the performance deficiency had the potential to lead to a more significant safety concern. Specifically, by not performing activities affecting quality in accordance with a procedure the licensee could manipulate equipment and challenge the operators by causing unexpected transients or impact safety related equipment. Using IMC 0609, Appendix G, Shutdown Operations Significance Determination Process, Attachment 1, issued May 9, 2014, the finding was screened against the Mitigating Systems cornerstone and determined to be of very low safety significance because the finding did not represent a loss of system safety function, it did not represent an actual loss of function of a single train or two separate trains for greater than its allowed outage time, it did not represent an actual loss of safety function of one or more non-TS trains of equipment during shutdown for equipment designated as risk significant for greater than 24 hours, and it did not degrade a functional auto-isolation of residual heat removal (RHR) on low reactor vessel level. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of challenging the unknown which states, individuals stop when faced with uncertain conditions. Risks are evaluated and managed before proceeding. Contrary to this, when challenged with uncertain conditions, the licensee did not stop and properly Page 5 of 11
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| 1Q/2016 Inspection Findings - Clinton evaluate the issues before proceeding, resulting in adverse impacts to safety related equipment and activities. (H.11)
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| Inspection Report# : 2015004 (pdf)
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| Significance: Oct 09, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Generate Issue Reports for Conditions Adverse to Quality The inspectors identified a finding of very low safety significance, and an associated NCV of Title 10, Code of Federal Regulations, Part 50, Appendix B, Criterion II, Quality Assurance Program, for the failure to perform activities in accordance with procedure PI-AA-125, Corrective Action Program, Revision 2, which was a Quality Assurance Program implementing procedure. Specifically, the inspectors identified six examples where the licensee failed to generate IRs for conditions adverse to quality (CAQ) as required by PI-AA-125, until prompted by the inspectors. The licensee documented the issue in the CAP as IR 2518477, and planned on reviewing the apparent cause evaluation to determine if additional actions needed to be taken.
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| The performance deficiency was more than minor in accordance with IMC 0612, "Power Reactor Inspection Reports,"
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| Appendix B, "Issue Screening," dated September 7, 2012, because, if left uncorrected the performance deficiency had the potential to lead to a more significant safety concern. Specifically, by not identifying and documenting conditions adverse to quality the issues would not go through the screening and review process in accordance with the corrective action procedure, which could impact the identification of conditions affecting operability. The finding was screened against the Mitigating Systems cornerstone, and determined to be of very low safety significance because the it did not represent a loss of safety system or function, it did not represent an actual loss of function of a single train of two separate trains for greater than its allowed outage time and it did not represent a loss of function of a non-technical specification system designated as highly safety-significant within the licensees Maintenance Rule Program for greater than 24 hours. The inspectors determined this finding affected the cross-cutting area of problem identification and resolution in the aspect of identification where the organization implements a CAP with a threshold for identifying issues and individuals identify issues completely, accurately and in a timely manner in accordance with the program. Specifically, the licensee failed to identify issues completely, accurately and in a timely manner, causing them to not recognize issues as CAQs, and therefore not follow their process for handling these issues.
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| Inspection Report# : 2015007 (pdf)
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| Significance: Sep 30, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation FAILURE TO IMPLEMENT AND COMPLY WITH TRANSIENT EQUIPMENT/MATERIALS PROGRAM The inspectors identified a green finding and an associated NCV of 10 CFR 50, Appendix B, Criterion V Instructions, Procedures, and Drawings for the licensees failure to implement and comply with station procedure CPS 1019.05, Transient Equipment/Materials, Revision 23, to ensure that transient equipment and materials are controlled so there is no impact to safe operation of plant equipment. Specifically, on numerous occasions the inspectors identified equipment and materials improperly staged, improperly secured or in areas without engineering evaluations. The licensee documented the issue in the CAP as action requests (AR) 02507167 and AR 02529227. In each occasion identified by the inspectors the licensee subsequently removed the items identified to restore compliance with the station procedures.
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| The inspectors determined the licensees failure to implement and comply with station procedures to ensure that transient equipment and materials are controlled so there is no impact to safe operation of plant equipment was a performance deficiency. The performance deficiency was determined to be more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Screening, dated Page 6 of 11
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| 1Q/2016 Inspection Findings - Clinton September 7, 2012, because if left uncorrected it had the potential to lead to a more significant safety concern.
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| Specifically, transient equipment and material in proximity of safety related components has the potential of impacting these components during a seismic event, potentially rendering them unable to fulfill their safety function.
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| The performance deficiency is also associated with the protection against external factors attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability and capability of systems that response to initiating events to prevent undesirable consequences, and is therefore a finding.
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| Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process (SDP) for Findings at Power, issued June 19, 2012, the finding was screened against the Mitigating Systems cornerstone and determined to be of very low safety significance (Green) because the finding did not represent a loss of system or function, it did not represent an actual loss of function of at least a single train for >
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| its TS allowed outage time and it did not represent an actual loss of one or more not TS trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of field presence where leaders are commonly seen in the work areas of the plant observing, coaching, reinforcing standards and expectation. Deviations from standards and expectations are corrected promptly. Specifically, after various examples of material placement being an issue, the licensee didnt perform in field observations, caching and reinforcement of standards and expectations in the identified areas.
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| Inspection Report# : 2015003 (pdf)
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| Barrier Integrity Significance: Sep 30, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation FAILURE TO OBTAIN A LICENSE AMENDMENT PRIOR TO MAKING MODIFICATIONS TO SECONDARY CONTAINMENT The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.59(d)(1), Changes, Tests, and Experiments for the licensees failure to provide a written evaluation, which provided the basis for determining that the change to the secondary containment completed on December 18, 2014 did not require a license amendment.
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| Specifically, the licensee made a change pursuant to 10 CFR 50.59(c), to the secondary containment, and eliminated the tornado wind and tornado missile loading condition from the FB Railroad Airlock (the enclosure walls and roof) and associated outer door (1SD1-31) Seismic Category I requirements and did not provide a written evaluation to provide a basis for the determination that this change would not result in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system or component important to safety.
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| The inspectors determined that the licensees failure to provide a written evaluation, which provided the basis for determining that the change to the secondary containment completed on December 18, 2014 did not require a license amendment was a performance deficiency. Specifically, the licensee made a change pursuant to 10 CFR 50.59(c) to the secondary containment and eliminated the tornado wind and tornado missile loading condition from the FB Railroad Airlock (the enclosure walls and roof) and associated outer door and did not provide a written evaluation to provide a basis for the determination that this change would not result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety. The performance deficiency was determined to be more than minor in accordance with IMC 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," dated September 7, 2012, because it was associated with the design control attribute of the barrier integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system and containment) protect the public from radionuclide releases Page 7 of 11
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| 1Q/2016 Inspection Findings - Clinton caused by accidents or events. In addition, the associated violation was determined to be more than minor because the inspectors could not reasonably determine if the changes to secondary containment would have required NRC prior approval. The licensee documented the issue in the CAP as action request (AR) 02534694. The licensee is complying with technical specifications anytime the inner railroad bay door is opened by entering the applicable action statements, evaluating weather conditions and impact to plant risk and establishing the necessary mitigating actions required prior to opening the door. Violations of 10 CFR 50.59 are dispositioned using the traditional enforcement process instead of the SDP because they are considered to be violations that potentially impede or impact the regulatory process. However, if possible, the underlying technical issue is evaluated under the SDP to determine the severity of the violation. In this case, the inspectors used IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings at Power, issued June 19, 2012, the finding was screened against the barrier integrity cornerstone and determined to be of very low safety significance (Green) because the finding did not represent a degradation only of the radiological barrier function for the Standby Gas Treatment (SBGT) system nor did it represent a degradation of the function of the control room against smoke or toxic atmosphere. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of procedure adherence where individuals follow processes, procedures and work instructions. Specifically, the licensee failed to follow the 50.59 regulatory process as defined in station procedure LS-AA-104-1000, 50.59 Resource Manual, Revision 9.
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| Inspection Report# : 2015003 (pdf)
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| Significance: Sep 30, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation FAILURE TO ENTER APPROPRIATE TS ACTION STATEMENT FOR INOPERABLE RADIATION MONITORS DURING OPDRV ACTIVITIES The inspectors identified a green finding and associated NCV of T.S. 3.3.6.1 Primary Containment and Drywell Isolation Instrumentation and 3.3.6.2 Secondary Containment Isolation Instrumentation, for the failure to enter the appropriate action statement and take the associated actions related to inoperable containment radiation monitor instrumentation during operations with the potential to drain the reactor vessel. Specifically, with the containment ventilation dampers closed, the containment radiation monitor instrumentation would not be able to perform its safety function of sending a containment isolation signal for elevated containment radiation levels as required during OPDRVS. At the time of discovery the licensee had already concluded OPDRV activities and was therefore no longer in a mode of applicability. The licensee documented the issue in the CAP as action request (AR) 2566708. When this issue was identified the maintenance on the VR/VQ system was complete and no OPDRVs were in progress, therefor the T.S. noncompliance was no longer in effect.
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| The inspectors determined that the failure to enter T.S. 3.3.6.1 and 3.3.6.2 when the radiation monitor instrumentation was not able to perform its safety function during an OPDRV, was a performance deficiency. Specifically, the licensee failed to recognize that when the containment ventilation dampers were closed, the radiation monitors could not detect the radiation levels in primary containment and therefore could not fulfill their safety function of sending containment isolation signals in the case of elevated radiation levels in containment. The performance deficiency was more than minor in accordance with IMC 0612, Power Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, because, it was associated with the SSC and Barrier Performance attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events, and is therefore a finding.
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| Specifically, the automatic containment isolation signal function of the radiation monitors was impacted when the containment ventilation dampers were closed during OPDRV operations. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings, dated May 9, 2014, the finding was screened against the Barrier Integrity cornerstone and determined to need a detailed risk evaluation because the finding Page 8 of 11
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| 1Q/2016 Inspection Findings - Clinton represents a degradation of the ability to close or isolate the containment. Using Appendix G Exhibit 4, Barrier Integrity Screening Questions, the Senior Reactor Analyst (SRA) determined that the finding degraded the ability to close or isolate the containment per Section B, Containment Barrier, Question 6. Therefore, the evaluation was continued using IMC 0609 Appendix H, Containment Integrity Significance Determination Process. The SRA determined this to be a Type B finding, because it was related to a degraded condition that had implications for containment integrity without affecting the likelihood of core damage. The SRA used Section 6.2 of Appendix H, Approach for Assessing Type B Findings at Shutdown. Based on information from the inspectors, during all OPDRV time windows, the reactor water level was confirmed to be greater than the minimum level required for movement of irradiated fuel assemblies (i.e., greater than 228 above the flange). This plant condition meets the definition of Plant Operating State 3 (POS 3) of Appendix H. Therefore, based on the plant being in POS 3 during the OPDRV time windows, the finding screens as Green based on Step 2.1 of Section 6.2 of Appendix H. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of conservative bias where individuals use decision making practices that emphasize prudent choices over those that are simple allowable. A proposed action is determined to be safe in order to proceed, rather than unsafe in order to stop.
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| Specifically, the licensee relied solely on the successful completion of the surveillance requirements to determine the radiation monitor instrumentation was operable rather than considering the impact the closed dampers would have on their ability to fulfill their safety function.
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| Inspection Report# : 2015003 (pdf)
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| Emergency Preparedness Occupational Radiation Safety Significance: Jun 30, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation CONTRACT WORKERS NOT MONITORED FOR OCCUPATIONAL RADIATION EXPOSURE The inspectors identified a finding of very-low safety significance and an associated NCV of Technical Specification (TS) 5.4.1, Procedures, for the failure to monitor the radiation dose received by a group of workers as required by station procedure RP-AA-210, Dosimetry Issue, Usage, and Control. Specifically, contractor employees who did not wear individual dosimetry were not monitored by the usage of an Area Badging Program and the workers were not excluded from wearing individual dosimetry by the usage of medical isotopes or external radioactivity being detected, or a previously performed evaluation by RP Supervision. The licensee documented the issue in the licensees CAP as action request AR 02452005. The trailer was relocated to a distance further away from the radioactive material storage area. This reduced the radiation dose rate in the trailer.
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| The inspectors determined that the issue of concern was a performance deficiency because the licensee did not monitor a group of workers using one or more methods as required by procedure, RP-AA-210, Dosimetry Issue, Usage and Control. The licensee did not assign radiation dosimetry to each worker, nor was an Area Badging Program in place. The inspectors determined that the cause of the performance deficiency was reasonably within the licensees ability to foresee and correct and should have been prevented. The issue was not subject to traditional enforcement since the concern did not have a significant safety consequence, did not impact the NRCs ability to perform its regulatory function, and was not willful. The performance deficiency was determined to be of more than minor safety significance in accordance with IMC 0612, Appendix B, Issue Screening, issued September 7, 2012, because it was associated with the program and process attribute of the Occupational Radiation Safety Cornerstone, Page 9 of 11
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| 1Q/2016 Inspection Findings - Clinton and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation. Specifically, the licensee could not demonstrate compliance with other sections of 10 CFR Part 20, such as occupational dose limits, and records and reporting of individual monitoring results. The inspectors also reviewed the guidance in IMC 0612, Appendix E, Examples of Minor Issues, and did not find any similar examples.
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| In accordance with IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, issued August 19, 2008, the inspectors determined that the finding had very low safety significance (Green) because the finding: (1) did not involve as-low-as-reasonably-achievable planning and controls; (2) did not involve a radiological overexposure; (3) there was not a substantial potential for an overexposure; and (4) there was no compromised ability to assess dose. This finding has a cross-cutting aspect in the area of Human Performance, Change Management, because the primary cause of the finding was due to inadequate change management.
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| Specifically, licensee supervision incorrectly located the trailer near a posted radiation area without performing an appropriate evaluation to ensure the personnel or area was correctly monitored. [H.3]
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| Inspection Report# : 2015002 (pdf)
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| Public Radiation Safety Security Although the Security Cornerstone is included in the Reactor Oversight Process assessment program, the Commission has decided that specific information related to findings and performance indicators pertaining to the Security Cornerstone will not be publicly available to ensure that security information is not provided to a possible adversary.
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| Other than the fact that a finding or performance indicator is Green or Greater-Than-Green, security related information will not be displayed on the public web page. Therefore, the cover letters to security inspection reports may be viewed.
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| Miscellaneous Significance: N/A Dec 31, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation FAILURE TO UPDATE THE FINAL SAFETY ANALYSIS REPORT (FSAR) - HYDROGEN WATER CHEMISTRY SYSTEM The inspectors identified a Severity Level IV Violation of Title 10 Code of Federal Regulations (CFR) 50.71(e),
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| "Periodic Update of the FSAR", for the licensee's failure to update the FSAR after installing a hydrogen water chemistry system into the plant to reduce rates of intergranular stress corrosion cracking (IGSCC) in recirculation piping and reactor vessel internals. Specifically, the licensee did not update Section 5.4.15, "Hydrogen Water Chemistry System" of the FSAR to include a design basis and description of process and system used to periodically injection noble metals. The licensee entered this issue into the corrective action program as AR 02594259 and is revising the FSAR include additional the design basis and additional system description for noble metal injection.
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| The inspectors determined that the failure to update the FSAR in accordance with 10 CFR 50.71(e), "Periodic Update of the FSAR", with the design basis and description of the process and system used to periodically injection noble Page 10 of 11
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| 1Q/2016 Inspection Findings - Clinton metals was a performance deficiency warranting a significance evaluation. The inspectors reviewed this issue in accrodance with NRC inspection manual chapter 0612 and the NRC enforcement manual. Violations of 10 CFR 50.71 (e) are dispositioned using the traditional enforcement process because they are considered to be violations that potentially impede or impact the regulatory process. The inspectors reviewed section 6.1.d.3 of the NRC Enforcement Policy and determined this violation was Severity :Level IV because the licensee's failure to update the FSAR as required by 10 CFR 50.71(e) had not yet resulted in any unacceptable change to the facility or procedures. No cross cutting aspect was assigned because cross cutting aspects are not assigned to traditional enforcement only violations.
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| Inspection Report# : 2015004 (pdf)
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| Last modified : April 05, 2016 Page 11 of 11
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| 1Q/2016 Inspection Findings - Clinton Clinton 1Q/2016 Plant Inspection Findings Initiating Events Significance: Dec 31, 2015 Identified By: NRC Item Type: FIN Finding Failure to Follow Station Procedures for Plant Activities The inspectors identified a finding of very low safety significance for the failure to ensure that activities were accomplished in accordance with prescribed procedures as required by station procedure HU-AA-104-101 Procedure Use and Adherence. Specifically, the inspectors identified two examples where the licensee failed to adhere to prescribed station procedures when performing activities in the plant. The licensee placed both issues in their corrective action program as AR 02600726 and addressed the nonconformances created by the failure to follow the procedures. The licensee planned to perform an apparent cause evaluation to determine why there was an adverse trend related to procedure adherence.
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| The inspectors determined that the failure to perform activities in accordance with prescribed procedures as required by station procedure HU-AA-104-101, Procedure Use and Adherence, was a performance deficiency. Specifically, the inspectors identified two instances where the licensee failed to follow procedures when performing activities in the plant. The performance deficiency was more than minor in accordance with IMC 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," dated September 7, 2012, because, if left uncorrected the performance deficiency had the potential to lead to a more significant safety concern. Specifically, by not performing activities in accordance with a procedure the licensee could manipulate equipment and challenge the operators, and cause unexpected transients. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process (SDP) for Findings at Power, issued June 19, 2012, the finding was screened against the Initiating Events cornerstone and determined to be of very low safety significance because the finding did not cause a reactor trip or the loss of mitigation equipment and it did not involve the complete or partial loss of a support system that contributes to the likelihood of, or cause, an initiating event. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of challenging the unknown which stated, individuals stop when faced with uncertain conditions. Risks are evaluated and managed before proceeding. Contrary to this, when challenged with unknown conditions, the licensee did not stop and properly evaluate the issues before proceeding, resulting in adverse impacts to station equipment. (H.11)
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| Inspection Report# : 2015004 (pdf)
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| Significance: Sep 30, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation FAILURE TO FOLLOW PROCEDURE LEAVES CONTROL ROOM CABINET DOORS UNATTENDED IN SEISMICALLY UNANALYSED CONDITION The inspectors identified a finding of very low safety significance (Green) and an associated NCV of 10 CFR50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to maintain control room doors in a seismically analyzed condition, in accordance with station procedure CPS 1014.11, 6900/4160/480v Switchgear/Circuit Breaker Operability Program, Revision 5a. Specifically, on several occasions the licensee failed Page 1 of 14
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| 1Q/2016 Inspection Findings - Clinton to maintain control room cabinet doors in seismically qualified positions, while performing maintenance or trouble shooting activities, by leaving the doors open and unattended. The licensee documented the issue in the Corrective Action Program (CAP) as action request (AR) 02518477. The licensee has revised the station procedure to ensure control room cabinet doors either remain latched closed or are completely removed when unattended and has issued a standing order to ensure the requirements are reinforced.
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| The performance deficiency was more than minor in accordance with IMC 0612, "Power Reactor Inspection Reports,"
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| Appendix B, "Issue Screening," dated September 7, 2012, because, it was associated with the configuration control performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations and is therefore a finding. Specifically, leaving the control doors in a seismically unanalyzed condition could challenge critical safety functions during a seismic event. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process (SDP) for Findings at Power, issued June 19, 2012, the finding was screened against the Initiating Events cornerstone and determined to be of very low safety significance (Green) because the finding did not result in exceeding the reactor coolant system leak rate for a small loss of coolant accident (LOCA), cause a reactor trip, involve the complete or partial loss of a support system that contributes to the likelihood of, or caused, an initiating event and did not affect mitigation equipment. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of resources where leaders ensure that personnel, equipment, procedures and other resources are available and adequate to support nuclear safety. Specifically, the licensee failed to ensure the personnel performing maintenance and troubleshooting had adequate documentation in written work instructions to maintain control room cabinets in seismically analyzed conditions.
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| Inspection Report# : 2015003 (pdf)
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| Significance: Jun 30, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation POST MAINTENANCE TEST FAILED TO DEMONSTRATE REQUIRED FLOW THROUGH RCIC ROOM COOLER A self-revealed finding of very low safety significance and an associated Non-Cited Violation of 10 CFR 50 Appendix B, Criterion XI, Test Control, was documented by the inspectors for the failure to perform adequate post maintenance testing that would assure that the Reactor Core Isolation Cooling (RCIC) room cooler would perform its intended function when restored to service following maintenance. Specifically, the licensee declared the room cooler operable with insufficient cooling flow through the cooler. The licensee documented the issue in the licensees corrective action program (CAP) as action request (AR) 02447013. The licensee operated the RCIC Room Cooler outlet valve from its throttled position to fully open to flush the seat and the upstream piping and positioned the valve to maintain the required flow to restore the cooler to an operable condition.
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| The failure to perform adequate post maintenance testing that would assure that the RCIC Room cooler would perform its intended function when restored to service following maintenance is a performance deficiency. The performance deficiency was determined to be more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, because it was associated with the procedure quality attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences and is, therefore, a finding. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings at Power, dated June 19, 2012, the finding was screened against the Mitigating Systems cornerstone and determined to need a detailed risk evaluation because the finding represents the loss of a system and/or function. The Region III Senior Reactor Analysts (SRAs) evaluated the finding using the Clinton Station Standardized Plant Analysis Risk (SPAR) Model Version 8.17, Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) Version 8.1.2. The SRAs reviewed the Page 2 of 14
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| 1Q/2016 Inspection Findings - Clinton licensees Apparent Cause Investigation Report IR 2447013. The exposure time was assumed to be 150.5 hours based on information in that report. The SRAs modeled the condition using failure of the RCIC pump as a surrogate for failure of the RCIC room cooler. The basic event representing the RCIC pump failure-to-run was set to True for the 150.5 hour duration. The result was a ?CDF of 9.98E-08/yr. The dominant sequence was a station blackout initiating event; failure of high pressure core spray; failure of reactor core isolation cooling; and failure to recover offsite or emergency AC power within 30-minutes. Based on the detailed risk evaluation, the finding is best characterized as a finding of very low safety-significance (Green). The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of work management where the organization implements a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. The work process includes the identification and management of risk commensurate to the work and the need for coordination with different groups of job activities. Specifically, the licensee failed to plan and execute adequate post maintenance testing that would have ensured the satisfactory operation of the RCIC Room cooler following planned maintenance.
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| [H.5]
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| Inspection Report# : 2015002 (pdf)
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| Significance: Jun 30, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation FAILURE TO TRANSLATE SUFFICIENT GLAND STRESS TO PACKING GLAND NUTS RESULTED IN VALVE PACKING FAILURE AND PLANT SHUTDOWN A finding of very low safety significance and an associated Non-Cited Violation of 10 CFR50, Appendix B, Criterion III, Design Control, was self-revealed on January 19, 2015, when a steam leak developed from the RCIC system inboard steam isolation valve (1E51F0063) stem packing. Specifically, the licensee failed to identify and implement a torque value for the gland packing nuts for the RCIC system inboard steam isolation valve 1E51F0063 to overcome service induced consolidation and prevent packing leakage. This resulted in a plant down power to 83 percent and subsequent plant shutdown due to increasing unidentified reactor coolant system leakage. The licensee documented the issue in the licensees CAP as AR 02439437. The licensee repacked the valve utilizing the station procedure CPS 8120.37, Valve Packing Installation, and applicable SealPro data sheet. A four ring set of A.P. Services graphite packing was installed with a new live load assembly sized to a new torque value of 59 ft-lbs. and the valve packing was tested to verify no leakage.
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| The inspectors determined that the failure to apply sufficient packing gland torque to overcome service induced consolidation and prevent packing leakage on the RCIC system inboard steam isolation valve was a performance deficiency. The performance deficiency was more than minor in accordance with IMC 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," dated September 7, 2012, because, it was associated with the equipment performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations and is therefore a finding. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings at Power, issued June 19, 2012, the finding was screened against the Initiating Events cornerstone and determined to be of very low safety significance (Green) because the finding did not result in exceeding the RCS leak rate for a small LOCA, cause a reactor trip, involve the complete or partial loss of a support system that contributes to the likelihood of, or caused, an initiating event and did not affect mitigation equipment. The inspectors determined that no cross-cutting aspect would be associated with this finding since the performance deficiency occurred in 2010 and was not representative of current licensee performance in the of valve packing.
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| Inspection Report# : 2015002 (pdf)
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| Significance: Jun 30, 2015 Page 3 of 14
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| 1Q/2016 Inspection Findings - Clinton Identified By: NRC Item Type: FIN Finding FAILURE TO EVALUATE THE OPERATIONAL IMPACT OF THE TDRFP LOCKOUT SWITCH POSITION A self-revealed finding was identified for failure to evaluate the consequences of an adverse condition, in accordance with the operational decision making process. Specifically, contrary to station procedure OP-AA-106-101-1006 Operational Decision Making Process, Revision 14, the licensee failed to adequately implement the procedure to ensure the consequences of leaving the switch in the lockout position were evaluated, which resulted in the loss of the manual trip function for the A turbine driven reactor feed pump (TDRFP). The licensee documented the issue in the licensees CAP as action request (AR) 02440052. The licensee repaired the ground condition and returned the switch to its normal position. The licensee also revised the surveillance procedure to document the limitations associated with putting the emergency governor trip test and lockout switch in the lockout position.
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| The inspectors determined that the failure to adequately implement the procedure to ensure the consequences of leaving the switch in the lockout position were evaluated, which resulted in the loss of the manual trip function for the A TDRFP, was a performance deficiency. Specifically, by not evaluating leaving the emergency governor trip test and lockout switch in the lockout position, the licensee lost the ability to manually trip the A TDRFP, which challenged the operators during the reactor shutdown, and nearly resulted in a Level 8 reactor SCRAM. The performance deficiency was more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, because if left uncorrected the performance deficiency had the potential to lead to a more significant safety concern. The performance deficiency was also associated with the configuration control attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown and power operations and is therefore a finding. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings at Power, issued June 19, 2012, the finding was screened against the Initiating Events cornerstone and determined to be of very low safety significance (Green) because the finding did not cause a reactor trip or the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of resources where leaders ensure that personnel, equipment, procedures, and other resources are available and adequate to support nuclear safety. Specifically, the surveillance procedure for the RFPT emergency governor and trip mechanism test Section 2.1.1 stated if an actual signal was generated during testing, the lockout valve would de-energize to allow the trip mechanism to operate and trip the RFPT, which led to the understanding that the trip functions were unaffected by the switch position. (H.1)
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| Inspection Report# : 2015002 (pdf)
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| Mitigating Systems Significance: Mar 31, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Operability Determination Failed to Examine Test Failures The inspectors identified a finding of very low safety significance and an associated non-cited violation of 10, Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion V, Instructions Procedures and Drawings, for the failure to follow Station Procedure OP-AA-108-115, Operability Determinations, Revision 16. Specifically, after valve 1SX027C, a valve required for residual heat removal operability, failed a surveillance test, the licensee did not base the operability determination on a detailed examination of the deficiency and did not document a basis for why a reasonable expectation of operability existed. The licensee entered this issue into their corrective action program Page 4 of 14
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| 1Q/2016 Inspection Findings - Clinton (CAP) as Action Request (AR) 02553168 and AR 02558101. The licensee revised the in-service testing program evaluation for valve 1SX027C and documented additional details to support declaring the valve operable.
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| The inspectors determined the failure to follow Station Procedure OP-AA-108-115 was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to correctly perform an operability evaluation for valve 1SX027C had the potential to allow an inoperable condition to go undetected. Using IMC 0609, , Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2012, the finding was screened against the Mitigating Systems Cornerstone and determined to be of very low safety significance because the finding: was not a deficiency affecting the design or qualification of a mitigating system; did not represent a loss of system and/or function; did not represent an actual loss of function of a single train for greater than its Technical Specification (TS) allowed outage time; and did not represent an actual loss of function of one or more non-TS trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. The inspectors determined this finding affected the cross-cutting area of human performance, in the aspect of resources, where leaders ensure that personnel, equipment, procedures, and other resources are available and adequate to support nuclear safety.
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| Specifically, Station Procedure CPS 9053.04, provided guidance that the valve could remain operable for 96 hours without providing an appropriate basis.
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| Inspection Report# : 2016001 (pdf)
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| Significance: Mar 31, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Inadequate Extent of Condition Associate with an ACE The inspectors identified a finding of very low safety significance and an associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion II, Quality Assurance Program, for the failure to follow a Quality Assurance Program implementing procedure. Specifically, the licensee failed to perform an adequate extent of condition review as required by PI-AA-125, Corrective Action Program, while evaluating a lack of proficiency in applying the licensing basis for structures, systems and components (SSCs) when implementing the 50.59 process. The licensee documented this issue in their CAP as AR 02641397. Immediate corrective actions included a review of the extent of condition performed by the engineering department and a recommended action of expanding the scope of the review to include additional 50.59 evaluations.
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| The inspectors determined the failure to follow a Quality Assurance Program implementing procedure was more than minor because if left uncorrected it had the potential to lead to a more significant safety concern. Specifically, if the extent of condition review is too narrowly assessed there is the potential for other safety significant systems to have been impacted by a lack of proficiency in applying the licensing basis. As a result, the SSCs may not perform their intended safety function as defined in the Updated Safety Analysis Report. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings at Power, issued June 19, 2012, the finding was screened against all cornerstones and determined to be of very low safety significance because there was no reasonable indication that the criteria in Appendix A were met. The inspectors determined this finding affected the cross-cutting area of human performance, in the aspect of procedure adherence, where individuals follow processes, procedures and work instructions. Specifically, the licensee did not effectively adhere to all available portions of CAP procedures, which led to a narrowly focused extent of condition.
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| Inspection Report# : 2016001 (pdf)
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| Page 5 of 14
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| 1Q/2016 Inspection Findings - Clinton Significance: Feb 04, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Perform and Adequate Equipment Apparent Cause Evaluation (Section 4OA4)
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| The inspectors identified a finding of very-low safety significance (Green), and an associated Non-Cited Violation of Title 10, Code of Federal Regulations, Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to follow Step 4.3.4 of procedure PI-AA-125, Corrective Action Program Procedure.
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| Specifically, the licensee failed to perform Class B Equipment Apparent Cause Evaluation (EACE) 2381871, 1SX01PC Failed to Start for Testing, in accordance with PI-AA-125-1003, Apparent Cause Evaluation Manual, because they: (1) failed to analyze each causal factor to determine contributing causes as required by Step 4.4.1.2; and (2) failed to assign an effectiveness review for the EACE as required by Step 4.4.9.1. The licensee entered this finding into their Corrective Action Program and revised their EACE to: (1) include three contributing causes; (2) upgrade a corrective action to a corrective action to prevent recurrence; and (3) assign an effectiveness review to determine the effectiveness of the corrective action to prevent recurrence.
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| The performance deficiency was determined to be more than minor because if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, an effectiveness review is required to provide assurance that the Division 3 SX pump design change is successful in preventing recurrence of pump failure before another pump failure occurs, which would be a more significant safety concern. The finding impacted the Mitigating Systems Cornerstone and screened as having very-low safety significance (Green) because although the finding is a deficiency ultimately affecting the design or qualification of the Division 3 SX pump, the pump still maintains its operability. The inspectors determined this finding had an associated cross-cutting aspect in the area of Human Performance (Conservative Bias) because although a B Apparent Cause Evaluation may have been allowable for investigating the failure of the Division 3 SX pump, had an A Root Cause Analysis been performed, a more rigorous investigation process would have been used to identify contributing causes, assign corrective actions, and identify effectiveness reviews for the failure of the Division 3 SX pump. [H.14] (Section 4OA4.02.03.f)
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| Inspection Report# : 2016008 (pdf)
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| Significance: Dec 31, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Perform Activities Affecting Quality in Accordance with Prescribed Procedures The inspectors identified a finding of very low safety significance and an associated Non-Cited Violation of 10 CFR50, Appendix B, Criterion V, Instructions Procedures and Drawings, for the failure to ensure that activities affecting quality were accomplished in accordance with the appropriate instructions, procedures and drawings.
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| Specifically, the inspectors identified two examples where the licensee failed to perform activities affecting quality in accordance with prescribed procedures. The licensee entered this issue into their corrective action program as action request (AR) 02600726 and planned to perform an apparent cause evaluation to address the trend. Separate action requests were also written and immediate corrective actions were taken for each identified example to address the nonconformances created by the failure to follow procedures.
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| The inspectors determined that the failure to ensure that activities affecting quality were accomplished in accordance with the appropriate instructions, procedures and drawings as required by 10 CFR 50 Appendix B Criterion V, was a performance deficiency. Specifically, the inspectors identified two instances where the licensee failed to follow procedures resulting in impacts to safety related equipment and processes. The performance deficiency was more than minor in accordance with Inspection Manual Chapter (IMC) 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," dated September 7, 2012, because, if left uncorrected the performance deficiency had the potential to lead to a more significant safety concern. Specifically, by not performing activities affecting quality in accordance Page 6 of 14
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| 1Q/2016 Inspection Findings - Clinton with a procedure the licensee could manipulate equipment and challenge the operators by causing unexpected transients or impact safety related equipment. Using IMC 0609, Appendix G, Shutdown Operations Significance Determination Process, Attachment 1, issued May 9, 2014, the finding was screened against the Mitigating Systems cornerstone and determined to be of very low safety significance because the finding did not represent a loss of system safety function, it did not represent an actual loss of function of a single train or two separate trains for greater than its allowed outage time, it did not represent an actual loss of safety function of one or more non-TS trains of equipment during shutdown for equipment designated as risk significant for greater than 24 hours, and it did not degrade a functional auto-isolation of residual heat removal (RHR) on low reactor vessel level. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of challenging the unknown which states, individuals stop when faced with uncertain conditions. Risks are evaluated and managed before proceeding. Contrary to this, when challenged with uncertain conditions, the licensee did not stop and properly evaluate the issues before proceeding, resulting in adverse impacts to safety related equipment and activities. (H.11)
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| Inspection Report# : 2015004 (pdf)
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| Significance: Oct 09, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Generate Issue Reports for Conditions Adverse to Quality The inspectors identified a finding of very low safety significance, and an associated NCV of Title 10, Code of Federal Regulations, Part 50, Appendix B, Criterion II, Quality Assurance Program, for the failure to perform activities in accordance with procedure PI-AA-125, Corrective Action Program, Revision 2, which was a Quality Assurance Program implementing procedure. Specifically, the inspectors identified six examples where the licensee failed to generate IRs for conditions adverse to quality (CAQ) as required by PI-AA-125, until prompted by the inspectors. The licensee documented the issue in the CAP as IR 2518477, and planned on reviewing the apparent cause evaluation to determine if additional actions needed to be taken.
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| The performance deficiency was more than minor in accordance with IMC 0612, "Power Reactor Inspection Reports,"
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| Appendix B, "Issue Screening," dated September 7, 2012, because, if left uncorrected the performance deficiency had the potential to lead to a more significant safety concern. Specifically, by not identifying and documenting conditions adverse to quality the issues would not go through the screening and review process in accordance with the corrective action procedure, which could impact the identification of conditions affecting operability. The finding was screened against the Mitigating Systems cornerstone, and determined to be of very low safety significance because the it did not represent a loss of safety system or function, it did not represent an actual loss of function of a single train of two separate trains for greater than its allowed outage time and it did not represent a loss of function of a non-technical specification system designated as highly safety-significant within the licensees Maintenance Rule Program for greater than 24 hours. The inspectors determined this finding affected the cross-cutting area of problem identification and resolution in the aspect of identification where the organization implements a CAP with a threshold for identifying issues and individuals identify issues completely, accurately and in a timely manner in accordance with the program. Specifically, the licensee failed to identify issues completely, accurately and in a timely manner, causing them to not recognize issues as CAQs, and therefore not follow their process for handling these issues.
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| Inspection Report# : 2015007 (pdf)
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| Significance: Sep 30, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation FAILURE TO IMPLEMENT AND COMPLY WITH TRANSIENT EQUIPMENT/MATERIALS PROGRAM The inspectors identified a green finding and an associated NCV of 10 CFR 50, Appendix B, Criterion V Instructions, Procedures, and Drawings for the licensees failure to implement and comply with station procedure Page 7 of 14
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| 1Q/2016 Inspection Findings - Clinton CPS 1019.05, Transient Equipment/Materials, Revision 23, to ensure that transient equipment and materials are controlled so there is no impact to safe operation of plant equipment. Specifically, on numerous occasions the inspectors identified equipment and materials improperly staged, improperly secured or in areas without engineering evaluations. The licensee documented the issue in the CAP as action requests (AR) 02507167 and AR 02529227. In each occasion identified by the inspectors the licensee subsequently removed the items identified to restore compliance with the station procedures.
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| The inspectors determined the licensees failure to implement and comply with station procedures to ensure that transient equipment and materials are controlled so there is no impact to safe operation of plant equipment was a performance deficiency. The performance deficiency was determined to be more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Screening, dated September 7, 2012, because if left uncorrected it had the potential to lead to a more significant safety concern.
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| Specifically, transient equipment and material in proximity of safety related components has the potential of impacting these components during a seismic event, potentially rendering them unable to fulfill their safety function.
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| The performance deficiency is also associated with the protection against external factors attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability and capability of systems that response to initiating events to prevent undesirable consequences, and is therefore a finding.
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| Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process (SDP) for Findings at Power, issued June 19, 2012, the finding was screened against the Mitigating Systems cornerstone and determined to be of very low safety significance (Green) because the finding did not represent a loss of system or function, it did not represent an actual loss of function of at least a single train for >
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| its TS allowed outage time and it did not represent an actual loss of one or more not TS trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of field presence where leaders are commonly seen in the work areas of the plant observing, coaching, reinforcing standards and expectation. Deviations from standards and expectations are corrected promptly. Specifically, after various examples of material placement being an issue, the licensee didnt perform in field observations, caching and reinforcement of standards and expectations in the identified areas.
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| Inspection Report# : 2015003 (pdf)
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| Barrier Integrity Significance: Mar 31, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Identify a Degraded Safety-Related Support
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| . The inspectors identified a finding of very low safety significance and an associated non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for the failure to identify a condition adverse to quality.
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| Specifically, the licensee failed to identify that a safety-related support associated with control room ventilation B was degraded to the point it no longer conformed to the seismic analysis and required an evaluation to determine whether it was still capable of performing its safety function during a seismic event. This issue was entered into the licensees CAP as AR 2639317. The licensees immediate corrective actions included performing an evaluation that concluded the remaining three supports would be able to withstand the stresses imposed during a seismic event and creating an action to update the seismic calculation to incorporate the evaluation performed for the degraded support.
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| The licensee also planned to re-apply a coating to the supports as well as research and install insulation that was more breathable to minimize moisture accumulation and preclude any further degradation.
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| Page 8 of 14
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| 1Q/2016 Inspection Findings - Clinton The inspectors determined that the failure to identify a condition adverse to quality in accordance with 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, was more than minor because if left uncorrected it had the potential to lead to a more significant safety concern. Specifically, by failing to identify the support was degraded, and correct the condition, the loss of material due to corrosion could potentially progress to the point where the remaining supports would no longer be able to perform their safety function. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings at Power, issued June 19, 2012, the finding was screened against the Barrier Integrity Cornerstone and determined to be of very low safety significance because the finding did not represent a degradation of the barrier function of the control room against radiological conditions or a smoke or toxic atmosphere. The inspectors determined this finding affected the cross-cutting area of problem identification and resolution, in the aspect of evaluation, which states, The organization thoroughly evaluates issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, the licensee failed to thoroughly evaluate the issue identified by the inspectors and therefore did not recognize the degradation on the supports constituted a condition adverse to quality.
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| Inspection Report# : 2016001 (pdf)
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| Significance: Mar 31, 2016 Identified By: Self-Revealing Item Type: NCV Non-Cited Violation Failure to Assess and Manage Risk Increase for a Proposed Maintenance Activity A self-revealed finding of very low safety significance and an associated non-cited violation of 10 CFR 50.65 (a)(4) was identified on January 20, 2016, due to the licensees failure to assess and manage the risk increase from a proposed maintenance activity. Specifically, the licensee failed to manage the risk associated with racking out the continuous containment purge (CCP) A breaker, which resulted in the loss of both CCP trains, and led to an increase in primary to secondary containment differential pressure which exceeded the TS value. The licensee entered this issue into their CAP as AR 02614832. The proposed corrective actions to address this issue included creating a checklist to ensure validation of initial conditions is performed and providing training that reinforces the need to properly screen work order tasks with the appropriate risk factors.
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| The inspectors determined that the failure to assess and manage the risk increase of a proposed maintenance activity, as required by 10 CFR 50.65 (a)(4), was more than minor because it was associated with the maintenance procedure quality attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, by not properly assessing the risk of racking out the CCP A breaker the licensee did not recognize the CCP B train would be impacted, which resulted in exceeding the TS value for primary to secondary containment differential pressure. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings at Power, issued June 19, 2012, the finding was screened against the Barrier Integrity Cornerstone and determined to be of very low safety significance because the finding did not represent an actual open pathway in the physical reactor containment, containment isolation system or heat removal components and it did not involve an actual reduction in function of hydrogen igniters in the reactor containment. The inspectors identified a cross-cutting aspect in the area of human performance, in the aspect of challenging the unknown, which states, individuals stop when faced with uncertain conditions; risks are evaluated and managed before proceeding. Specifically, when the licensee was preparing the work package for maintenance on the CCP system it was uncertain what activities had already been completed as part of a concurrent evolution. Instead of stopping and validating the configuration of plant equipment, assumptions were made, and the risk of the activity was not properly assessed or managed.
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| Inspection Report# : 2016001 (pdf)
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| Page 9 of 14
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| 1Q/2016 Inspection Findings - Clinton Significance: Sep 30, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation FAILURE TO OBTAIN A LICENSE AMENDMENT PRIOR TO MAKING MODIFICATIONS TO SECONDARY CONTAINMENT The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.59(d)(1), Changes, Tests, and Experiments for the licensees failure to provide a written evaluation, which provided the basis for determining that the change to the secondary containment completed on December 18, 2014 did not require a license amendment.
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| Specifically, the licensee made a change pursuant to 10 CFR 50.59(c), to the secondary containment, and eliminated the tornado wind and tornado missile loading condition from the FB Railroad Airlock (the enclosure walls and roof) and associated outer door (1SD1-31) Seismic Category I requirements and did not provide a written evaluation to provide a basis for the determination that this change would not result in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system or component important to safety.
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| The inspectors determined that the licensees failure to provide a written evaluation, which provided the basis for determining that the change to the secondary containment completed on December 18, 2014 did not require a license amendment was a performance deficiency. Specifically, the licensee made a change pursuant to 10 CFR 50.59(c) to the secondary containment and eliminated the tornado wind and tornado missile loading condition from the FB Railroad Airlock (the enclosure walls and roof) and associated outer door and did not provide a written evaluation to provide a basis for the determination that this change would not result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety. The performance deficiency was determined to be more than minor in accordance with IMC 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," dated September 7, 2012, because it was associated with the design control attribute of the barrier integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system and containment) protect the public from radionuclide releases caused by accidents or events. In addition, the associated violation was determined to be more than minor because the inspectors could not reasonably determine if the changes to secondary containment would have required NRC prior approval. The licensee documented the issue in the CAP as action request (AR) 02534694. The licensee is complying with technical specifications anytime the inner railroad bay door is opened by entering the applicable action statements, evaluating weather conditions and impact to plant risk and establishing the necessary mitigating actions required prior to opening the door. Violations of 10 CFR 50.59 are dispositioned using the traditional enforcement process instead of the SDP because they are considered to be violations that potentially impede or impact the regulatory process. However, if possible, the underlying technical issue is evaluated under the SDP to determine the severity of the violation. In this case, the inspectors used IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings at Power, issued June 19, 2012, the finding was screened against the barrier integrity cornerstone and determined to be of very low safety significance (Green) because the finding did not represent a degradation only of the radiological barrier function for the Standby Gas Treatment (SBGT) system nor did it represent a degradation of the function of the control room against smoke or toxic atmosphere. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of procedure adherence where individuals follow processes, procedures and work instructions. Specifically, the licensee failed to follow the 50.59 regulatory process as defined in station procedure LS-AA-104-1000, 50.59 Resource Manual, Revision 9.
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| Inspection Report# : 2015003 (pdf)
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| Significance: Sep 30, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation FAILURE TO ENTER APPROPRIATE TS ACTION STATEMENT FOR INOPERABLE RADIATION Page 10 of 14
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| 1Q/2016 Inspection Findings - Clinton MONITORS DURING OPDRV ACTIVITIES The inspectors identified a green finding and associated NCV of T.S. 3.3.6.1 Primary Containment and Drywell Isolation Instrumentation and 3.3.6.2 Secondary Containment Isolation Instrumentation, for the failure to enter the appropriate action statement and take the associated actions related to inoperable containment radiation monitor instrumentation during operations with the potential to drain the reactor vessel. Specifically, with the containment ventilation dampers closed, the containment radiation monitor instrumentation would not be able to perform its safety function of sending a containment isolation signal for elevated containment radiation levels as required during OPDRVS. At the time of discovery the licensee had already concluded OPDRV activities and was therefore no longer in a mode of applicability. The licensee documented the issue in the CAP as action request (AR) 2566708. When this issue was identified the maintenance on the VR/VQ system was complete and no OPDRVs were in progress, therefor the T.S. noncompliance was no longer in effect.
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| The inspectors determined that the failure to enter T.S. 3.3.6.1 and 3.3.6.2 when the radiation monitor instrumentation was not able to perform its safety function during an OPDRV, was a performance deficiency. Specifically, the licensee failed to recognize that when the containment ventilation dampers were closed, the radiation monitors could not detect the radiation levels in primary containment and therefore could not fulfill their safety function of sending containment isolation signals in the case of elevated radiation levels in containment. The performance deficiency was more than minor in accordance with IMC 0612, Power Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, because, it was associated with the SSC and Barrier Performance attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events, and is therefore a finding.
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| Specifically, the automatic containment isolation signal function of the radiation monitors was impacted when the containment ventilation dampers were closed during OPDRV operations. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings, dated May 9, 2014, the finding was screened against the Barrier Integrity cornerstone and determined to need a detailed risk evaluation because the finding represents a degradation of the ability to close or isolate the containment. Using Appendix G Exhibit 4, Barrier Integrity Screening Questions, the Senior Reactor Analyst (SRA) determined that the finding degraded the ability to close or isolate the containment per Section B, Containment Barrier, Question 6. Therefore, the evaluation was continued using IMC 0609 Appendix H, Containment Integrity Significance Determination Process. The SRA determined this to be a Type B finding, because it was related to a degraded condition that had implications for containment integrity without affecting the likelihood of core damage. The SRA used Section 6.2 of Appendix H, Approach for Assessing Type B Findings at Shutdown. Based on information from the inspectors, during all OPDRV time windows, the reactor water level was confirmed to be greater than the minimum level required for movement of irradiated fuel assemblies (i.e., greater than 228 above the flange). This plant condition meets the definition of Plant Operating State 3 (POS 3) of Appendix H. Therefore, based on the plant being in POS 3 during the OPDRV time windows, the finding screens as Green based on Step 2.1 of Section 6.2 of Appendix H. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of conservative bias where individuals use decision making practices that emphasize prudent choices over those that are simple allowable. A proposed action is determined to be safe in order to proceed, rather than unsafe in order to stop.
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| Specifically, the licensee relied solely on the successful completion of the surveillance requirements to determine the radiation monitor instrumentation was operable rather than considering the impact the closed dampers would have on their ability to fulfill their safety function.
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| Inspection Report# : 2015003 (pdf)
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| Emergency Preparedness Page 11 of 14
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| 1Q/2016 Inspection Findings - Clinton Occupational Radiation Safety Significance: Jun 30, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation CONTRACT WORKERS NOT MONITORED FOR OCCUPATIONAL RADIATION EXPOSURE The inspectors identified a finding of very-low safety significance and an associated NCV of Technical Specification (TS) 5.4.1, Procedures, for the failure to monitor the radiation dose received by a group of workers as required by station procedure RP-AA-210, Dosimetry Issue, Usage, and Control. Specifically, contractor employees who did not wear individual dosimetry were not monitored by the usage of an Area Badging Program and the workers were not excluded from wearing individual dosimetry by the usage of medical isotopes or external radioactivity being detected, or a previously performed evaluation by RP Supervision. The licensee documented the issue in the licensees CAP as action request AR 02452005. The trailer was relocated to a distance further away from the radioactive material storage area. This reduced the radiation dose rate in the trailer.
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| The inspectors determined that the issue of concern was a performance deficiency because the licensee did not monitor a group of workers using one or more methods as required by procedure, RP-AA-210, Dosimetry Issue, Usage and Control. The licensee did not assign radiation dosimetry to each worker, nor was an Area Badging Program in place. The inspectors determined that the cause of the performance deficiency was reasonably within the licensees ability to foresee and correct and should have been prevented. The issue was not subject to traditional enforcement since the concern did not have a significant safety consequence, did not impact the NRCs ability to perform its regulatory function, and was not willful. The performance deficiency was determined to be of more than minor safety significance in accordance with IMC 0612, Appendix B, Issue Screening, issued September 7, 2012, because it was associated with the program and process attribute of the Occupational Radiation Safety Cornerstone, and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation. Specifically, the licensee could not demonstrate compliance with other sections of 10 CFR Part 20, such as occupational dose limits, and records and reporting of individual monitoring results. The inspectors also reviewed the guidance in IMC 0612, Appendix E, Examples of Minor Issues, and did not find any similar examples.
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| In accordance with IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, issued August 19, 2008, the inspectors determined that the finding had very low safety significance (Green) because the finding: (1) did not involve as-low-as-reasonably-achievable planning and controls; (2) did not involve a radiological overexposure; (3) there was not a substantial potential for an overexposure; and (4) there was no compromised ability to assess dose. This finding has a cross-cutting aspect in the area of Human Performance, Change Management, because the primary cause of the finding was due to inadequate change management.
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| Specifically, licensee supervision incorrectly located the trailer near a posted radiation area without performing an appropriate evaluation to ensure the personnel or area was correctly monitored. [H.3]
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| Inspection Report# : 2015002 (pdf)
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| Public Radiation Safety Security Although the Security Cornerstone is included in the Reactor Oversight Process assessment program, the Commission has decided that specific information related to findings and performance indicators pertaining to the Security Page 12 of 14
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| 1Q/2016 Inspection Findings - Clinton Cornerstone will not be publicly available to ensure that security information is not provided to a possible adversary.
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| Other than the fact that a finding or performance indicator is Green or Greater-Than-Green, security related information will not be displayed on the public web page. Therefore, the cover letters to security inspection reports may be viewed.
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| Miscellaneous Significance: N/A Mar 31, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Report a Condition that Could Have Prevented Fulfillment of a Safety Function The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.72(b)(3)(v) for failing to report an event or condition, that at the time of discovery could have prevented the fulfillment of a safety function, to the NRC within eight hours. Specifically, control room operators placed both divisions of reactor water cleanup differential flow instruments in bypass, which rendered the instruments inoperable and resulted in a loss of the isolation function.
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| The licensee entered this issue into the CAP as AR 02645140 and created an action to submit an licensee event report under 10 CFR 50.73(a)(2)(v).
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| The inspectors determined that the failure to report an event or condition, that at the time of discovery could have prevented the fulfillment of a safety function, to the NRC within 8 hours as required by 10 CFR 50.72(b)(3)(v) was a performance deficiency. The inspectors reviewed this issue in accordance with IMC 0612 and the Enforcement Manual. Violations of 10 CFR 50.72 are dispositioned using the traditional enforcement process because they are considered to be violations that potentially impede or impact the regulatory process. The inspectors reviewed Section 6.9.d.9 of the NRC Enforcement Policy and determined this violation was Severity Level IV because the licensees failure to make the report, as required by 10 CFR 50.72, did not cause the NRC to reconsider a regulatory position or undertake substantial further inquiry. No cross-cutting aspect was assigned because cross-cutting aspects are not assigned to traditional enforcement only violations.
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| Inspection Report# : 2016001 (pdf)
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| Significance: N/A Mar 31, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Report Condition Prohibited by Technical Specifications The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.73(a)(2)(i)(B) for failing to report to the NRC, within 60 days of discovery, a condition prohibited by the plants TS. Specifically, the licensee failed to notify the NRC of two instances where they failed to comply with TS 3.3.6.1 and TS 3.3.6.2 and enter the limiting condition for operation action statements when required. The licensee entered this issue into their CAP as AR 02619114 and subsequently issued a licensee event report on March 16, 2016.
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| The inspectors determined that the failure to report a condition prohibited by the plants TS as required by 10 CFR 50.73(a)(2)(i)(B), within 60 days of discovery, was a performance deficiency. The inspectors reviewed this issue in accordance with IMC 0612 and the Enforcement Manual. Violations of 10 CFR 50.73 are dispositioned using the traditional enforcement process because they are considered to be violations that potentially impede or impact the regulatory process. The inspectors reviewed Section 6.9.d.9 of the NRC Enforcement Policy and determined this violation was Severity Level IV because the licensees failure to make the report, as required by 10 CFR 50.73, did not cause the NRC to reconsider a regulatory position or undertake substantial further inquiry. No cross-cutting aspect was assigned because cross-cutting aspects are not assigned to traditional enforcement only violations.
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| Page 13 of 14
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| 1Q/2016 Inspection Findings - Clinton Inspection Report# : 2016001 (pdf)
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| Significance: N/A Dec 31, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation FAILURE TO UPDATE THE FINAL SAFETY ANALYSIS REPORT (FSAR) - HYDROGEN WATER CHEMISTRY SYSTEM The inspectors identified a Severity Level IV Violation of Title 10 Code of Federal Regulations (CFR) 50.71(e),
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| "Periodic Update of the FSAR", for the licensee's failure to update the FSAR after installing a hydrogen water chemistry system into the plant to reduce rates of intergranular stress corrosion cracking (IGSCC) in recirculation piping and reactor vessel internals. Specifically, the licensee did not update Section 5.4.15, "Hydrogen Water Chemistry System" of the FSAR to include a design basis and description of process and system used to periodically injection noble metals. The licensee entered this issue into the corrective action program as AR 02594259 and is revising the FSAR include additional the design basis and additional system description for noble metal injection.
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| The inspectors determined that the failure to update the FSAR in accordance with 10 CFR 50.71(e), "Periodic Update of the FSAR", with the design basis and description of the process and system used to periodically injection noble metals was a performance deficiency warranting a significance evaluation. The inspectors reviewed this issue in accrodance with NRC inspection manual chapter 0612 and the NRC enforcement manual. Violations of 10 CFR 50.71 (e) are dispositioned using the traditional enforcement process because they are considered to be violations that potentially impede or impact the regulatory process. The inspectors reviewed section 6.1.d.3 of the NRC Enforcement Policy and determined this violation was Severity :Level IV because the licensee's failure to update the FSAR as required by 10 CFR 50.71(e) had not yet resulted in any unacceptable change to the facility or procedures. No cross cutting aspect was assigned because cross cutting aspects are not assigned to traditional enforcement only violations.
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| Inspection Report# : 2015004 (pdf)
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| Last modified : July 11, 2016 Page 14 of 14
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| 2Q/2016 Inspection Findings - Clinton Clinton 2Q/2016 Plant Inspection Findings Initiating Events Significance: Dec 31, 2015 Identified By: NRC Item Type: FIN Finding Failure to Follow Station Procedures for Plant Activities The inspectors identified a finding of very low safety significance for the failure to ensure that activities were accomplished in accordance with prescribed procedures as required by station procedure HU-AA-104-101 Procedure Use and Adherence. Specifically, the inspectors identified two examples where the licensee failed to adhere to prescribed station procedures when performing activities in the plant. The licensee placed both issues in their corrective action program as AR 02600726 and addressed the nonconformances created by the failure to follow the procedures. The licensee planned to perform an apparent cause evaluation to determine why there was an adverse trend related to procedure adherence.
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| The inspectors determined that the failure to perform activities in accordance with prescribed procedures as required by station procedure HU-AA-104-101, Procedure Use and Adherence, was a performance deficiency. Specifically, the inspectors identified two instances where the licensee failed to follow procedures when performing activities in the plant. The performance deficiency was more than minor in accordance with IMC 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," dated September 7, 2012, because, if left uncorrected the performance deficiency had the potential to lead to a more significant safety concern. Specifically, by not performing activities in accordance with a procedure the licensee could manipulate equipment and challenge the operators, and cause unexpected transients. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process (SDP) for Findings at Power, issued June 19, 2012, the finding was screened against the Initiating Events cornerstone and determined to be of very low safety significance because the finding did not cause a reactor trip or the loss of mitigation equipment and it did not involve the complete or partial loss of a support system that contributes to the likelihood of, or cause, an initiating event. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of challenging the unknown which stated, individuals stop when faced with uncertain conditions. Risks are evaluated and managed before proceeding. Contrary to this, when challenged with unknown conditions, the licensee did not stop and properly evaluate the issues before proceeding, resulting in adverse impacts to station equipment. (H.11)
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| Inspection Report# : 2015004 (pdf)
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| Significance: Sep 30, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation FAILURE TO FOLLOW PROCEDURE LEAVES CONTROL ROOM CABINET DOORS UNATTENDED IN SEISMICALLY UNANALYSED CONDITION The inspectors identified a finding of very low safety significance (Green) and an associated NCV of 10 CFR50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to maintain control room doors in a seismically analyzed condition, in accordance with station procedure CPS 1014.11, 6900/4160/480v Switchgear/Circuit Breaker Operability Program, Revision 5a. Specifically, on several occasions the licensee failed Page 1 of 12
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| 2Q/2016 Inspection Findings - Clinton to maintain control room cabinet doors in seismically qualified positions, while performing maintenance or trouble shooting activities, by leaving the doors open and unattended. The licensee documented the issue in the Corrective Action Program (CAP) as action request (AR) 02518477. The licensee has revised the station procedure to ensure control room cabinet doors either remain latched closed or are completely removed when unattended and has issued a standing order to ensure the requirements are reinforced.
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| The performance deficiency was more than minor in accordance with IMC 0612, "Power Reactor Inspection Reports,"
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| Appendix B, "Issue Screening," dated September 7, 2012, because, it was associated with the configuration control performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations and is therefore a finding. Specifically, leaving the control doors in a seismically unanalyzed condition could challenge critical safety functions during a seismic event. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process (SDP) for Findings at Power, issued June 19, 2012, the finding was screened against the Initiating Events cornerstone and determined to be of very low safety significance (Green) because the finding did not result in exceeding the reactor coolant system leak rate for a small loss of coolant accident (LOCA), cause a reactor trip, involve the complete or partial loss of a support system that contributes to the likelihood of, or caused, an initiating event and did not affect mitigation equipment. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of resources where leaders ensure that personnel, equipment, procedures and other resources are available and adequate to support nuclear safety. Specifically, the licensee failed to ensure the personnel performing maintenance and troubleshooting had adequate documentation in written work instructions to maintain control room cabinets in seismically analyzed conditions.
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| Inspection Report# : 2015003 (pdf)
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| Mitigating Systems Significance: Mar 31, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Operability Determination Failed to Examine Test Failures The inspectors identified a finding of very low safety significance and an associated non-cited violation of 10, Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion V, Instructions Procedures and Drawings, for the failure to follow Station Procedure OP-AA-108-115, Operability Determinations, Revision 16. Specifically, after valve 1SX027C, a valve required for residual heat removal operability, failed a surveillance test, the licensee did not base the operability determination on a detailed examination of the deficiency and did not document a basis for why a reasonable expectation of operability existed. The licensee entered this issue into their corrective action program (CAP) as Action Request (AR) 02553168 and AR 02558101. The licensee revised the in-service testing program evaluation for valve 1SX027C and documented additional details to support declaring the valve operable.
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| The inspectors determined the failure to follow Station Procedure OP-AA-108-115 was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to correctly perform an operability evaluation for valve 1SX027C had the potential to allow an inoperable condition to go undetected. Using IMC 0609, , Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2012, the finding was screened against the Mitigating Systems Cornerstone and Page 2 of 12
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| 2Q/2016 Inspection Findings - Clinton determined to be of very low safety significance because the finding: was not a deficiency affecting the design or qualification of a mitigating system; did not represent a loss of system and/or function; did not represent an actual loss of function of a single train for greater than its Technical Specification (TS) allowed outage time; and did not represent an actual loss of function of one or more non-TS trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. The inspectors determined this finding affected the cross-cutting area of human performance, in the aspect of resources, where leaders ensure that personnel, equipment, procedures, and other resources are available and adequate to support nuclear safety.
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| Specifically, Station Procedure CPS 9053.04, provided guidance that the valve could remain operable for 96 hours without providing an appropriate basis.
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| Inspection Report# : 2016001 (pdf)
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| Significance: Mar 31, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Inadequate Extent of Condition Associate with an ACE The inspectors identified a finding of very low safety significance and an associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion II, Quality Assurance Program, for the failure to follow a Quality Assurance Program implementing procedure. Specifically, the licensee failed to perform an adequate extent of condition review as required by PI-AA-125, Corrective Action Program, while evaluating a lack of proficiency in applying the licensing basis for structures, systems and components (SSCs) when implementing the 50.59 process. The licensee documented this issue in their CAP as AR 02641397. Immediate corrective actions included a review of the extent of condition performed by the engineering department and a recommended action of expanding the scope of the review to include additional 50.59 evaluations.
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| The inspectors determined the failure to follow a Quality Assurance Program implementing procedure was more than minor because if left uncorrected it had the potential to lead to a more significant safety concern. Specifically, if the extent of condition review is too narrowly assessed there is the potential for other safety significant systems to have been impacted by a lack of proficiency in applying the licensing basis. As a result, the SSCs may not perform their intended safety function as defined in the Updated Safety Analysis Report. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings at Power, issued June 19, 2012, the finding was screened against all cornerstones and determined to be of very low safety significance because there was no reasonable indication that the criteria in Appendix A were met. The inspectors determined this finding affected the cross-cutting area of human performance, in the aspect of procedure adherence, where individuals follow processes, procedures and work instructions. Specifically, the licensee did not effectively adhere to all available portions of CAP procedures, which led to a narrowly focused extent of condition.
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| Inspection Report# : 2016001 (pdf)
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| Significance: Feb 04, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Perform and Adequate Equipment Apparent Cause Evaluation (Section 4OA4)
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| The inspectors identified a finding of very-low safety significance (Green), and an associated Non-Cited Violation of Title 10, Code of Federal Regulations, Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to follow Step 4.3.4 of procedure PI-AA-125, Corrective Action Program Procedure.
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| Specifically, the licensee failed to perform Class B Equipment Apparent Cause Evaluation (EACE) 2381871, 1SX01PC Failed to Start for Testing, in accordance with PI-AA-125-1003, Apparent Cause Evaluation Manual, because they: (1) failed to analyze each causal factor to determine contributing causes as required by Step 4.4.1.2; and Page 3 of 12
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| 2Q/2016 Inspection Findings - Clinton (2) failed to assign an effectiveness review for the EACE as required by Step 4.4.9.1. The licensee entered this finding into their Corrective Action Program and revised their EACE to: (1) include three contributing causes; (2) upgrade a corrective action to a corrective action to prevent recurrence; and (3) assign an effectiveness review to determine the effectiveness of the corrective action to prevent recurrence.
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| The performance deficiency was determined to be more than minor because if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, an effectiveness review is required to provide assurance that the Division 3 SX pump design change is successful in preventing recurrence of pump failure before another pump failure occurs, which would be a more significant safety concern. The finding impacted the Mitigating Systems Cornerstone and screened as having very-low safety significance (Green) because although the finding is a deficiency ultimately affecting the design or qualification of the Division 3 SX pump, the pump still maintains its operability. The inspectors determined this finding had an associated cross-cutting aspect in the area of Human Performance (Conservative Bias) because although a B Apparent Cause Evaluation may have been allowable for investigating the failure of the Division 3 SX pump, had an A Root Cause Analysis been performed, a more rigorous investigation process would have been used to identify contributing causes, assign corrective actions, and identify effectiveness reviews for the failure of the Division 3 SX pump. [H.14] (Section 4OA4.02.03.f)
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| Inspection Report# : 2016008 (pdf)
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| Significance: Dec 31, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Perform Activities Affecting Quality in Accordance with Prescribed Procedures The inspectors identified a finding of very low safety significance and an associated Non-Cited Violation of 10 CFR50, Appendix B, Criterion V, Instructions Procedures and Drawings, for the failure to ensure that activities affecting quality were accomplished in accordance with the appropriate instructions, procedures and drawings.
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| Specifically, the inspectors identified two examples where the licensee failed to perform activities affecting quality in accordance with prescribed procedures. The licensee entered this issue into their corrective action program as action request (AR) 02600726 and planned to perform an apparent cause evaluation to address the trend. Separate action requests were also written and immediate corrective actions were taken for each identified example to address the nonconformances created by the failure to follow procedures.
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| The inspectors determined that the failure to ensure that activities affecting quality were accomplished in accordance with the appropriate instructions, procedures and drawings as required by 10 CFR 50 Appendix B Criterion V, was a performance deficiency. Specifically, the inspectors identified two instances where the licensee failed to follow procedures resulting in impacts to safety related equipment and processes. The performance deficiency was more than minor in accordance with Inspection Manual Chapter (IMC) 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," dated September 7, 2012, because, if left uncorrected the performance deficiency had the potential to lead to a more significant safety concern. Specifically, by not performing activities affecting quality in accordance with a procedure the licensee could manipulate equipment and challenge the operators by causing unexpected transients or impact safety related equipment. Using IMC 0609, Appendix G, Shutdown Operations Significance Determination Process, Attachment 1, issued May 9, 2014, the finding was screened against the Mitigating Systems cornerstone and determined to be of very low safety significance because the finding did not represent a loss of system safety function, it did not represent an actual loss of function of a single train or two separate trains for greater than its allowed outage time, it did not represent an actual loss of safety function of one or more non-TS trains of equipment during shutdown for equipment designated as risk significant for greater than 24 hours, and it did not degrade a functional auto-isolation of residual heat removal (RHR) on low reactor vessel level. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of challenging the unknown which states, individuals stop when faced with uncertain conditions. Risks are evaluated and managed before proceeding. Contrary to this, when challenged with uncertain conditions, the licensee did not stop and properly Page 4 of 12
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| 2Q/2016 Inspection Findings - Clinton evaluate the issues before proceeding, resulting in adverse impacts to safety related equipment and activities. (H.11)
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| Inspection Report# : 2015004 (pdf)
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| Significance: Oct 09, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Generate Issue Reports for Conditions Adverse to Quality The inspectors identified a finding of very low safety significance, and an associated NCV of Title 10, Code of Federal Regulations, Part 50, Appendix B, Criterion II, Quality Assurance Program, for the failure to perform activities in accordance with procedure PI-AA-125, Corrective Action Program, Revision 2, which was a Quality Assurance Program implementing procedure. Specifically, the inspectors identified six examples where the licensee failed to generate IRs for conditions adverse to quality (CAQ) as required by PI-AA-125, until prompted by the inspectors. The licensee documented the issue in the CAP as IR 2518477, and planned on reviewing the apparent cause evaluation to determine if additional actions needed to be taken.
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| The performance deficiency was more than minor in accordance with IMC 0612, "Power Reactor Inspection Reports,"
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| Appendix B, "Issue Screening," dated September 7, 2012, because, if left uncorrected the performance deficiency had the potential to lead to a more significant safety concern. Specifically, by not identifying and documenting conditions adverse to quality the issues would not go through the screening and review process in accordance with the corrective action procedure, which could impact the identification of conditions affecting operability. The finding was screened against the Mitigating Systems cornerstone, and determined to be of very low safety significance because the it did not represent a loss of safety system or function, it did not represent an actual loss of function of a single train of two separate trains for greater than its allowed outage time and it did not represent a loss of function of a non-technical specification system designated as highly safety-significant within the licensees Maintenance Rule Program for greater than 24 hours. The inspectors determined this finding affected the cross-cutting area of problem identification and resolution in the aspect of identification where the organization implements a CAP with a threshold for identifying issues and individuals identify issues completely, accurately and in a timely manner in accordance with the program. Specifically, the licensee failed to identify issues completely, accurately and in a timely manner, causing them to not recognize issues as CAQs, and therefore not follow their process for handling these issues.
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| Inspection Report# : 2015007 (pdf)
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| Significance: Sep 30, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation FAILURE TO IMPLEMENT AND COMPLY WITH TRANSIENT EQUIPMENT/MATERIALS PROGRAM The inspectors identified a green finding and an associated NCV of 10 CFR 50, Appendix B, Criterion V Instructions, Procedures, and Drawings for the licensees failure to implement and comply with station procedure CPS 1019.05, Transient Equipment/Materials, Revision 23, to ensure that transient equipment and materials are controlled so there is no impact to safe operation of plant equipment. Specifically, on numerous occasions the inspectors identified equipment and materials improperly staged, improperly secured or in areas without engineering evaluations. The licensee documented the issue in the CAP as action requests (AR) 02507167 and AR 02529227. In each occasion identified by the inspectors the licensee subsequently removed the items identified to restore compliance with the station procedures.
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| The inspectors determined the licensees failure to implement and comply with station procedures to ensure that transient equipment and materials are controlled so there is no impact to safe operation of plant equipment was a performance deficiency. The performance deficiency was determined to be more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Screening, dated Page 5 of 12
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| 2Q/2016 Inspection Findings - Clinton September 7, 2012, because if left uncorrected it had the potential to lead to a more significant safety concern.
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| Specifically, transient equipment and material in proximity of safety related components has the potential of impacting these components during a seismic event, potentially rendering them unable to fulfill their safety function.
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| The performance deficiency is also associated with the protection against external factors attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability and capability of systems that response to initiating events to prevent undesirable consequences, and is therefore a finding.
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| Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process (SDP) for Findings at Power, issued June 19, 2012, the finding was screened against the Mitigating Systems cornerstone and determined to be of very low safety significance (Green) because the finding did not represent a loss of system or function, it did not represent an actual loss of function of at least a single train for >
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| its TS allowed outage time and it did not represent an actual loss of one or more not TS trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of field presence where leaders are commonly seen in the work areas of the plant observing, coaching, reinforcing standards and expectation. Deviations from standards and expectations are corrected promptly. Specifically, after various examples of material placement being an issue, the licensee didnt perform in field observations, caching and reinforcement of standards and expectations in the identified areas.
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| Inspection Report# : 2015003 (pdf)
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| Barrier Integrity Significance: Mar 31, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Identify a Degraded Safety-Related Support
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| . The inspectors identified a finding of very low safety significance and an associated non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for the failure to identify a condition adverse to quality.
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| Specifically, the licensee failed to identify that a safety-related support associated with control room ventilation B was degraded to the point it no longer conformed to the seismic analysis and required an evaluation to determine whether it was still capable of performing its safety function during a seismic event. This issue was entered into the licensees CAP as AR 2639317. The licensees immediate corrective actions included performing an evaluation that concluded the remaining three supports would be able to withstand the stresses imposed during a seismic event and creating an action to update the seismic calculation to incorporate the evaluation performed for the degraded support.
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| The licensee also planned to re-apply a coating to the supports as well as research and install insulation that was more breathable to minimize moisture accumulation and preclude any further degradation.
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| The inspectors determined that the failure to identify a condition adverse to quality in accordance with 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, was more than minor because if left uncorrected it had the potential to lead to a more significant safety concern. Specifically, by failing to identify the support was degraded, and correct the condition, the loss of material due to corrosion could potentially progress to the point where the remaining supports would no longer be able to perform their safety function. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings at Power, issued June 19, 2012, the finding was screened against the Barrier Integrity Cornerstone and determined to be of very low safety significance because the finding did not represent a degradation of the barrier function of the control room against radiological conditions or a smoke or toxic atmosphere. The inspectors determined this finding affected the cross-cutting area of problem identification and resolution, in the aspect of evaluation, which states, The organization Page 6 of 12
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| 2Q/2016 Inspection Findings - Clinton thoroughly evaluates issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, the licensee failed to thoroughly evaluate the issue identified by the inspectors and therefore did not recognize the degradation on the supports constituted a condition adverse to quality.
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| Inspection Report# : 2016001 (pdf)
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| Significance: Mar 31, 2016 Identified By: Self-Revealing Item Type: NCV Non-Cited Violation Failure to Assess and Manage Risk Increase for a Proposed Maintenance Activity A self-revealed finding of very low safety significance and an associated non-cited violation of 10 CFR 50.65 (a)(4) was identified on January 20, 2016, due to the licensees failure to assess and manage the risk increase from a proposed maintenance activity. Specifically, the licensee failed to manage the risk associated with racking out the continuous containment purge (CCP) A breaker, which resulted in the loss of both CCP trains, and led to an increase in primary to secondary containment differential pressure which exceeded the TS value. The licensee entered this issue into their CAP as AR 02614832. The proposed corrective actions to address this issue included creating a checklist to ensure validation of initial conditions is performed and providing training that reinforces the need to properly screen work order tasks with the appropriate risk factors.
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| The inspectors determined that the failure to assess and manage the risk increase of a proposed maintenance activity, as required by 10 CFR 50.65 (a)(4), was more than minor because it was associated with the maintenance procedure quality attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, by not properly assessing the risk of racking out the CCP A breaker the licensee did not recognize the CCP B train would be impacted, which resulted in exceeding the TS value for primary to secondary containment differential pressure. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings at Power, issued June 19, 2012, the finding was screened against the Barrier Integrity Cornerstone and determined to be of very low safety significance because the finding did not represent an actual open pathway in the physical reactor containment, containment isolation system or heat removal components and it did not involve an actual reduction in function of hydrogen igniters in the reactor containment. The inspectors identified a cross-cutting aspect in the area of human performance, in the aspect of challenging the unknown, which states, individuals stop when faced with uncertain conditions; risks are evaluated and managed before proceeding. Specifically, when the licensee was preparing the work package for maintenance on the CCP system it was uncertain what activities had already been completed as part of a concurrent evolution. Instead of stopping and validating the configuration of plant equipment, assumptions were made, and the risk of the activity was not properly assessed or managed.
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| Inspection Report# : 2016001 (pdf)
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| Significance: Feb 11, 2016 Identified By: NRC Item Type: FIN Finding Failure to Perform Adequate Evaluation of Crane and Crane Support Structure Elements A finding of very low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors for the failure of the licensees design control measures to provide for the verifying or checking the adequacy of design of the fuel handling building crane and crane support structure elements. Specifically, calculations involving the crane trolley rails, crane rail clips, and crane rail clip bolts had not been verified or checked to ensure the design basis requirements of American Society of Mechanical Engineers (ASME) NOG-1-2004; American Institute of Steel Construction (AISC), 7th Edition; and Updated Safety Page 7 of 12
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| 2Q/2016 Inspection Findings - Clinton Analysis Report (USAR) Section 3.8.4.5 were included. The licensee documented these issues in its corrective action program and initiated actions to restore compliance.
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| The performance deficiency was determined to be more than minor because if left uncorrected the performance deficiency could lead to a more significant safety concern if independent spent fuel storage installation (ISFSI) loading was conducted. The inspectors determined the finding could be evaluated using the Significance Determination Process in accordance with Inspection Manual Chapter (IMC) 0609, The Significance Determination Process for Findings At-Power, Appendix A, Exhibit 3 - Barrier Integrity Screening Questions (Section D). Based on answering No to all the questions in Exhibit 3, Section D, the inspectors determined the finding to be of very low safety significance (Green). The inspectors identified a Human Performance, Design Margin (H.6) cross-cutting aspect associated with this finding. Specifically, the licensee failed to ensure the crane trolley rails, crane rail clips, and crane rail clip bolts reflected the intended design margins of the design and licensing basis.
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| Inspection Report# : 2016010 (pdf)
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| Significance: Sep 30, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation FAILURE TO OBTAIN A LICENSE AMENDMENT PRIOR TO MAKING MODIFICATIONS TO SECONDARY CONTAINMENT The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.59(d)(1), Changes, Tests, and Experiments for the licensees failure to provide a written evaluation, which provided the basis for determining that the change to the secondary containment completed on December 18, 2014 did not require a license amendment.
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| Specifically, the licensee made a change pursuant to 10 CFR 50.59(c), to the secondary containment, and eliminated the tornado wind and tornado missile loading condition from the FB Railroad Airlock (the enclosure walls and roof) and associated outer door (1SD1-31) Seismic Category I requirements and did not provide a written evaluation to provide a basis for the determination that this change would not result in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system or component important to safety.
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| The inspectors determined that the licensees failure to provide a written evaluation, which provided the basis for determining that the change to the secondary containment completed on December 18, 2014 did not require a license amendment was a performance deficiency. Specifically, the licensee made a change pursuant to 10 CFR 50.59(c) to the secondary containment and eliminated the tornado wind and tornado missile loading condition from the FB Railroad Airlock (the enclosure walls and roof) and associated outer door and did not provide a written evaluation to provide a basis for the determination that this change would not result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety. The performance deficiency was determined to be more than minor in accordance with IMC 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," dated September 7, 2012, because it was associated with the design control attribute of the barrier integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system and containment) protect the public from radionuclide releases caused by accidents or events. In addition, the associated violation was determined to be more than minor because the inspectors could not reasonably determine if the changes to secondary containment would have required NRC prior approval. The licensee documented the issue in the CAP as action request (AR) 02534694. The licensee is complying with technical specifications anytime the inner railroad bay door is opened by entering the applicable action statements, evaluating weather conditions and impact to plant risk and establishing the necessary mitigating actions required prior to opening the door. Violations of 10 CFR 50.59 are dispositioned using the traditional enforcement process instead of the SDP because they are considered to be violations that potentially impede or impact the regulatory process. However, if possible, the underlying technical issue is evaluated under the SDP to determine the severity of the violation. In this case, the inspectors used IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings at Power, issued June 19, 2012, the finding was screened against the barrier integrity cornerstone and determined to be of very low safety significance Page 8 of 12
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| 2Q/2016 Inspection Findings - Clinton (Green) because the finding did not represent a degradation only of the radiological barrier function for the Standby Gas Treatment (SBGT) system nor did it represent a degradation of the function of the control room against smoke or toxic atmosphere. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of procedure adherence where individuals follow processes, procedures and work instructions. Specifically, the licensee failed to follow the 50.59 regulatory process as defined in station procedure LS-AA-104-1000, 50.59 Resource Manual, Revision 9.
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| Inspection Report# : 2015003 (pdf)
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| Significance: Sep 30, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation FAILURE TO ENTER APPROPRIATE TS ACTION STATEMENT FOR INOPERABLE RADIATION MONITORS DURING OPDRV ACTIVITIES The inspectors identified a green finding and associated NCV of T.S. 3.3.6.1 Primary Containment and Drywell Isolation Instrumentation and 3.3.6.2 Secondary Containment Isolation Instrumentation, for the failure to enter the appropriate action statement and take the associated actions related to inoperable containment radiation monitor instrumentation during operations with the potential to drain the reactor vessel. Specifically, with the containment ventilation dampers closed, the containment radiation monitor instrumentation would not be able to perform its safety function of sending a containment isolation signal for elevated containment radiation levels as required during OPDRVS. At the time of discovery the licensee had already concluded OPDRV activities and was therefore no longer in a mode of applicability. The licensee documented the issue in the CAP as action request (AR) 2566708. When this issue was identified the maintenance on the VR/VQ system was complete and no OPDRVs were in progress, therefor the T.S. noncompliance was no longer in effect.
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| The inspectors determined that the failure to enter T.S. 3.3.6.1 and 3.3.6.2 when the radiation monitor instrumentation was not able to perform its safety function during an OPDRV, was a performance deficiency. Specifically, the licensee failed to recognize that when the containment ventilation dampers were closed, the radiation monitors could not detect the radiation levels in primary containment and therefore could not fulfill their safety function of sending containment isolation signals in the case of elevated radiation levels in containment. The performance deficiency was more than minor in accordance with IMC 0612, Power Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, because, it was associated with the SSC and Barrier Performance attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events, and is therefore a finding.
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| Specifically, the automatic containment isolation signal function of the radiation monitors was impacted when the containment ventilation dampers were closed during OPDRV operations. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings, dated May 9, 2014, the finding was screened against the Barrier Integrity cornerstone and determined to need a detailed risk evaluation because the finding represents a degradation of the ability to close or isolate the containment. Using Appendix G Exhibit 4, Barrier Integrity Screening Questions, the Senior Reactor Analyst (SRA) determined that the finding degraded the ability to close or isolate the containment per Section B, Containment Barrier, Question 6. Therefore, the evaluation was continued using IMC 0609 Appendix H, Containment Integrity Significance Determination Process. The SRA determined this to be a Type B finding, because it was related to a degraded condition that had implications for containment integrity without affecting the likelihood of core damage. The SRA used Section 6.2 of Appendix H, Approach for Assessing Type B Findings at Shutdown. Based on information from the inspectors, during all OPDRV time windows, the reactor water level was confirmed to be greater than the minimum level required for movement of irradiated fuel assemblies (i.e., greater than 228 above the flange). This plant condition meets the definition of Plant Operating State 3 (POS 3) of Appendix H. Therefore, based on the plant being in POS 3 during the OPDRV time windows, the finding screens as Green based on Step 2.1 of Section 6.2 of Appendix H. The Page 9 of 12
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| 2Q/2016 Inspection Findings - Clinton inspectors determined this finding affected the cross-cutting area of human performance in the aspect of conservative bias where individuals use decision making practices that emphasize prudent choices over those that are simple allowable. A proposed action is determined to be safe in order to proceed, rather than unsafe in order to stop.
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| Specifically, the licensee relied solely on the successful completion of the surveillance requirements to determine the radiation monitor instrumentation was operable rather than considering the impact the closed dampers would have on their ability to fulfill their safety function.
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| Inspection Report# : 2015003 (pdf)
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| Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Security Although the Security Cornerstone is included in the Reactor Oversight Process assessment program, the Commission has decided that specific information related to findings and performance indicators pertaining to the Security Cornerstone will not be publicly available to ensure that security information is not provided to a possible adversary.
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| Other than the fact that a finding or performance indicator is Green or Greater-Than-Green, security related information will not be displayed on the public web page. Therefore, the cover letters to security inspection reports may be viewed.
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| Miscellaneous Significance: N/A Mar 31, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Report a Condition that Could Have Prevented Fulfillment of a Safety Function The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.72(b)(3)(v) for failing to report an event or condition, that at the time of discovery could have prevented the fulfillment of a safety function, to the NRC within eight hours. Specifically, control room operators placed both divisions of reactor water cleanup differential flow instruments in bypass, which rendered the instruments inoperable and resulted in a loss of the isolation function.
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| The licensee entered this issue into the CAP as AR 02645140 and created an action to submit an licensee event report under 10 CFR 50.73(a)(2)(v).
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| The inspectors determined that the failure to report an event or condition, that at the time of discovery could have Page 10 of 12
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| 2Q/2016 Inspection Findings - Clinton prevented the fulfillment of a safety function, to the NRC within 8 hours as required by 10 CFR 50.72(b)(3)(v) was a performance deficiency. The inspectors reviewed this issue in accordance with IMC 0612 and the Enforcement Manual. Violations of 10 CFR 50.72 are dispositioned using the traditional enforcement process because they are considered to be violations that potentially impede or impact the regulatory process. The inspectors reviewed Section 6.9.d.9 of the NRC Enforcement Policy and determined this violation was Severity Level IV because the licensees failure to make the report, as required by 10 CFR 50.72, did not cause the NRC to reconsider a regulatory position or undertake substantial further inquiry. No cross-cutting aspect was assigned because cross-cutting aspects are not assigned to traditional enforcement only violations.
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| Inspection Report# : 2016001 (pdf)
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| Significance: N/A Mar 31, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Report Condition Prohibited by Technical Specifications The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.73(a)(2)(i)(B) for failing to report to the NRC, within 60 days of discovery, a condition prohibited by the plants TS. Specifically, the licensee failed to notify the NRC of two instances where they failed to comply with TS 3.3.6.1 and TS 3.3.6.2 and enter the limiting condition for operation action statements when required. The licensee entered this issue into their CAP as AR 02619114 and subsequently issued a licensee event report on March 16, 2016.
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| The inspectors determined that the failure to report a condition prohibited by the plants TS as required by 10 CFR 50.73(a)(2)(i)(B), within 60 days of discovery, was a performance deficiency. The inspectors reviewed this issue in accordance with IMC 0612 and the Enforcement Manual. Violations of 10 CFR 50.73 are dispositioned using the traditional enforcement process because they are considered to be violations that potentially impede or impact the regulatory process. The inspectors reviewed Section 6.9.d.9 of the NRC Enforcement Policy and determined this violation was Severity Level IV because the licensees failure to make the report, as required by 10 CFR 50.73, did not cause the NRC to reconsider a regulatory position or undertake substantial further inquiry. No cross-cutting aspect was assigned because cross-cutting aspects are not assigned to traditional enforcement only violations.
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| Inspection Report# : 2016001 (pdf)
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| Significance: N/A Dec 31, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation FAILURE TO UPDATE THE FINAL SAFETY ANALYSIS REPORT (FSAR) - HYDROGEN WATER CHEMISTRY SYSTEM The inspectors identified a Severity Level IV Violation of Title 10 Code of Federal Regulations (CFR) 50.71(e),
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| "Periodic Update of the FSAR", for the licensee's failure to update the FSAR after installing a hydrogen water chemistry system into the plant to reduce rates of intergranular stress corrosion cracking (IGSCC) in recirculation piping and reactor vessel internals. Specifically, the licensee did not update Section 5.4.15, "Hydrogen Water Chemistry System" of the FSAR to include a design basis and description of process and system used to periodically injection noble metals. The licensee entered this issue into the corrective action program as AR 02594259 and is revising the FSAR include additional the design basis and additional system description for noble metal injection.
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| The inspectors determined that the failure to update the FSAR in accordance with 10 CFR 50.71(e), "Periodic Update of the FSAR", with the design basis and description of the process and system used to periodically injection noble metals was a performance deficiency warranting a significance evaluation. The inspectors reviewed this issue in accrodance with NRC inspection manual chapter 0612 and the NRC enforcement manual. Violations of 10 CFR 50.71 (e) are dispositioned using the traditional enforcement process because they are considered to be violations that potentially impede or impact the regulatory process. The inspectors reviewed section 6.1.d.3 of the NRC Enforcement Page 11 of 12
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| 2Q/2016 Inspection Findings - Clinton Policy and determined this violation was Severity :Level IV because the licensee's failure to update the FSAR as required by 10 CFR 50.71(e) had not yet resulted in any unacceptable change to the facility or procedures. No cross cutting aspect was assigned because cross cutting aspects are not assigned to traditional enforcement only violations.
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| Inspection Report# : 2015004 (pdf)
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| Last modified : August 29, 2016 Page 12 of 12
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| 3Q/2016 Inspection Findings - Clinton Clinton 3Q/2016 Plant Inspection Findings Initiating Events Significance: Sep 30, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Prevent Recurrence of a Significant Condition Adverse to Quality The inspectors documented a self-revealing finding of very low safety significance and an NCV of 10 CFR 50, Appendix B, Criterion XVI for the licensee's failure to take corrective action to preclude repetition of a significant condition adverse to quality (SCAQ). After identifying IGSCC on main steam flex hoses in 2007 and concluding the leakage constituted a SCAQ, the licensee's corrective actions to prevent recurrence failed to prevent pressure boundary leakage at the same location in 2016. The licensee entered this issue into their corrective action program as AR 02670593. The affected hoses were replaced. The licensee is also developing a design change to address at least one of the three factors that contributes to IGSCC.
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| The inspectors determined that the licensee's failure to take corrective actions to prevent recurrence of an SCAQ was a performance deficiency and more than minor because it was associated with the equipment performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown and at power operations. The finding was screened as low safety significant because it did not result in exceeding the RCS leak rate for a small LOCA and did not affect systems used to mitigate a LOCA.
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| Inspection Report# : 2016003 (pdf)
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| Significance: Jun 30, 2016 Identified By: NRC Item Type: FIN Finding Material Unsecure in the Secured Material Zone The inspectors identified a finding of very low safety significance for the failure to ensure material placed within the transformer secured material zone, was secured as required by station procedure MA-AA-716-026, Station Housekeeping/Material Condition Program, Revision 14. Specifically, the inspectors identified unsecured scaffold poles and knuckles within the licensee established secure material zone. The licensee has entered this issue into their corrective action program (CAP) as action request AR 02668245. The material was immediately removed from the secured zone by the licensee.
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| The inspectors determined the licensees failure to ensure material within the secured material zone was adequately secured in accordance with procedure MA-AA-716-026, Station Housekeeping/Material Condition Program, was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations, and is therefore a finding. Specifically, by not securing material in the vicinity of main power transformers, the material could become a missile and impact the transformers causing a potential reactor SCRAM. The finding was screened against the Initiating Events Cornerstone and determined to be of very low safety significance (Green) because the finding did not involve the complete or partial loss of a support system that contributes to the likelihood of, or cause an initiating event and did not affect mitigation equipment. The inspectors determined that this finding Page 1 of 17
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| 3Q/2016 Inspection Findings - Clinton had a cross-cutting aspect in the area of human performance in the aspect of field presence, where leaders are commonly seen in the work areas of the plant observing, coaching, and reinforcing standards and expectations.
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| Specifically, since initial identification of the issue the inspectors have noted that while discussions on when to perform walkdowns took place, the supervisors or managers did not ensure sufficient field presence to reinforce the standards and expectations, leading to material continuing to be easily found by the inspectors. (H.2)
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| Inspection Report# : 2016002 (pdf)
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| Significance: Jun 30, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Perform Surface Examination Prior to Weld Repair of a Reactor Water Cleanup System Pipe The inspectors identified a finding of very-low safety significance and associated NCV of 10 CFR 50.55a(g)(4).
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| Specifically, the licensee failed to perform a surface examination to detect cracking on reactor water cleanup small-bore piping prior to performing a weld repair. The license documented the issue in the CAP as AR 02671726 and AR 02685332 and performed an operability review. The licensee has prepared a work order to perform a surface exam of the existing weld and surrounding area.
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| The inspectors determined that the failure to perform the surface examination prior to weld repair of RWCU pipe 1G33C001B as required by 10 CFR 50.55a(g)(4) was a performance deficiency. The inspectors determined that this issue was more-than-minor in accordance with IMC 0612, Appendix B, because it adversely affect the Initiating Events Cornerstone attribute of barrier integrity and because the answer to the question of If left uncorrected, would the performance deficiency have the potential to lead to a more significant safety concern? was yes. Specifically, the lack of a surface exam may result in the entire defect not being removed during the repair and the potential existed for a cracked pipe to remain in service. This could lead to a repeat leak of reactor coolant. The inspectors determined this finding was of very-low safety significance (Green) based on answering no to Question A.1 and A.2 of the Exhibit 1, Initiating Events Screening Questions, in IMC 0609, Attachment A, The Significance Determination Process (SDP) for Findings At-Power. Specifically, the inspectors answered no to the screening question associated with a reactor coolant system leak exceeding the leak rate for a small loss of coolant accident (LOCA) and no to the screening question associated with systems used to mitigate a LOCA. A subsequent visual examination of the weld repair revealed an absence of cracking and the licensee also planned to perform a follow-up surface examination of the repaired area to look for cracking. The inspectors determined that this finding had a cross-cutting aspect in the area of human performance in the aspect of resources, where leaders ensure that personnel, equipment, procedures and other station resources are available and adequate to support nuclear safety. Specifically, the work order that performed the work did not specify a surface examination of the base metal prior to welding. (H.1)
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| Inspection Report# : 2016002 (pdf)
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| Significance: Jun 30, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Obtain License Amendment prior to Operating Reactor Water Cleanup Bypass Switches The inspectors identified a Severity Level IV NCV of 10 CFR 50.59(a)(1), Changes, Tests, and Experiments, and an associated Green finding for the licensees failure to perform an adequate written safety evaluation to provide the basis that changes to the Updated Safety Analysis Report (USAR) and station procedures did not involve an unreviewed safety question. Specifically, the licensee changed the USAR and station procedures to allow operators to a defeat the safety function of the reactor water cleanup (RWCU) isolation valves to prevent unwarranted isolation signals during normal operation without obtaining prior Commission approval. The licensee entered this issue into their CAP as AR 02685337 and will be changing station procedures to prevent placing the RWCU leak detection divisional bypass switches in bypass except for instrument channel maintenance, testing or calibration.
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| The inspectors determined that the licensees failure to perform an adequate written safety evaluation to provide the Page 2 of 17
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| 3Q/2016 Inspection Findings - Clinton basis that changes to the USAR and station procedures did not involve an unreviewed safety question was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the equipment performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The finding was screened against the Initiating Events cornerstone and determined to be of very low safety significance (Green) because the finding did not result in exceeding the RCS leak rate for a small LOCA and did not affect other systems used to mitigate a LOCA resulting in a total loss of their function (e.g. Interfacing System LOCA). No cross cutting aspect was assigned because the inspectors determined the performance deficiency was not indicative of current plant performance. Violations of 10 CFR 50.59 are dispositioned using the traditional enforcement process because they are considered to be violations that potentially impede or impact the regulatory process. The inspectors reviewed Section 6.1.d.2 of the NRC Enforcement Policy and determined this violation was Severity Level IV because the resulting changes were evaluated by the SDP as having very low safety significance.
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| Inspection Report# : 2016002 (pdf)
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| Significance: Dec 31, 2015 Identified By: NRC Item Type: FIN Finding Failure to Follow Station Procedures for Plant Activities The inspectors identified a finding of very low safety significance for the failure to ensure that activities were accomplished in accordance with prescribed procedures as required by station procedure HU-AA-104-101 Procedure Use and Adherence. Specifically, the inspectors identified two examples where the licensee failed to adhere to prescribed station procedures when performing activities in the plant. The licensee placed both issues in their corrective action program as AR 02600726 and addressed the nonconformances created by the failure to follow the procedures. The licensee planned to perform an apparent cause evaluation to determine why there was an adverse trend related to procedure adherence.
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| The inspectors determined that the failure to perform activities in accordance with prescribed procedures as required by station procedure HU-AA-104-101, Procedure Use and Adherence, was a performance deficiency. Specifically, the inspectors identified two instances where the licensee failed to follow procedures when performing activities in the plant. The performance deficiency was more than minor in accordance with IMC 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," dated September 7, 2012, because, if left uncorrected the performance deficiency had the potential to lead to a more significant safety concern. Specifically, by not performing activities in accordance with a procedure the licensee could manipulate equipment and challenge the operators, and cause unexpected transients. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process (SDP) for Findings at Power, issued June 19, 2012, the finding was screened against the Initiating Events cornerstone and determined to be of very low safety significance because the finding did not cause a reactor trip or the loss of mitigation equipment and it did not involve the complete or partial loss of a support system that contributes to the likelihood of, or cause, an initiating event. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of challenging the unknown which stated, individuals stop when faced with uncertain conditions. Risks are evaluated and managed before proceeding. Contrary to this, when challenged with unknown conditions, the licensee did not stop and properly evaluate the issues before proceeding, resulting in adverse impacts to station equipment. (H.11)
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| Inspection Report# : 2015004 (pdf)
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| Mitigating Systems Page 3 of 17
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| 3Q/2016 Inspection Findings - Clinton Significance: Sep 30, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Exceeded Technical Specification Allowed Outage Time for Main Turbine Bypass System The inspectors identified a finding of very low safety significance and an associated NCV of Technical Specification 3.7.6, Main Turbine Bypass System for the licensees failure to meet the limiting conditions for operation and complete the associated required actions after making a deficient change to the turbine bypass valve surveillance testing frequency. Specifically, with the main turbine bypass system inoperable and without the Core Operating Limits Report (COLR) limits for thermal power, minimum critical power ratio (MCPR), and linear heat generation rate (LGHR) with the main turbine by pass system inoperable applied, thermal power was not reduced to less than 21.6 percent of rated thermal power within six hours. The licensee entered this issue into their corrective action program as AR 02690657. The licensee restored compliance by applying the COLR limits for reactor thermal power, MCPR and LGHR.
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| The inspectors determined the failure to meet the limiting conditions for operation and complete the associated required actions prior to the end of the specified completion times was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was screened against the Mitigating Systems cornerstone and determined to be of very low safety significance because all of the associated questions in IMC 0609, Appendix A, were answered no. The inspectors determined this finding affected the cross-cutting area of human performance, in the aspect of change management, where leaders use a systematic process for evaluating and implementing change so that nuclear safety remains the overriding priority because the licensees change management process was not fully utilized by senior management when evaluating and implementing a change to the turbine bypass valve surveillance testing frequency. (H.3)
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| Inspection Report# : 2016003 (pdf)
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| Significance: Sep 30, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Perform a 50.59 Screening for Changing the Frequency of Exercising the Turbine Bypass Valves The inspectors identified a Severity Level IV NCV of 10 CFR 50.59 4(d)(1), Changes, Tests, and Experiments, and an associated Green finding for the licensees failure to perform a written evaluation which provided the bases for determining that changing the turbine bypass valve surveillance testing frequency from every 31 days, as specified in the Updated Safety Analysis Report, to once a year did not require a license amendment. The licensee has entered this issue into their corrective action program as AR 02720163. The licensee is currently evaluating the issue in accordance with their procedure for changes to the facility.
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| The inspectors determined that the licensees failure to perform a written evaluation to provide the basis for the determination that a change to the facility, a change to a procedure, or a change to a test or experiment did not require a license amendment was a performance deficiency. The performance deficiency was more than minor in accordance with IMC 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," dated September 7, 2012, because, it was associated with the procedure quality attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2012, the finding was screened against the Mitigating Systems cornerstone and determined to be of very low safety significance because all of the associated questions in IMC 0609, Appendix A, were answered no. Violations of 10 Page 4 of 17
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| 3Q/2016 Inspection Findings - Clinton CFR 50.59 are dispositioned using the traditional enforcement process because they are considered to be violations that potentially impede or impact the regulatory process. The inspectors reviewed Section 6.1.d.2 of the NRC Enforcement Policy and determined this violation was Severity Level IV because the resulting changes were evaluated by the SDP as having very low safety significance. The inspectors determined this finding affected the cross-cutting area of human performance, in the aspect of consistent process, where individuals use a consistent, systematic approach to make decisions. The licensee made a decision to proceed with implementation of a change to the turbine bypass valve surveillance testing frequency after a plant oversight committee review in lieu of following their consistent, systematic process for evaluating changes to the USAR. (H.13)
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| Inspection Report# : 2016003 (pdf)
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| Significance: Aug 09, 2016 Identified By: NRC Item Type: FIN Finding Failure to have hose configurations that were verified to be able to ensure a timely and successful implementation of a FLEX strategy Green. Two examples of a finding of very low safety significance was identified by the inspectors for the licensees failure to have hose configurations that were verified to be able to ensure a timely and successful implementation of a flexible response (FLEX) strategy. Specifically, the licensee did not ensure through evaluations, calculations, analyses or any other means that the strategy for maintaining core cooling, containment heat removal and Spent Fuel Pool (SFP) cooling during a Beyond-Design-Basis External Event (BDBEE) flooding scenario would be capable of fulfilling its function. No violation of NRC requirements were identified.
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| The performance deficiency is more than minor because it was associated with the mitigating systems cornerstone objective attribute of protection against external factors, specifically the BDBEE flood hazard, and it adversely affected the cornerstone attribute of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Issues identified through TI-191 are evaluated through a cross-regional panel using IMC 0609, Appendix M, Significance Determination Process Using Qualitative Criteria, as informed by draft Appendix O, Post Fukushima Mitigation Strategies Significance Determination Process. The finding was determined to be of very low safety significance (Green). The inspectors concluded that the cause of the finding involved a cross-cutting component in the Human Performance area of Design Margins because the organization did not ensure the selected strategy contained the required verification that it could be successfully implemented. [H.6]
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| Inspection Report# : 2016007 (pdf)
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| Significance: Jun 30, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Inspection Fails to Identify Safety Related Cables Submerged in Cable Vault The inspectors identified a finding of very low safety significance (Green) and associated NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action Program, for the failure to identify a condition adverse to quality.
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| Specifically, the licensee failed to identify that portions of the Division 1 SX safety related cables, which are not rated for submergence, were under water. The licensee entered this issue into their corrective action program as action requests AR 02648804 and AR 02648507. Operators took actions to pump out the water to ensure the cables were returned to a dry condition.
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| The inspectors determined the licensees failure to identify a condition adverse to quality was contrary to 10 CFR 50, Appendix B, Criterion XVI, Corrective Action Program, and was a performance deficiency. The performance deficiency was determined to be more than minor because the finding was associated with the Mitigating Systems Page 5 of 17
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| 3Q/2016 Inspection Findings - Clinton Cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, because the SX cables are not rated for submergence, they could degrade and affect the reliability of the SX system. The finding was screened against the Mitigating Systems cornerstone and determined to be of very low safety significance (Green) because the inspectors answer Yes to the question does the SSC maintain its operability or functionality. Specifically, the SX system submerged cables did not cause the SX system to be inoperable or nonfunctional. The inspectors determined that this finding had a cross-cutting aspect in the area of human performance in the aspect of resources, where leaders ensure that personnel, equipment and other resources are available and adequate to support nuclear safety. Specifically, the individuals performing the inspection did not have the necessary resources, such as training, procedures, drawings or a detailed pre-job brief, to identify the cables sloped downwards in the cable vault and were submerged in the water. (H.1)
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| Inspection Report# : 2016002 (pdf)
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| Significance: Jun 30, 2016 Identified By: NRC Item Type: FIN Finding Failure to Properly Install Cable Vault Mitigating Equipment The inspectors identified a finding of very low safety significance (Green) for the failure to incorporate human performance standards when developing work package instructions in accordance with MA-AA-716-010,Maintenance Planning, Revision 23. Specifically, the licensee did not assure the cable vault dewatering system installation and maintenance work order (WO) included the appropriate details to troubleshoot and install the cable vault sump pumps and float switches. This resulted in installation of the equipment in a manner that prevented detection and removal of water from the cable vaults, allowing cables to remain submerged undetected. The licensee entered this issue into their CAP as AR 02668245. The corrective actions performed by the licensee included placing the sump pumps in the right location and adjusting the float switches to ensure the indications would alert operators when the vaults needed to be pumped.
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| The inspectors determined that the failure to incorporate human performance standards when developing work package instructions in accordance with MA-AA-716-010, Maintenance Planning, Revision 23, was a performance deficiency. The performance deficiency was determined to be more than minor because if left uncorrected the performance had the potential to lead to a more significant safety concern. Specifically, by not appropriately installing the sump pumps and float switches, the cables would be allowed to remain submerged undetected. The finding was screened against the Mitigating Systems cornerstone and determined to be of very low safety significance (Green) because the inspectors answer Yes to the question does the SSC maintain its operability or functionality?. This finding has a cross-cutting aspect in the area of human performance in the aspect of conservative bias, where individuals use decision making practices that emphasize prudent choices over those that are simply allowable. Specifically, because the licensee classified the cable vault dewatering system as a maintenance tool, they decided it was not necessary to include specific instructions within the WOs related to ensure the troubleshooting and re-installation activities were performed appropriately. (H.14)
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| Inspection Report# : 2016002 (pdf)
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| Significance: Jun 30, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Inadequate Ultrasonic Examination Method Used to Detect Crack-Like Flaws On May 17, 2016, the inspectors identified a finding of very-low safety significance and associated NCV of 10 CFR 50, Appendix B, Criterion IX, Control of Special Processes, for the licensees failure to ensure that nondestructive testing was controlled and accomplished using qualified procedures in accordance with applicable codes and standards. Specifically, the licensee did not implement an angle beam ultrasonic (UT) examination to detect cracking in a degraded SX pipe prior to implementation of a weld overlay repair. The licensee subsequently performed the required UT examination to confirm the absence of cracks and documented the issue in the CAP in AR 02671724.
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| Page 6 of 17
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| 3Q/2016 Inspection Findings - Clinton The inspectors determined that this finding was more than minor because if left uncorrected, the failure to perform the UT would become a more significant safety concern. Specifically, if left uncorrected, the use of an unqualified UT examination for detection of cracks could result undetected cracks that propagate to failure during service. The inspectors determined this finding was of very low safety significance (Green) based on answering yes to the questions in Part A of Exhibit 2, Mitigating Systems Screening Questions, in IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Specifically, the inspectors answered yes to the screening question If the finding is a deficiency affecting the design or qualification of a mitigating SSC [structures, systems, or components], does the SSC maintain its operability or functionality? because the licensee subsequently performed appropriate UT examination to confirm that cracks were not present. The finding had a cross-cutting aspect in the area of Human Performance for Procedure Adherence, because the licensee failed to follow processes, procedures, and work instructions to ensure that the appropriate UT examination was applied to the degraded SX pipe.
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| (H.8)
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| Inspection Report# : 2016002 (pdf)
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| Significance: Mar 31, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Operability Determination Failed to Examine Test Failures The inspectors identified a finding of very low safety significance and an associated non-cited violation of 10, Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion V, Instructions Procedures and Drawings, for the failure to follow Station Procedure OP-AA-108-115, Operability Determinations, Revision 16. Specifically, after valve 1SX027C, a valve required for residual heat removal operability, failed a surveillance test, the licensee did not base the operability determination on a detailed examination of the deficiency and did not document a basis for why a reasonable expectation of operability existed. The licensee entered this issue into their corrective action program (CAP) as Action Request (AR) 02553168 and AR 02558101. The licensee revised the in-service testing program evaluation for valve 1SX027C and documented additional details to support declaring the valve operable.
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| The inspectors determined the failure to follow Station Procedure OP-AA-108-115 was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to correctly perform an operability evaluation for valve 1SX027C had the potential to allow an inoperable condition to go undetected. Using IMC 0609, , Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2012, the finding was screened against the Mitigating Systems Cornerstone and determined to be of very low safety significance because the finding: was not a deficiency affecting the design or qualification of a mitigating system; did not represent a loss of system and/or function; did not represent an actual loss of function of a single train for greater than its Technical Specification (TS) allowed outage time; and did not represent an actual loss of function of one or more non-TS trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. The inspectors determined this finding affected the cross-cutting area of human performance, in the aspect of resources, where leaders ensure that personnel, equipment, procedures, and other resources are available and adequate to support nuclear safety.
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| Specifically, Station Procedure CPS 9053.04, provided guidance that the valve could remain operable for 96 hours without providing an appropriate basis.
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| Inspection Report# : 2016001 (pdf)
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| Significance: Mar 31, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Page 7 of 17
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| 3Q/2016 Inspection Findings - Clinton Inadequate Extent of Condition Associate with an ACE The inspectors identified a finding of very low safety significance and an associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion II, Quality Assurance Program, for the failure to follow a Quality Assurance Program implementing procedure. Specifically, the licensee failed to perform an adequate extent of condition review as required by PI-AA-125, Corrective Action Program, while evaluating a lack of proficiency in applying the licensing basis for structures, systems and components (SSCs) when implementing the 50.59 process. The licensee documented this issue in their CAP as AR 02641397. Immediate corrective actions included a review of the extent of condition performed by the engineering department and a recommended action of expanding the scope of the review to include additional 50.59 evaluations.
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| The inspectors determined the failure to follow a Quality Assurance Program implementing procedure was more than minor because if left uncorrected it had the potential to lead to a more significant safety concern. Specifically, if the extent of condition review is too narrowly assessed there is the potential for other safety significant systems to have been impacted by a lack of proficiency in applying the licensing basis. As a result, the SSCs may not perform their intended safety function as defined in the Updated Safety Analysis Report. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings at Power, issued June 19, 2012, the finding was screened against all cornerstones and determined to be of very low safety significance because there was no reasonable indication that the criteria in Appendix A were met. The inspectors determined this finding affected the cross-cutting area of human performance, in the aspect of procedure adherence, where individuals follow processes, procedures and work instructions. Specifically, the licensee did not effectively adhere to all available portions of CAP procedures, which led to a narrowly focused extent of condition.
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| Inspection Report# : 2016001 (pdf)
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| Significance: Feb 04, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Perform and Adequate Equipment Apparent Cause Evaluation (Section 4OA4)
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| The inspectors identified a finding of very-low safety significance (Green), and an associated Non-Cited Violation of Title 10, Code of Federal Regulations, Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to follow Step 4.3.4 of procedure PI-AA-125, Corrective Action Program Procedure.
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| Specifically, the licensee failed to perform Class B Equipment Apparent Cause Evaluation (EACE) 2381871, 1SX01PC Failed to Start for Testing, in accordance with PI-AA-125-1003, Apparent Cause Evaluation Manual, because they: (1) failed to analyze each causal factor to determine contributing causes as required by Step 4.4.1.2; and (2) failed to assign an effectiveness review for the EACE as required by Step 4.4.9.1. The licensee entered this finding into their Corrective Action Program and revised their EACE to: (1) include three contributing causes; (2) upgrade a corrective action to a corrective action to prevent recurrence; and (3) assign an effectiveness review to determine the effectiveness of the corrective action to prevent recurrence.
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| The performance deficiency was determined to be more than minor because if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, an effectiveness review is required to provide assurance that the Division 3 SX pump design change is successful in preventing recurrence of pump failure before another pump failure occurs, which would be a more significant safety concern. The finding impacted the Mitigating Systems Cornerstone and screened as having very-low safety significance (Green) because although the finding is a deficiency ultimately affecting the design or qualification of the Division 3 SX pump, the pump still maintains its operability. The inspectors determined this finding had an associated cross-cutting aspect in the area of Human Performance (Conservative Bias) because although a B Apparent Cause Evaluation may have been allowable for investigating the failure of the Division 3 SX pump, had an A Root Cause Analysis been performed, a more rigorous investigation process would have been used to identify contributing causes, assign corrective actions, and identify effectiveness reviews for the failure of the Division 3 SX pump. [H.14] (Section 4OA4.02.03.f)
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| Page 8 of 17
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| 3Q/2016 Inspection Findings - Clinton Inspection Report# : 2016008 (pdf)
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| Significance: Dec 31, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Perform Activities Affecting Quality in Accordance with Prescribed Procedures The inspectors identified a finding of very low safety significance and an associated Non-Cited Violation of 10 CFR50, Appendix B, Criterion V, Instructions Procedures and Drawings, for the failure to ensure that activities affecting quality were accomplished in accordance with the appropriate instructions, procedures and drawings.
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| Specifically, the inspectors identified two examples where the licensee failed to perform activities affecting quality in accordance with prescribed procedures. The licensee entered this issue into their corrective action program as action request (AR) 02600726 and planned to perform an apparent cause evaluation to address the trend. Separate action requests were also written and immediate corrective actions were taken for each identified example to address the nonconformances created by the failure to follow procedures.
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| The inspectors determined that the failure to ensure that activities affecting quality were accomplished in accordance with the appropriate instructions, procedures and drawings as required by 10 CFR 50 Appendix B Criterion V, was a performance deficiency. Specifically, the inspectors identified two instances where the licensee failed to follow procedures resulting in impacts to safety related equipment and processes. The performance deficiency was more than minor in accordance with Inspection Manual Chapter (IMC) 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," dated September 7, 2012, because, if left uncorrected the performance deficiency had the potential to lead to a more significant safety concern. Specifically, by not performing activities affecting quality in accordance with a procedure the licensee could manipulate equipment and challenge the operators by causing unexpected transients or impact safety related equipment. Using IMC 0609, Appendix G, Shutdown Operations Significance Determination Process, Attachment 1, issued May 9, 2014, the finding was screened against the Mitigating Systems cornerstone and determined to be of very low safety significance because the finding did not represent a loss of system safety function, it did not represent an actual loss of function of a single train or two separate trains for greater than its allowed outage time, it did not represent an actual loss of safety function of one or more non-TS trains of equipment during shutdown for equipment designated as risk significant for greater than 24 hours, and it did not degrade a functional auto-isolation of residual heat removal (RHR) on low reactor vessel level. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of challenging the unknown which states, individuals stop when faced with uncertain conditions. Risks are evaluated and managed before proceeding. Contrary to this, when challenged with uncertain conditions, the licensee did not stop and properly evaluate the issues before proceeding, resulting in adverse impacts to safety related equipment and activities. (H.11)
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| Inspection Report# : 2015004 (pdf)
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| Significance: Oct 09, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Generate Issue Reports for Conditions Adverse to Quality The inspectors identified a finding of very low safety significance, and an associated NCV of Title 10, Code of Federal Regulations, Part 50, Appendix B, Criterion II, Quality Assurance Program, for the failure to perform activities in accordance with procedure PI-AA-125, Corrective Action Program, Revision 2, which was a Quality Assurance Program implementing procedure. Specifically, the inspectors identified six examples where the licensee failed to generate IRs for conditions adverse to quality (CAQ) as required by PI-AA-125, until prompted by the inspectors. The licensee documented the issue in the CAP as IR 2518477, and planned on reviewing the apparent cause evaluation to determine if additional actions needed to be taken.
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| Page 9 of 17
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| 3Q/2016 Inspection Findings - Clinton The performance deficiency was more than minor in accordance with IMC 0612, "Power Reactor Inspection Reports,"
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| Appendix B, "Issue Screening," dated September 7, 2012, because, if left uncorrected the performance deficiency had the potential to lead to a more significant safety concern. Specifically, by not identifying and documenting conditions adverse to quality the issues would not go through the screening and review process in accordance with the corrective action procedure, which could impact the identification of conditions affecting operability. The finding was screened against the Mitigating Systems cornerstone, and determined to be of very low safety significance because the it did not represent a loss of safety system or function, it did not represent an actual loss of function of a single train of two separate trains for greater than its allowed outage time and it did not represent a loss of function of a non-technical specification system designated as highly safety-significant within the licensees Maintenance Rule Program for greater than 24 hours. The inspectors determined this finding affected the cross-cutting area of problem identification and resolution in the aspect of identification where the organization implements a CAP with a threshold for identifying issues and individuals identify issues completely, accurately and in a timely manner in accordance with the program. Specifically, the licensee failed to identify issues completely, accurately and in a timely manner, causing them to not recognize issues as CAQs, and therefore not follow their process for handling these issues.
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| Inspection Report# : 2015007 (pdf)
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| Barrier Integrity Significance: Sep 30, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Scope Fuel Building Ventilation Pressure Control into Maintenance Rule The inspectors identified a finding of very low safety significance and an NCV of 10 CFR 50.65 (b) for the licensees failure to scope a non-safety related structure, system and component (SSC), whose function is used in one or more Emergency Operating Procedures (EOP) and whose failure could cause actuation of a safety-related system, into maintenance rule. Specifically, the licensee failed to scope the non-safety related fuel building pressure control function into their maintenance rule program. The licensee has entered this issue into their corrective action program as AR 02716300. The licensee is scoping the pressure control function of fuel building ventilation into maintenance rule.
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| The inspectors determined that the licensees failure to scope a non-safety related system whose function is used in one or more EOPs and whose failure caused the actuation of a safety-related system into maintenance rule was a performance deficiency. The performance deficiency was determined to be more than minor because it affects the SSC and barrier performance attribute of the Barrier Integrity cornerstone and adversely affects the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. The finding was determined to be of very low safety significance because the inspectors answered yes to the question does the finding only represent a degradation of the radiological barrier function provided for the control room, or auxiliary building, or spent fuel pool, SBGT system (BWR)?. The inspectors determined this finding affected the cross-cutting area of human performance, in the aspect of avoiding complacency, because the licensee identified water intrusion of the fuel building pressure sensing line was a longstanding, latent, known problem and failed to recognize and appropriately challenge how the function was scoped into maintenance rule. (H.12)
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| Inspection Report# : 2016003 (pdf)
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| Page 10 of 17
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| 3Q/2016 Inspection Findings - Clinton Significance: Sep 30, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Spent Fuel Pool Liner Design not Verified per Code The inspectors identified a finding of very low safety significance and a NCV of 10 CFR Part 50, Appendix B, Criterion III for the failure of the licensee's design control measures to provide for the verifying or checking of the adequacy of design of the spent fuel pool liner. Specifically, calculations involving the liner had not been verified or checked to ensure the design basis requirements of ASME Boiler and Pressure Vessel Code, Section III, Division II, were included. The licensee initiated AR 02690744 and initiated actions to restore compliance.
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| This performance deficiency was more than minor because if left uncorrected it could lead to a more significant safety concern if independent spent fuel storage installation loading was conducted. The inspectors determined the finding was of very low safety significance because each of the Barrier Integrity screening questions was answered no. The inspectors determined this issue was cross cutting in the Human Performance, Design Margin area because the licensee failed to carefully guard their design margins and ensure the margins were only changed through a systematic and rigorous process.
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| Inspection Report# : 2016003 (pdf)
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| Significance: Jun 30, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Lack of Acceptance Criteria for Containment Visual Examinations The inspectors identified a finding of very-low safety significance and associated NCV of 10 CFR 50.55a(g)(4).
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| Specifically, the licensee failed to define acceptance criteria for containment visual examinations. Consequently, active containment liner degradation on a containment penetration was identified and returned to service without comparing to defined acceptance criteria. The licensee verified through visual examination that the liner thickness was marginally affected by the corrosion and documented this issue in the Corrective Action System in AR 02671728.
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| The inspectors determined that the failure to define and incorporate acceptance criteria in the containment visual examination procedure as required by 10 CFR 50.55a(g)(4) was a performance deficiency. The inspectors determined that this issue was more than minor in accordance with IMC 0612, Appendix B, Issue Screening, dated September 7, 2012, because the inspectors answered yes to the more than minor question If left uncorrected, would the performance deficiency have the potential to lead to a more significant safety concern in that active containment penetration degradation may not be properly evaluated and/or promptly corrected. Specifically, the inspectors were concerned that without acceptance standards, unacceptable containment degradation may be returned to service and adversely affect containment leakage or structural integrity. The inspectors determined this finding was of very-low safety significance (Green) based on answering no to Questions B.1 and B.2 of the Exhibit 3, Barrier Integrity Screening Questions, in IMC 0609, Attachment A, The Significance Determination Process (SDP) for Findings At-Power, issued on June 19, 2012. Specifically, the inspectors answered no to the screening question associated with an actual open pathway (e.g., breach) in the containment and no to the question associated with reduction in function of hydrogen igniters in containment. A subsequent visual examination performed by the licensee confirmed only marginal degradation of the liner thickness.
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| The inspectors determined that this finding had a cross-cutting aspect in the area of human performance in the aspect of consistent process, where individuals use a consistent, systematic approach to make decisions. Specifically, the lack of acceptance criteria allowed various interpretations for disposing of identified conditions that were inconsistent.
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| (H.13)
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| Inspection Report# : 2016002 (pdf)
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| Page 11 of 17
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| 3Q/2016 Inspection Findings - Clinton Significance: Mar 31, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Identify a Degraded Safety-Related Support
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| . The inspectors identified a finding of very low safety significance and an associated non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for the failure to identify a condition adverse to quality.
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| Specifically, the licensee failed to identify that a safety-related support associated with control room ventilation B was degraded to the point it no longer conformed to the seismic analysis and required an evaluation to determine whether it was still capable of performing its safety function during a seismic event. This issue was entered into the licensees CAP as AR 2639317. The licensees immediate corrective actions included performing an evaluation that concluded the remaining three supports would be able to withstand the stresses imposed during a seismic event and creating an action to update the seismic calculation to incorporate the evaluation performed for the degraded support.
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| The licensee also planned to re-apply a coating to the supports as well as research and install insulation that was more breathable to minimize moisture accumulation and preclude any further degradation.
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| The inspectors determined that the failure to identify a condition adverse to quality in accordance with 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, was more than minor because if left uncorrected it had the potential to lead to a more significant safety concern. Specifically, by failing to identify the support was degraded, and correct the condition, the loss of material due to corrosion could potentially progress to the point where the remaining supports would no longer be able to perform their safety function. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings at Power, issued June 19, 2012, the finding was screened against the Barrier Integrity Cornerstone and determined to be of very low safety significance because the finding did not represent a degradation of the barrier function of the control room against radiological conditions or a smoke or toxic atmosphere. The inspectors determined this finding affected the cross-cutting area of problem identification and resolution, in the aspect of evaluation, which states, The organization thoroughly evaluates issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, the licensee failed to thoroughly evaluate the issue identified by the inspectors and therefore did not recognize the degradation on the supports constituted a condition adverse to quality.
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| Inspection Report# : 2016001 (pdf)
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| Significance: Mar 31, 2016 Identified By: Self-Revealing Item Type: NCV Non-Cited Violation Failure to Assess and Manage Risk Increase for a Proposed Maintenance Activity A self-revealed finding of very low safety significance and an associated non-cited violation of 10 CFR 50.65 (a)(4) was identified on January 20, 2016, due to the licensees failure to assess and manage the risk increase from a proposed maintenance activity. Specifically, the licensee failed to manage the risk associated with racking out the continuous containment purge (CCP) A breaker, which resulted in the loss of both CCP trains, and led to an increase in primary to secondary containment differential pressure which exceeded the TS value. The licensee entered this issue into their CAP as AR 02614832. The proposed corrective actions to address this issue included creating a checklist to ensure validation of initial conditions is performed and providing training that reinforces the need to properly screen work order tasks with the appropriate risk factors.
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| The inspectors determined that the failure to assess and manage the risk increase of a proposed maintenance activity, as required by 10 CFR 50.65 (a)(4), was more than minor because it was associated with the maintenance procedure quality attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, by not properly assessing the risk of racking out the CCP A breaker the licensee did not Page 12 of 17
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| 3Q/2016 Inspection Findings - Clinton recognize the CCP B train would be impacted, which resulted in exceeding the TS value for primary to secondary containment differential pressure. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings at Power, issued June 19, 2012, the finding was screened against the Barrier Integrity Cornerstone and determined to be of very low safety significance because the finding did not represent an actual open pathway in the physical reactor containment, containment isolation system or heat removal components and it did not involve an actual reduction in function of hydrogen igniters in the reactor containment. The inspectors identified a cross-cutting aspect in the area of human performance, in the aspect of challenging the unknown, which states, individuals stop when faced with uncertain conditions; risks are evaluated and managed before proceeding. Specifically, when the licensee was preparing the work package for maintenance on the CCP system it was uncertain what activities had already been completed as part of a concurrent evolution. Instead of stopping and validating the configuration of plant equipment, assumptions were made, and the risk of the activity was not properly assessed or managed.
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| Inspection Report# : 2016001 (pdf)
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| Significance: Feb 11, 2016 Identified By: NRC Item Type: FIN Finding Failure to Perform Adequate Evaluation of Crane and Crane Support Structure Elements A finding of very low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors for the failure of the licensees design control measures to provide for the verifying or checking the adequacy of design of the fuel handling building crane and crane support structure elements. Specifically, calculations involving the crane trolley rails, crane rail clips, and crane rail clip bolts had not been verified or checked to ensure the design basis requirements of American Society of Mechanical Engineers (ASME) NOG-1-2004; American Institute of Steel Construction (AISC), 7th Edition; and Updated Safety Analysis Report (USAR) Section 3.8.4.5 were included. The licensee documented these issues in its corrective action program and initiated actions to restore compliance.
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| The performance deficiency was determined to be more than minor because if left uncorrected the performance deficiency could lead to a more significant safety concern if independent spent fuel storage installation (ISFSI) loading was conducted. The inspectors determined the finding could be evaluated using the Significance Determination Process in accordance with Inspection Manual Chapter (IMC) 0609, The Significance Determination Process for Findings At-Power, Appendix A, Exhibit 3 - Barrier Integrity Screening Questions (Section D). Based on answering No to all the questions in Exhibit 3, Section D, the inspectors determined the finding to be of very low safety significance (Green). The inspectors identified a Human Performance, Design Margin (H.6) cross-cutting aspect associated with this finding. Specifically, the licensee failed to ensure the crane trolley rails, crane rail clips, and crane rail clip bolts reflected the intended design margins of the design and licensing basis.
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| Inspection Report# : 2016010 (pdf)
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| Emergency Preparedness Occupational Radiation Safety Page 13 of 17
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| 3Q/2016 Inspection Findings - Clinton Public Radiation Safety Security Although the Security Cornerstone is included in the Reactor Oversight Process assessment program, the Commission has decided that specific information related to findings and performance indicators pertaining to the Security Cornerstone will not be publicly available to ensure that security information is not provided to a possible adversary.
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| Other than the fact that a finding or performance indicator is Green or Greater-Than-Green, security related information will not be displayed on the public web page. Therefore, the cover letters to security inspection reports may be viewed.
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| Security Although the Security Cornerstone is included in the Reactor Oversight Process assessment program, the Commission has decided that specific information related to findings and performance indicators pertaining to the Security Cornerstone will not be publicly available to ensure that security information is not provided to a possible adversary.
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| Other than the fact that a finding or performance indicator is Green or Greater-Than-Green, security related information will not be displayed on the public web page. Therefore, the cover letters to security inspection reports may be viewed.
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| Security Although the Security Cornerstone is included in the Reactor Oversight Process assessment program, the Commission has decided that specific information related to findings and performance indicators pertaining to the Security Cornerstone will not be publicly available to ensure that security information is not provided to a possible adversary.
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| Other than the fact that a finding or performance indicator is Green or Greater-Than-Green, security related information will not be displayed on the public web page. Therefore, the cover letters to security inspection reports may be viewed.
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| Miscellaneous Significance: N/A Jun 30, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Update the Updated Safety Analysis Report (USAR) - Peak Suppression Pool Temperature The inspectors identified a Severity Level IV NCV of 10 CFR 50.71(e), Periodic Update of the [Final Safety Analysis Report] FSAR, for the licensees failure to update the USAR after updating a Safety Analysis Calculation.
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| Specifically, the licensee did not update the USAR Section A3.8.3.1 and Table 15.2.9-1 to coincide with the most recent updates to the accident analysis of record. The licensee initiated AR 2664276 to document the discrepancy in the peak suppression pool temperature throughout the USAR and initiated actions to revise FSAR Section A3.8.3.1 Page 14 of 17
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| 3Q/2016 Inspection Findings - Clinton and Table 15.2.9-1 to coincide with the most recent revision to EPU-T0400.
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| The inspectors determined that the failure to update the USAR in accordance with 10 CFR 50.71(e), Periodic Update of the FSAR, with the most accurate version of calculated peak suppression pool temperature during an accident was a performance deficiency. The performance deficiency was determined to be minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012; however, the reactor oversite programs significance determination process does not specifically consider the regulatory process impact in its assessment of licensee performance, therefore, it was necessary to address this violation which impeded the NRCs ability to regulate using traditional enforcement to adequately deter non-compliance. The inspectors reviewed this issue in accordance with IMC 0612 and the NRC Enforcement Policy. Violations of 10 CFR 50.71(e) are dispositioned using the traditional enforcement process because they are considered to be violations that potentially impede or impact the regulatory process. The inspectors reviewed Section 6.1.d.3 of the NRC Enforcement Policy and determined this violation was Severity Level IV because the licensees failure to update the USAR as required by 10 CFR 50.71(e) had not yet resulted in any unacceptable change to the facility or procedures. No cross cutting aspect was assigned because traditional enforcement violations are not assessed for cross cutting aspects.
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| Inspection Report# : 2016002 (pdf)
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| Significance: N/A Jun 30, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Update the Updated Safety Analysis Report (USAR) - Condensate and Feedwater System The inspectors identified a Severity Level IV NCV of Title 10 Code of Federal Regulations (CFR) 50.71(e), Periodic Update of the FSAR, for the licensees failure to update the FSAR after implementation of license amendment 149, for extended power uprate. Specifically, the licensee did not update USAR Section 10.4.7.1.2 Performance Requirements, for the condensate and feedwater system with the design requirements for a reactor thermal power rating of 3473 MWt. The licensee entered the issue into their CAP as AR 02656128 and is preparing a technical change package to update the USAR.
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| The inspectors determined that the failure to update the USAR in accordance with 10 CFR 50.71(e), Periodic Update of the FSAR, with the design requirements for the condensate and feedwater system for a reactor thermal power rating of 3473 MWt was a performance deficiency. The performance deficiency was determined to be minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012; however, the reactor oversite programs significance determination process does not specifically consider the regulatory process impact in its assessment of licensee performance, therefore, it was necessary to address this violation which impeded the NRCs ability to regulate using traditional enforcement to adequately deter non-compliance. The inspectors reviewed this issue in accordance with NRC IMC 0612 and the NRC Enforcement Policy.
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| Violations of 10 CFR 50.71(e) are dispositioned using the traditional enforcement process because they are considered to be violations that potentially impede or impact the regulatory process. The inspectors reviewed Section 6.1.d.3 of the NRC Enforcement Policy and determined this violation was Severity Level IV because the licensees failure to update the USAR as required by 10 CFR 50.71(e) had not yet resulted in any unacceptable change to the facility or procedures. No cross cutting aspect was assigned because traditional enforcement violations are not assessed for cross cutting aspects.
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| Inspection Report# : 2016002 (pdf)
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| Significance: N/A Mar 31, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Report a Condition that Could Have Prevented Fulfillment of a Safety Function The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.72(b)(3)(v) for failing to report an event or condition, that at the time of discovery could have prevented the fulfillment of a safety function, to the NRC within eight hours. Specifically, control room operators placed both divisions of reactor water cleanup differential flow instruments in bypass, which rendered the instruments inoperable and resulted in a loss of the isolation function.
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| The licensee entered this issue into the CAP as AR 02645140 and created an action to submit an licensee event report Page 15 of 17
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| 3Q/2016 Inspection Findings - Clinton under 10 CFR 50.73(a)(2)(v).
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| The inspectors determined that the failure to report an event or condition, that at the time of discovery could have prevented the fulfillment of a safety function, to the NRC within 8 hours as required by 10 CFR 50.72(b)(3)(v) was a performance deficiency. The inspectors reviewed this issue in accordance with IMC 0612 and the Enforcement Manual. Violations of 10 CFR 50.72 are dispositioned using the traditional enforcement process because they are considered to be violations that potentially impede or impact the regulatory process. The inspectors reviewed Section 6.9.d.9 of the NRC Enforcement Policy and determined this violation was Severity Level IV because the licensees failure to make the report, as required by 10 CFR 50.72, did not cause the NRC to reconsider a regulatory position or undertake substantial further inquiry. No cross-cutting aspect was assigned because cross-cutting aspects are not assigned to traditional enforcement only violations.
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| Inspection Report# : 2016001 (pdf)
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| Significance: N/A Mar 31, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Report Condition Prohibited by Technical Specifications The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.73(a)(2)(i)(B) for failing to report to the NRC, within 60 days of discovery, a condition prohibited by the plants TS. Specifically, the licensee failed to notify the NRC of two instances where they failed to comply with TS 3.3.6.1 and TS 3.3.6.2 and enter the limiting condition for operation action statements when required. The licensee entered this issue into their CAP as AR 02619114 and subsequently issued a licensee event report on March 16, 2016.
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| The inspectors determined that the failure to report a condition prohibited by the plants TS as required by 10 CFR 50.73(a)(2)(i)(B), within 60 days of discovery, was a performance deficiency. The inspectors reviewed this issue in accordance with IMC 0612 and the Enforcement Manual. Violations of 10 CFR 50.73 are dispositioned using the traditional enforcement process because they are considered to be violations that potentially impede or impact the regulatory process. The inspectors reviewed Section 6.9.d.9 of the NRC Enforcement Policy and determined this violation was Severity Level IV because the licensees failure to make the report, as required by 10 CFR 50.73, did not cause the NRC to reconsider a regulatory position or undertake substantial further inquiry. No cross-cutting aspect was assigned because cross-cutting aspects are not assigned to traditional enforcement only violations.
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| Inspection Report# : 2016001 (pdf)
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| Significance: N/A Dec 31, 2015 Identified By: NRC Item Type: NCV Non-Cited Violation FAILURE TO UPDATE THE FINAL SAFETY ANALYSIS REPORT (FSAR) - HYDROGEN WATER CHEMISTRY SYSTEM The inspectors identified a Severity Level IV Violation of Title 10 Code of Federal Regulations (CFR) 50.71(e),
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| "Periodic Update of the FSAR", for the licensee's failure to update the FSAR after installing a hydrogen water chemistry system into the plant to reduce rates of intergranular stress corrosion cracking (IGSCC) in recirculation piping and reactor vessel internals. Specifically, the licensee did not update Section 5.4.15, "Hydrogen Water Chemistry System" of the FSAR to include a design basis and description of process and system used to periodically injection noble metals. The licensee entered this issue into the corrective action program as AR 02594259 and is revising the FSAR include additional the design basis and additional system description for noble metal injection.
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| The inspectors determined that the failure to update the FSAR in accordance with 10 CFR 50.71(e), "Periodic Update of the FSAR", with the design basis and description of the process and system used to periodically injection noble metals was a performance deficiency warranting a significance evaluation. The inspectors reviewed this issue in Page 16 of 17
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| 3Q/2016 Inspection Findings - Clinton accrodance with NRC inspection manual chapter 0612 and the NRC enforcement manual. Violations of 10 CFR 50.71 (e) are dispositioned using the traditional enforcement process because they are considered to be violations that potentially impede or impact the regulatory process. The inspectors reviewed section 6.1.d.3 of the NRC Enforcement Policy and determined this violation was Severity :Level IV because the licensee's failure to update the FSAR as required by 10 CFR 50.71(e) had not yet resulted in any unacceptable change to the facility or procedures. No cross cutting aspect was assigned because cross cutting aspects are not assigned to traditional enforcement only violations.
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| Inspection Report# : 2015004 (pdf)
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| Last modified : December 08, 2016 Page 17 of 17
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| 4Q/2016 Inspection Findings - Clinton Clinton 4Q/2016 Plant Inspection Findings Initiating Events Significance: Sep 30, 2016 Identified By: Self-Revealing Item Type: NCV Non-Cited Violation Failure to Prevent Recurrence of a Significant Condition Adverse to Quality The inspectors documented a self-revealing finding of very low safety significance and an NCV of 10 CFR 50, Appendix B, Criterion XVI for the licensee's failure to take corrective action to preclude repetition of a significant condition adverse to quality (SCAQ). After identifying IGSCC on main steam flex hoses in 2007 and concluding the leakage constituted a SCAQ, the licensee's corrective actions to prevent recurrence failed to prevent pressure boundary leakage at the same location in 2016. The licensee entered this issue into their corrective action program as AR 02670593. The affected hoses were replaced. The licensee is also developing a design change to address at least one of the three factors that contributes to IGSCC.
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| The inspectors determined that the licensee's failure to take corrective actions to prevent recurrence of an SCAQ was a performance deficiency and more than minor because if left uncorrected pressure boundary leakage could become a more significant concern. Specifically, pressure boundary leakage is not allowed by TS and any leakage requires the plant to be shutdown. The finding was screened as low safety significant because it did not result in exceeding the RCS leak rate for a small LOCA and did not affect systems used to mitigate a LOCA. No cross cutting aspect was assigned as the original issue occurred greater than three years ago and was not reflective of current performance.
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| Inspection Report# : 2016003 (pdf)
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| Significance: Jun 30, 2016 Identified By: NRC Item Type: FIN Finding Material Unsecure in the Secured Material Zone The inspectors identified a finding of very low safety significance for the failure to ensure material placed within the transformer secured material zone, was secured as required by station procedure MA-AA-716-026, Station Housekeeping/Material Condition Program, Revision 14. Specifically, the inspectors identified unsecured scaffold poles and knuckles within the licensee established secure material zone. The licensee has entered this issue into their corrective action program (CAP) as action request AR 02668245. The material was immediately removed from the secured zone by the licensee.
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| The inspectors determined the licensees failure to ensure material within the secured material zone was adequately secured in accordance with procedure MA-AA-716-026, Station Housekeeping/Material Condition Program, was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations, and is therefore a finding. Specifically, by not securing material in the vicinity of main power transformers, the material could become a missile and impact the transformers causing a potential reactor SCRAM. The finding was screened against the Initiating Events Cornerstone and determined to be of very low safety significance (Green) because the finding did not involve the complete or partial loss of a support system that contributes to the likelihood of, or cause an initiating event and did not affect mitigation equipment. The inspectors determined that this finding Page 1 of 18
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| 4Q/2016 Inspection Findings - Clinton had a cross-cutting aspect in the area of human performance in the aspect of field presence, where leaders are commonly seen in the work areas of the plant observing, coaching, and reinforcing standards and expectations.
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| Specifically, since initial identification of the issue the inspectors have noted that while discussions on when to perform walkdowns took place, the supervisors or managers did not ensure sufficient field presence to reinforce the standards and expectations, leading to material continuing to be easily found by the inspectors. (H.2)
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| Inspection Report# : 2016002 (pdf)
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| Significance: Jun 30, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Perform Surface Examination Prior to Weld Repair of a Reactor Water Cleanup System Pipe The inspectors identified a finding of very-low safety significance and associated NCV of 10 CFR 50.55a(g)(4).
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| Specifically, the licensee failed to perform a surface examination to detect cracking on reactor water cleanup small-bore piping prior to performing a weld repair. The license documented the issue in the CAP as AR 02671726 and AR 02685332 and performed an operability review. The licensee has prepared a work order to perform a surface exam of the existing weld and surrounding area.
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| The inspectors determined that the failure to perform the surface examination prior to weld repair of RWCU pipe 1G33C001B as required by 10 CFR 50.55a(g)(4) was a performance deficiency. The inspectors determined that this issue was more-than-minor in accordance with IMC 0612, Appendix B, because it adversely affect the Initiating Events Cornerstone attribute of barrier integrity and because the answer to the question of If left uncorrected, would the performance deficiency have the potential to lead to a more significant safety concern? was yes. Specifically, the lack of a surface exam may result in the entire defect not being removed during the repair and the potential existed for a cracked pipe to remain in service. This could lead to a repeat leak of reactor coolant. The inspectors determined this finding was of very-low safety significance (Green) based on answering no to Question A.1 and A.2 of the Exhibit 1, Initiating Events Screening Questions, in IMC 0609, Attachment A, The Significance Determination Process (SDP) for Findings At-Power. Specifically, the inspectors answered no to the screening question associated with a reactor coolant system leak exceeding the leak rate for a small loss of coolant accident (LOCA) and no to the screening question associated with systems used to mitigate a LOCA. A subsequent visual examination of the weld repair revealed an absence of cracking and the licensee also planned to perform a follow-up surface examination of the repaired area to look for cracking. The inspectors determined that this finding had a cross-cutting aspect in the area of human performance in the aspect of resources, where leaders ensure that personnel, equipment, procedures and other station resources are available and adequate to support nuclear safety. Specifically, the work order that performed the work did not specify a surface examination of the base metal prior to welding. (H.1)
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| Inspection Report# : 2016002 (pdf)
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| Significance: Jun 30, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Obtain License Amendment prior to Operating Reactor Water Cleanup Bypass Switches The inspectors identified a Severity Level IV NCV of 10 CFR 50.59(a)(1), Changes, Tests, and Experiments, and an associated Green finding for the licensees failure to perform an adequate written safety evaluation to provide the basis that changes to the Updated Safety Analysis Report (USAR) and station procedures did not involve an unreviewed safety question. Specifically, the licensee changed the USAR and station procedures to allow operators to a defeat the safety function of the reactor water cleanup (RWCU) isolation valves to prevent unwarranted isolation signals during normal operation without obtaining prior Commission approval. The licensee entered this issue into their CAP as AR 02685337 and will be changing station procedures to prevent placing the RWCU leak detection divisional bypass switches in bypass except for instrument channel maintenance, testing or calibration.
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| The inspectors determined that the licensees failure to perform an adequate written safety evaluation to provide the Page 2 of 18
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| 4Q/2016 Inspection Findings - Clinton basis that changes to the USAR and station procedures did not involve an unreviewed safety question was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the equipment performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The finding was screened against the Initiating Events cornerstone and determined to be of very low safety significance (Green) because the finding did not result in exceeding the RCS leak rate for a small LOCA and did not affect other systems used to mitigate a LOCA resulting in a total loss of their function (e.g. Interfacing System LOCA). No cross cutting aspect was assigned because the inspectors determined the performance deficiency was not indicative of current plant performance. Violations of 10 CFR 50.59 are dispositioned using the traditional enforcement process because they are considered to be violations that potentially impede or impact the regulatory process. The inspectors reviewed Section 6.1.d.2 of the NRC Enforcement Policy and determined this violation was Severity Level IV because the resulting changes were evaluated by the SDP as having very low safety significance.
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| Inspection Report# : 2016002 (pdf)
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| Mitigating Systems Significance: Dec 01, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Promptly Identify that the Incapability of the RHR Design to Support TS Operability Requirements Was a CAQ (Section 4OA2.b(1))
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| Green: The team identified a finding of very-low safety significance (Green) and an associated NCV of Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensee failure to promptly identify that the incapability of the residual heat removal (RHR) design to support Technical Specifications (TS) operability requirements was a condition adverse to quality. Specifically, when reactor water temperature was greater than 150 degrees Fahrenheit, RHR could not be realigned from shutdown cooling mode of operations to provide the TS required functions of the emergency core cooling system, suppression pool cooling, containment spray, and feedwater leakage control system. The licensee captured this issue in their Corrective Action Program (CAP) as Action Request (AR) 02742439 and AR 03948042, and planned to submit a License Amendment Request to align TS requirements with the design capabilities.
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| The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of design control and adversely affected the associated cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the performance deficiency resulted in voluntarily declaring TS functions inoperable when performing shutdown cooling operations, which did not ensure the associated mitigating systems availability or capability to respond to an initiating event. The team determined that this finding was of very low safety significance (Green). Specifically, there were no known instances where the finding: (1) represented a loss of system safety function; (2) represented an actual loss of safety function of at least a single train or two separate safety systems out of service for greater than their TS allowed outage time; (3) involved non-TS trains of equipment; (4) involved a degradation of a functional RHR auto-isolation on low reactor vessel level; (5) impacted external event protection; or (6) involved fire brigade issues. The team did not identify a cross cutting aspect associated with this finding because it did not reflect current licensee performance since the performance deficiency occurred more than 3 years ago.
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| (Section 4OA2.b(1))
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| Inspection Report# : 2016009 (pdf)
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| Page 3 of 18
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| 4Q/2016 Inspection Findings - Clinton Significance: Sep 30, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Exceeded Technical Specification Allowed Outage Time for Main Turbine Bypass System The inspectors identified a finding of very low safety significance and an associated NCV of Technical Specification 3.7.6, Main Turbine Bypass System for the licensees failure to meet the limiting conditions for operation and complete the associated required actions after making a deficient change to the turbine bypass valve surveillance testing frequency. Specifically, with the main turbine bypass system inoperable and without the Core Operating Limits Report (COLR) limits for thermal power, minimum critical power ratio (MCPR), and linear heat generation rate (LGHR) with the main turbine by pass system inoperable applied, thermal power was not reduced to less than 21.6 percent of rated thermal power within six hours. The licensee entered this issue into their corrective action program as AR 02690657. The licensee restored compliance by applying the COLR limits for reactor thermal power, MCPR and LGHR.
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| The inspectors determined the failure to meet the limiting conditions for operation and complete the associated required actions prior to the end of the specified completion times was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was screened against the Mitigating Systems cornerstone and determined to be of very low safety significance because all of the associated questions in IMC 0609, Appendix A, were answered no. The inspectors determined this finding affected the cross-cutting area of human performance, in the aspect of change management, where leaders use a systematic process for evaluating and implementing change so that nuclear safety remains the overriding priority because the licensees change management process was not fully utilized by senior management when evaluating and implementing a change to the turbine bypass valve surveillance testing frequency. (H.3)
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| Inspection Report# : 2016003 (pdf)
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| Significance: Sep 30, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Perform a 50.59 Screening for Changing the Frequency of Exercising the Turbine Bypass Valves The inspectors identified a Severity Level IV NCV of 10 CFR 50.59 4(d)(1), Changes, Tests, and Experiments, and an associated Green finding for the licensees failure to perform a written evaluation which provided the bases for determining that changing the turbine bypass valve surveillance testing frequency from every 31 days, as specified in the Updated Safety Analysis Report, to once a year did not require a license amendment. The licensee has entered this issue into their corrective action program as AR 02720163. The licensee is currently evaluating the issue in accordance with their procedure for changes to the facility.
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| The inspectors determined that the licensees failure to perform a written evaluation to provide the basis for the determination that a change to the facility, a change to a procedure, or a change to a test or experiment did not require a license amendment was a performance deficiency. The performance deficiency was more than minor in accordance with IMC 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," dated September 7, 2012, because, it was associated with the procedure quality attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2012, the finding was screened against the Mitigating Systems cornerstone and determined to be of very low safety significance because all of the associated questions in IMC 0609, Appendix A, were answered no. Violations of 10 Page 4 of 18
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| 4Q/2016 Inspection Findings - Clinton CFR 50.59 are dispositioned using the traditional enforcement process because they are considered to be violations that potentially impede or impact the regulatory process. The inspectors reviewed Section 6.1.d.2 of the NRC Enforcement Policy and determined this violation was Severity Level IV because the resulting changes were evaluated by the SDP as having very low safety significance. The inspectors determined this finding affected the cross-cutting area of human performance, in the aspect of consistent process, where individuals use a consistent, systematic approach to make decisions. The licensee made a decision to proceed with implementation of a change to the turbine bypass valve surveillance testing frequency after a plant oversight committee review in lieu of following their consistent, systematic process for evaluating changes to the USAR. (H.13)
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| Inspection Report# : 2016003 (pdf)
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| Significance: Aug 09, 2016 Identified By: NRC Item Type: FIN Finding Failure to have hose configurations that were verified to be able to ensure a timely and successful implementation of a FLEX strategy Green. Two examples of a finding of very low safety significance was identified by the inspectors for the licensees failure to have hose configurations that were verified to be able to ensure a timely and successful implementation of a flexible response (FLEX) strategy. Specifically, the licensee did not ensure through evaluations, calculations, analyses or any other means that the strategy for maintaining core cooling, containment heat removal and Spent Fuel Pool (SFP) cooling during a Beyond-Design-Basis External Event (BDBEE) flooding scenario would be capable of fulfilling its function. No violation of NRC requirements were identified.
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| The performance deficiency is more than minor because it was associated with the mitigating systems cornerstone objective attribute of protection against external factors, specifically the BDBEE flood hazard, and it adversely affected the cornerstone attribute of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Issues identified through TI-191 are evaluated through a cross-regional panel using IMC 0609, Appendix M, Significance Determination Process Using Qualitative Criteria, as informed by draft Appendix O, Post Fukushima Mitigation Strategies Significance Determination Process. The finding was determined to be of very low safety significance (Green). The inspectors concluded that the cause of the finding involved a cross-cutting component in the Human Performance area of Design Margins because the organization did not ensure the selected strategy contained the required verification that it could be successfully implemented. [H.6]
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| Inspection Report# : 2016007 (pdf)
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| Significance: Jun 30, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Inspection Fails to Identify Safety Related Cables Submerged in Cable Vault The inspectors identified a finding of very low safety significance (Green) and associated NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action Program, for the failure to identify a condition adverse to quality.
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| Specifically, the licensee failed to identify that portions of the Division 1 SX safety related cables, which are not rated for submergence, were under water. The licensee entered this issue into their corrective action program as action requests AR 02648804 and AR 02648507. Operators took actions to pump out the water to ensure the cables were returned to a dry condition.
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| The inspectors determined the licensees failure to identify a condition adverse to quality was contrary to 10 CFR 50, Appendix B, Criterion XVI, Corrective Action Program, and was a performance deficiency. The performance deficiency was determined to be more than minor because the finding was associated with the Mitigating Systems Page 5 of 18
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| 4Q/2016 Inspection Findings - Clinton Cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, because the SX cables are not rated for submergence, they could degrade and affect the reliability of the SX system. The finding was screened against the Mitigating Systems cornerstone and determined to be of very low safety significance (Green) because the inspectors answer Yes to the question does the SSC maintain its operability or functionality. Specifically, the SX system submerged cables did not cause the SX system to be inoperable or nonfunctional. The inspectors determined that this finding had a cross-cutting aspect in the area of human performance in the aspect of resources, where leaders ensure that personnel, equipment and other resources are available and adequate to support nuclear safety. Specifically, the individuals performing the inspection did not have the necessary resources, such as training, procedures, drawings or a detailed pre-job brief, to identify the cables sloped downwards in the cable vault and were submerged in the water. (H.1)
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| Inspection Report# : 2016002 (pdf)
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| Significance: Jun 30, 2016 Identified By: NRC Item Type: FIN Finding Failure to Properly Install Cable Vault Mitigating Equipment The inspectors identified a finding of very low safety significance (Green) for the failure to incorporate human performance standards when developing work package instructions in accordance with MA-AA-716-010,Maintenance Planning, Revision 23. Specifically, the licensee did not assure the cable vault dewatering system installation and maintenance work order (WO) included the appropriate details to troubleshoot and install the cable vault sump pumps and float switches. This resulted in installation of the equipment in a manner that prevented detection and removal of water from the cable vaults, allowing cables to remain submerged undetected. The licensee entered this issue into their CAP as AR 02668245. The corrective actions performed by the licensee included placing the sump pumps in the right location and adjusting the float switches to ensure the indications would alert operators when the vaults needed to be pumped.
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| The inspectors determined that the failure to incorporate human performance standards when developing work package instructions in accordance with MA-AA-716-010, Maintenance Planning, Revision 23, was a performance deficiency. The performance deficiency was determined to be more than minor because if left uncorrected the performance had the potential to lead to a more significant safety concern. Specifically, by not appropriately installing the sump pumps and float switches, the cables would be allowed to remain submerged undetected. The finding was screened against the Mitigating Systems cornerstone and determined to be of very low safety significance (Green) because the inspectors answer Yes to the question does the SSC maintain its operability or functionality?. This finding has a cross-cutting aspect in the area of human performance in the aspect of conservative bias, where individuals use decision making practices that emphasize prudent choices over those that are simply allowable. Specifically, because the licensee classified the cable vault dewatering system as a maintenance tool, they decided it was not necessary to include specific instructions within the WOs related to ensure the troubleshooting and re-installation activities were performed appropriately. (H.14)
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| Inspection Report# : 2016002 (pdf)
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| Significance: Jun 30, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Inadequate Ultrasonic Examination Method Used to Detect Crack-Like Flaws On May 17, 2016, the inspectors identified a finding of very-low safety significance and associated NCV of 10 CFR 50, Appendix B, Criterion IX, Control of Special Processes, for the licensees failure to ensure that nondestructive testing was controlled and accomplished using qualified procedures in accordance with applicable codes and standards. Specifically, the licensee did not implement an angle beam ultrasonic (UT) examination to detect cracking in a degraded SX pipe prior to implementation of a weld overlay repair. The licensee subsequently performed the required UT examination to confirm the absence of cracks and documented the issue in the CAP in AR 02671724.
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| Page 6 of 18
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| 4Q/2016 Inspection Findings - Clinton The inspectors determined that this finding was more than minor because if left uncorrected, the failure to perform the UT would become a more significant safety concern. Specifically, if left uncorrected, the use of an unqualified UT examination for detection of cracks could result in undetected cracks that propagate to failure during service. The inspectors determined this finding was of very low safety significance (Green) based on answering yes to the questions in Part A of Exhibit 2, Mitigating Systems Screening Questions, in IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Specifically, the inspectors answered yes to the screening question If the finding is a deficiency affecting the design or qualification of a mitigating SSC [structures, systems, or components], does the SSC maintain its operability or functionality? because the licensee subsequently performed appropriate UT examination to confirm that cracks were not present. The finding had a cross-cutting aspect in the area of Human Performance for Procedure Adherence, because the licensee failed to follow processes, procedures, and work instructions to ensure that the appropriate UT examination was applied to the degraded SX pipe.
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| (H.8)
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| Inspection Report# : 2016002 (pdf)
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| Significance: Mar 31, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Operability Determination Failed to Examine Test Failures The inspectors identified a finding of very low safety significance and an associated non-cited violation of 10, Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion V, Instructions Procedures and Drawings, for the failure to follow Station Procedure OP-AA-108-115, Operability Determinations, Revision 16. Specifically, after valve 1SX027C, a valve required for residual heat removal operability, failed a surveillance test, the licensee did not base the operability determination on a detailed examination of the deficiency and did not document a basis for why a reasonable expectation of operability existed. The licensee entered this issue into their corrective action program (CAP) as Action Request (AR) 02553168 and AR 02558101. The licensee revised the in-service testing program evaluation for valve 1SX027C and documented additional details to support declaring the valve operable.
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| The inspectors determined the failure to follow Station Procedure OP-AA-108-115 was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to correctly perform an operability evaluation for valve 1SX027C had the potential to allow an inoperable condition to go undetected. Using IMC 0609, , Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2012, the finding was screened against the Mitigating Systems Cornerstone and determined to be of very low safety significance because the finding: was not a deficiency affecting the design or qualification of a mitigating system; did not represent a loss of system and/or function; did not represent an actual loss of function of a single train for greater than its Technical Specification (TS) allowed outage time; and did not represent an actual loss of function of one or more non-TS trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. The inspectors determined this finding affected the cross-cutting area of human performance, in the aspect of resources, where leaders ensure that personnel, equipment, procedures, and other resources are available and adequate to support nuclear safety.
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| Specifically, Station Procedure CPS 9053.04, provided guidance that the valve could remain operable for 96 hours without providing an appropriate basis.
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| Inspection Report# : 2016001 (pdf)
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| Significance: Mar 31, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Page 7 of 18
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| 4Q/2016 Inspection Findings - Clinton Inadequate Extent of Condition Associated with an ACE The inspectors identified a finding of very low safety significance and an associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion II, Quality Assurance Program, for the failure to follow a Quality Assurance Program implementing procedure. Specifically, the licensee failed to perform an adequate extent of condition review as required by PI-AA-125, Corrective Action Program, while evaluating a lack of proficiency in applying the licensing basis for structures, systems and components (SSCs) when implementing the 50.59 process. The licensee documented this issue in their CAP as AR 02641397. Immediate corrective actions included a review of the extent of condition performed by the engineering department and a recommended action of expanding the scope of the review to include additional 50.59 evaluations.
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| The inspectors determined the failure to follow a Quality Assurance Program implementing procedure was more than minor because if left uncorrected it had the potential to lead to a more significant safety concern. Specifically, if the extent of condition review is too narrowly assessed there is the potential for other safety significant systems to have been impacted by a lack of proficiency in applying the licensing basis. As a result, the SSCs may not perform their intended safety function as defined in the Updated Safety Analysis Report. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings at Power, issued June 19, 2012, the finding was screened against all cornerstones and determined to be of very low safety significance because there was no reasonable indication that the criteria in Appendix A were met. The inspectors determined this finding affected the cross-cutting area of human performance, in the aspect of procedure adherence, where individuals follow processes, procedures and work instructions. Specifically, the licensee did not effectively adhere to all available portions of CAP procedures, which led to a narrowly focused extent of condition.
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| Inspection Report# : 2016001 (pdf)
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| Significance: Feb 04, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Perform and Adequate Equipment Apparent Cause Evaluation (Section 4OA4)
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| The inspectors identified a finding of very-low safety significance (Green), and an associated Non-Cited Violation of Title 10, Code of Federal Regulations, Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to follow Step 4.3.4 of procedure PI-AA-125, Corrective Action Program Procedure.
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| Specifically, the licensee failed to perform Class B Equipment Apparent Cause Evaluation (EACE) 2381871, 1SX01PC Failed to Start for Testing, in accordance with PI-AA-125-1003, Apparent Cause Evaluation Manual, because they: (1) failed to analyze each causal factor to determine contributing causes as required by Step 4.4.1.2; and (2) failed to assign an effectiveness review for the EACE as required by Step 4.4.9.1. The licensee entered this finding into their Corrective Action Program and revised their EACE to: (1) include three contributing causes; (2) upgrade a corrective action to a corrective action to prevent recurrence; and (3) assign an effectiveness review to determine the effectiveness of the corrective action to prevent recurrence.
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| The performance deficiency was determined to be more than minor because if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, an effectiveness review is required to provide assurance that the Division 3 SX pump design change is successful in preventing a future recurrence of pump failure, which would be a more significant safety concern. The finding impacted the Mitigating Systems Cornerstone and screened as having very-low safety significance (Green) because although the finding is a deficiency ultimately affecting the design or qualification of the Division 3 SX pump, the pump still maintains its operability. The inspectors determined this finding had an associated cross-cutting aspect in the area of Human Performance (Conservative Bias) because although a B Apparent Cause Evaluation may have been allowable for investigating the failure of the Division 3 SX pump, had an A Root Cause Analysis been performed, a more rigorous investigation process would have been used to identify contributing causes, assign corrective actions, and identify effectiveness reviews for the failure of the Division 3 SX pump. [H.14] (Section 4OA4.02.03.f)
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| Page 8 of 18
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| 4Q/2016 Inspection Findings - Clinton Inspection Report# : 2016008 (pdf)
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| Barrier Integrity Significance: Dec 01, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Non Conservative Control Room Radiological Habitability Assessment (Section 1R21.3.b(1))
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| Green. The team identified a finding of very-low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensee failure to use a technically appropriate analytical methodology in the control room radiological habitability calculation. Specifically, the licensee used a methodology that inappropriately characterized the control room heating, ventilation and air conditioning (HVAC) system outside air intake design resulting in a calculated control room dose following a loss of coolant accident that exceeded the applicable limit. The licensee captured this issue in their CAP as AR 02742442, completed an operability evaluation, and issued an NRC event notification.
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| The performance deficiency was determined to be more-than-minor because it was associated with the Barrier Integrity cornerstone attribute of design control and affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events.
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| Specifically, the performance deficiency resulted in the control room expected dose following a loss of coolant accident to exceed the applicable limits prompting an operability evaluation. The finding screened as of very-low safety significance (Green) because it only represented a degradation of the radiological barrier function provided for the control room. Specifically, the finding did not affect the control room barrier function against smoke or a toxic atmosphere. The team did not identify a cross-cutting aspect associated with this finding because it was not confirmed to reflect current performance due to the age of the performance deficiency. Specifically, the affected calculations were performed more than 3 years ago. (Section 1R21.3.b(1))
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| Inspection Report# : 2016009 (pdf)
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| Significance: Dec 01, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Scope SFP Temperature and Level Instruments into the Maintenance Rule Program (Section 1R21.3.b(2))
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| Green. The team identified a finding of very-low safety significance (Green) and an associated NCV of Paragraph (b)
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| (2)(i) of 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, for the licensee failure to scope non-safety related mitigating structure, systems, and components (SSCs) used within an emergency operating procedure (EOP) into Maintenance Rule Program. Specifically, an EOP used spent fuel pool (SFP) low-level and high-temperature parameters as distinct entry criteria but the associated components were not included in the scope of the Maintenance Rule Program. The licensee captured the team concerns in their CAP as AR 02736193, performed an extent of condition to identify any other SSC addition to the EOPs requiring them to be added to the Maintenance Rule Program scope, and initiated plans to incorporate the affected SSCs into the Maintenance Rule Program scope.
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| The performance deficiency was determined to be more-than-minor because it was associated with the Barrier Integrity cornerstone attribute of SSC performance and affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events.
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| Page 9 of 18
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| 4Q/2016 Inspection Findings - Clinton Specifically, a key aspect of the Maintenance Rule is to ensure that maintenance activities are performed in a manner that provide reasonable assurance that SSCs within its scope perform reliably and are capable of providing their intended Maintenance Rule function(s). In the case of the SFP temperature instruments, the licensee was not performing preventive maintenance to ensure that degradation, such as instrument drift, did not adversely affect their ability to detect and alarm EOP entry conditions such that mitigating actions could be implemented to preserve secondary containment. The finding screened as of very-low safety significance (Green) because it only represented a degradation of the radiological barrier function provided for the control room. Specifically, the finding did not cause SFP temperature to exceed the maximum analyzed limit, a detectible release of radionuclides, water inventory to decrease below the analyzed limit, or an adverse effect to the SFP neutron absorber or fuel loading pattern. The team determined that the finding had a cross cutting aspect in the area of human performance because the licensee did not use a systematic process for evaluating and implementing changes when updating the affected EOP in 2015. (Section 1R21.3.b(2)) [H.3]
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| Inspection Report# : 2016009 (pdf)
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| Significance: N/A Dec 01, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Amend the UFSAR Indicating Choice to Comply with 10 CFR 50.68(b) (Section 1R21.3.b(3))
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| The team identified a Severity Level-IV NCV of 10 CFR 50.68, Criticality Accident Requirements, Paragraph (b)
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| (8), for the licensee failure to amend the Updated Final Safety Analysis Report (UFSAR) to indicate they chose to comply with 10 CFR 50.68(b). Specifically, in 2005, the licensee chose to comply with 10 CFR 50.68(b) but did not amend the UFSAR following the issuance of the associated license amendment. The licensee captured this issue in their CAP as AR 02741851, reasonably confirmed compliance with 10 CFR 50.68(b) requirements (1) through (7) was maintained, and initiated plans to update the UFSAR to specifically indicate that Clinton Power Station chose to comply with 10 CFR 50.68(b).
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| The Significance Determination Process does not specifically consider the impact to the regulatory process in its assessment of licensee performance. Therefore, it was necessary to address this violation, which potentially impacts the NRCs ability to regulate, using traditional enforcement to adequately deter non-compliance. Specifically, failure to update the UFSAR challenges the regulatory process because it serves as a reference document used, in part, for recurring safety analyses, evaluating License Amendment Request, and in preparation for and conduct of inspection activities. The team determined the traditional enforcement violation was a Severity Level-IV violation in accordance with Section 6.1.d.3 of the Enforcement Policy because the un-updated UFSAR had not been used to evaluate a facility or procedure change that resulted in a condition evaluated as having low-to-moderate or greater safety significance by the Significance Determination Process. However, it had a material impact on safety or licensed activities. Specifically, the un-updated UFSAR could be used to perform evaluations of facility or procedure changes, which would have the potential to result in unacceptable conditions and/or regulatory decisions. Traditional enforcement violations are not assessed for cross-cutting aspects. (Section 1R21.3.b(3))
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| Inspection Report# : 2016009 (pdf)
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| Significance: Dec 01, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Verify the Adequacy of Design Assumptions Related to Time Critical Operator Actions (Section 1R21.6.b(1))
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| Green. The team identified a finding of very-low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensee failure to verify the adequacy of design assumptions related to time critical operator actions made in calculations associated with the control room HVAC and RHR emergency SFP cooling functions. Subsequently, it was determined that operators did not fully understand the control room HVAC system operational demands and that the operational assumptions of the RHR emergency SFP cooling Page 10 of 18
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| 4Q/2016 Inspection Findings - Clinton design were unrealistic. The licensee captured these issues into the CAP as AR 02739012, AR 03943566, and AR 02741909; reasonably demonstrated that SFP makeup sources would be available to cope with a prolonged loss of SFP cooling; conducted operator training; and provided refined procedural guidance to ensure the control room HVAC system would be operated consistent with the design assumptions.
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| The performance deficiency was determined to be more-than-minor because it was associated with the Barrier Integrity cornerstone attribute of human performance and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the pilot validations of the control room HVAC system operational assumptions demonstrated a significant reduction in margin due to, in part, a lack of operator understanding of the operational assumptions.
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| Additionally, a preliminary review of procedures associated with SFP cooling and RHR determined the operational assumptions of the calculation related to RHR emergency SFP cooling were not bounding. The team determined that this finding was of very low safety significance (Green). Specifically, the control room HVAC system finding example only represented a degradation of the radiological barrier function provided for the control room in that it did not affect the control room barrier function against smoke or a toxic atmosphere. In addition, the finding example related to emergency SFP cooling did not cause SFP temperature to exceed the maximum analyzed limit, a detectible release of radionuclides, water inventory to decrease below the analyzed limit, or an adverse effect to the SFP neutron absorber or fuel loading pattern. The team determined that the finding had a cross-cutting aspect in the area of Human Performance because the operation and engineering organizations did not effectively communicate and coordinate their respective roles in developing the control room HVAC system validation in a manner that supported nuclear safety. (Section 1R21.6.b(1)) [H.4]
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| Inspection Report# : 2016009 (pdf)
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| Significance: Dec 01, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Follow the Operability Determination Process Following the Identification of a Control Room HVAC System Design Issue (Section 4OA2.b(2))
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| Green: The team identified a finding of very-low safety significance (Green), and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instruction, Procedures, and Drawings, for the licensee failure to follow the operability evaluation procedure after the identification of a significant design error associated with the control room HVAC system. Specifically, the licensee did not identify the affected safety function, and promptly restore or confirm system operability. The licensee captured these issues into the CAP as AR 03948266 and performed a preliminary engineering evaluation using another alternative analytical methodology that reasonably determined the control room HVAC system remained operable.
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| The performance deficiency was determined to be more-than-minor because it was associated with the Barrier Integrity cornerstone attribute of human performance and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the performance deficiency resulted in a condition where reasonable doubt on the operability of the control room HVAC system remained following the identification of a significant design error. The finding screened as of very low safety significance (Green) because it only represented a degradation of the radiological barrier function provided for the control room. Specifically, the finding did not affect the control room barrier function against smoke or a toxic atmosphere. The team identified that the finding had a cross-cutting aspect in the area of Human Performance because the licensee did not provide training to maintain a knowledgeable workforce that would facilitate an adequate implementation of the operability evaluation process following the identification of a non-conforming design-related issue. (Section 4OA2.b(2)) [H.9]
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| Inspection Report# : 2016009 (pdf)
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| Page 11 of 18
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| 4Q/2016 Inspection Findings - Clinton Significance: Sep 30, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Scope Fuel Building Ventilation Pressure Control into Maintenance Rule The inspectors identified a finding of very low safety significance and an NCV of 10 CFR 50.65 (b) for the licensees failure to scope a non-safety related structure, system and component (SSC), whose function is used in one or more Emergency Operating Procedures (EOP) and whose failure could cause actuation of a safety-related system, into maintenance rule. Specifically, the licensee failed to scope the non-safety related fuel building pressure control function into their maintenance rule program. The licensee has entered this issue into their corrective action program as AR 02716300. The licensee is scoping the pressure control function of fuel building ventilation into maintenance rule.
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| The inspectors determined that the licensees failure to scope a non-safety related system whose function is used in one or more EOPs and whose failure caused the actuation of a safety-related system into maintenance rule was a performance deficiency. The performance deficiency was determined to be more than minor because it affects the SSC and barrier performance attribute of the Barrier Integrity cornerstone and adversely affects the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. The finding was determined to be of very low safety significance because the inspectors answered yes to the question does the finding only represent a degradation of the radiological barrier function provided for the control room, or auxiliary building, or spent fuel pool, SBGT system (BWR)?. The inspectors determined this finding affected the cross-cutting area of human performance, in the aspect of avoiding complacency, because the licensee identified water intrusion of the fuel building pressure sensing line was a longstanding, latent, known problem and failed to recognize and appropriately challenge how the function was scoped into maintenance rule. (H.12)
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| Inspection Report# : 2016003 (pdf)
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| Significance: Sep 30, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Spent Fuel Pool Liner Design not Verified per Code The inspectors identified a finding of very low safety significance and a NCV of 10 CFR Part 50, Appendix B, Criterion III for the failure of the licensee's design control measures to provide for the verifying or checking of the adequacy of design of the spent fuel pool liner. Specifically, calculations involving the liner had not been verified or checked to ensure the design basis requirements of ASME Boiler and Pressure Vessel Code, Section III, Division II, were included. The licensee initiated AR 02690744 and initiated actions to restore compliance.
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| This performance deficiency was more than minor because if left uncorrected it could lead to a more significant safety concern if independent spent fuel storage installation loading was conducted. The inspectors determined the finding was of very low safety significance because each of the Barrier Integrity screening questions was answered no. The inspectors determined this issue was cross cutting in the Human Performance, Design Margin area because the licensee failed to carefully guard their design margins and ensure the margins were only changed through a systematic and rigorous process.
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| Inspection Report# : 2016003 (pdf)
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| Significance: Jun 30, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Page 12 of 18
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| 4Q/2016 Inspection Findings - Clinton Lack of Acceptance Criteria for Containment Visual Examinations The inspectors identified a finding of very-low safety significance and associated NCV of 10 CFR 50.55a(g)(4).
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| Specifically, the licensee failed to define acceptance criteria for containment visual examinations. Consequently, active containment liner degradation on a containment penetration was identified and returned to service without comparing to defined acceptance criteria. The licensee verified through visual examination that the liner thickness was marginally affected by the corrosion and documented this issue in the Corrective Action System in AR 02671728.
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| The inspectors determined that the failure to define and incorporate acceptance criteria in the containment visual examination procedure as required by 10 CFR 50.55a(g)(4) was a performance deficiency. The inspectors determined that this issue was more than minor in accordance with IMC 0612, Appendix B, Issue Screening, dated September 7, 2012, because the inspectors answered yes to the more than minor question If left uncorrected, would the performance deficiency have the potential to lead to a more significant safety concern in that active containment penetration degradation may not be properly evaluated and/or promptly corrected. Specifically, the inspectors were concerned that without acceptance standards, unacceptable containment degradation may be returned to service and adversely affect containment leakage or structural integrity. The inspectors determined this finding was of very-low safety significance (Green) based on answering no to Questions B.1 and B.2 of the Exhibit 3, Barrier Integrity Screening Questions, in IMC 0609, Attachment A, The Significance Determination Process (SDP) for Findings At-Power, issued on June 19, 2012. Specifically, the inspectors answered no to the screening question associated with an actual open pathway (e.g., breach) in the containment and no to the question associated with reduction in function of hydrogen igniters in containment. A subsequent visual examination performed by the licensee confirmed only marginal degradation of the liner thickness.
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| The inspectors determined that this finding had a cross-cutting aspect in the area of human performance in the aspect of consistent process, where individuals use a consistent, systematic approach to make decisions. Specifically, the lack of acceptance criteria allowed various interpretations for disposing of identified conditions that were inconsistent.
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| (H.13)
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| Inspection Report# : 2016002 (pdf)
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| Significance: Mar 31, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Identify a Degraded Safety-Related Support
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| . The inspectors identified a finding of very low safety significance and an associated non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for the failure to identify a condition adverse to quality.
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| Specifically, the licensee failed to identify that a safety-related support associated with control room ventilation B was degraded to the point it no longer conformed to the seismic analysis and required an evaluation to determine whether it was still capable of performing its safety function during a seismic event. This issue was entered into the licensees CAP as AR 2639317. The licensees immediate corrective actions included performing an evaluation that concluded the remaining three supports would be able to withstand the stresses imposed during a seismic event and creating an action to update the seismic calculation to incorporate the evaluation performed for the degraded support.
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| The licensee also planned to re-apply a coating to the supports as well as research and install insulation that was more breathable to minimize moisture accumulation and preclude any further degradation.
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| The inspectors determined that the failure to identify a condition adverse to quality in accordance with 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, was more than minor because if left uncorrected it had the potential to lead to a more significant safety concern. Specifically, by failing to identify the support was degraded, and correct the condition, the loss of material due to corrosion could potentially progress to the point where the remaining supports would no longer be able to perform their safety function. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings at Power, issued June 19, 2012, the finding was screened against the Barrier Integrity Cornerstone and determined to be of very low safety significance because the finding did not represent a degradation of the barrier function of the control room Page 13 of 18
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| 4Q/2016 Inspection Findings - Clinton against radiological conditions or a smoke or toxic atmosphere. The inspectors determined this finding affected the cross-cutting area of problem identification and resolution, in the aspect of evaluation, which states, The organization thoroughly evaluates issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, the licensee failed to thoroughly evaluate the issue identified by the inspectors and therefore did not recognize the degradation on the supports constituted a condition adverse to quality.
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| Inspection Report# : 2016001 (pdf)
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| Significance: Mar 31, 2016 Identified By: Self-Revealing Item Type: NCV Non-Cited Violation Failure to Assess and Manage Risk Increase for a Proposed Maintenance Activity A self-revealed finding of very low safety significance and an associated non-cited violation of 10 CFR 50.65 (a)(4) was identified on January 20, 2016, due to the licensees failure to assess and manage the risk increase from a proposed maintenance activity. Specifically, the licensee failed to manage the risk associated with racking out the continuous containment purge (CCP) A breaker, which resulted in the loss of both CCP trains, and led to an increase in primary to secondary containment differential pressure which exceeded the TS value. The licensee entered this issue into their CAP as AR 02614832. The proposed corrective actions to address this issue included creating a checklist to ensure validation of initial conditions is performed and providing training that reinforces the need to properly screen work order tasks with the appropriate risk factors.
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| The inspectors determined that the failure to assess and manage the risk increase of a proposed maintenance activity, as required by 10 CFR 50.65 (a)(4), was more than minor because it was associated with the maintenance procedure quality attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, by not properly assessing the risk of racking out the CCP A breaker the licensee did not recognize the CCP B train would be impacted, which resulted in exceeding the TS value for primary to secondary containment differential pressure. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings at Power, issued June 19, 2012, the finding was screened against the Barrier Integrity Cornerstone and determined to be of very low safety significance because the finding did not represent an actual open pathway in the physical reactor containment, containment isolation system or heat removal components and it did not involve an actual reduction in function of hydrogen igniters in the reactor containment. The inspectors identified a cross-cutting aspect in the area of human performance, in the aspect of challenging the unknown, which states, individuals stop when faced with uncertain conditions; risks are evaluated and managed before proceeding. Specifically, when the licensee was preparing the work package for maintenance on the CCP system it was uncertain what activities had already been completed as part of a concurrent evolution. Instead of stopping and validating the configuration of plant equipment, assumptions were made, and the risk of the activity was not properly assessed or managed.
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| Inspection Report# : 2016001 (pdf)
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| Significance: Feb 11, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Perform Adequate Evaluation of Crane and Crane Support Structure Elements A finding of very low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors for the failure of the licensees design control measures to provide for the verifying or checking the adequacy of design of the fuel handling building crane and crane support structure elements. Specifically, calculations involving the crane trolley rails, crane rail clips, and crane rail clip bolts Page 14 of 18
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| 4Q/2016 Inspection Findings - Clinton had not been verified or checked to ensure the design basis requirements of American Society of Mechanical Engineers (ASME) NOG-1-2004; American Institute of Steel Construction (AISC), 7th Edition; and Updated Safety Analysis Report (USAR) Section 3.8.4.5 were included. The licensee documented these issues in its corrective action program and initiated actions to restore compliance.
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| The performance deficiency was determined to be more than minor because if left uncorrected the performance deficiency could lead to a more significant safety concern if independent spent fuel storage installation (ISFSI) loading was conducted. The inspectors determined the finding could be evaluated using the Significance Determination Process in accordance with Inspection Manual Chapter (IMC) 0609, The Significance Determination Process for Findings At-Power, Appendix A, Exhibit 3 - Barrier Integrity Screening Questions (Section D). Based on answering No to all the questions in Exhibit 3, Section D, the inspectors determined the finding to be of very low safety significance (Green). The inspectors identified a Human Performance, Design Margin (H.6) cross-cutting aspect associated with this finding. Specifically, the licensee failed to ensure the crane trolley rails, crane rail clips, and crane rail clip bolts reflected the intended design margins of the design and licensing basis.
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| Inspection Report# : 2016010 (pdf)
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| Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Security Although the Security Cornerstone is included in the Reactor Oversight Process assessment program, the Commission has decided that specific information related to findings and performance indicators pertaining to the Security Cornerstone will not be publicly available to ensure that security information is not provided to a possible adversary.
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| Other than the fact that a finding or performance indicator is Green or Greater-Than-Green, security related information will not be displayed on the public web page. Therefore, the cover letters to security inspection reports may be viewed.
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| Security Although the Security Cornerstone is included in the Reactor Oversight Process assessment program, the Commission has decided that specific information related to findings and performance indicators pertaining to the Security Cornerstone will not be publicly available to ensure that security information is not provided to a possible adversary.
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| Other than the fact that a finding or performance indicator is Green or Greater-Than-Green, security related information will not be displayed on the public web page. Therefore, the cover letters to security inspection reports Page 15 of 18
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| 4Q/2016 Inspection Findings - Clinton may be viewed.
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| Security Although the Security Cornerstone is included in the Reactor Oversight Process assessment program, the Commission has decided that specific information related to findings and performance indicators pertaining to the Security Cornerstone will not be publicly available to ensure that security information is not provided to a possible adversary.
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| Other than the fact that a finding or performance indicator is Green or Greater-Than-Green, security related information will not be displayed on the public web page. Therefore, the cover letters to security inspection reports may be viewed.
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| Miscellaneous Significance: N/A Jun 30, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Update the Updated Safety Analysis Report (USAR) - Peak Suppression Pool Temperature The inspectors identified a Severity Level IV NCV of 10 CFR 50.71(e), Periodic Update of the [Final Safety Analysis Report] FSAR, for the licensees failure to update the USAR after updating a Safety Analysis Calculation.
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| Specifically, the licensee did not update the USAR Section A3.8.3.1 and Table 15.2.9-1 to coincide with the most recent updates to the accident analysis of record. The licensee initiated AR 2664276 to document the discrepancy in the peak suppression pool temperature throughout the USAR and initiated actions to revise FSAR Section A3.8.3.1 and Table 15.2.9-1 to coincide with the most recent revision to EPU-T0400.
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| The inspectors determined that the failure to update the USAR in accordance with 10 CFR 50.71(e), Periodic Update of the FSAR, with the most accurate version of calculated peak suppression pool temperature during an accident was a performance deficiency. The performance deficiency was determined to be minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012; however, the reactor oversight programs significance determination process does not specifically consider the regulatory process impact in its assessment of licensee performance, therefore, it was necessary to address this violation which impeded the NRCs ability to regulate using traditional enforcement to adequately deter non-compliance. The inspectors reviewed this issue in accordance with IMC 0612 and the NRC Enforcement Policy. Violations of 10 CFR 50.71(e) are dispositioned using the traditional enforcement process because they are considered to be violations that potentially impede or impact the regulatory process. The inspectors reviewed Section 6.1.d.3 of the NRC Enforcement Policy and determined this violation was Severity Level IV because the licensees failure to update the USAR as required by 10 CFR 50.71(e) had not yet resulted in any unacceptable change to the facility or procedures. No cross cutting aspect was assigned because traditional enforcement violations are not assessed for cross cutting aspects.
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| Inspection Report# : 2016002 (pdf)
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| Significance: N/A Jun 30, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Update the Updated Safety Analysis Report (USAR) - Condensate and Feedwater System The inspectors identified a Severity Level IV NCV of Title 10 Code of Federal Regulations (CFR) 50.71(e), Periodic Update of the FSAR, for the licensees failure to update the FSAR after implementation of license amendment 149, for extended power uprate. Specifically, the licensee did not update USAR Section 10.4.7.1.2 Performance Requirements, for the condensate and feedwater system with the design requirements for a reactor thermal power Page 16 of 18
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| 4Q/2016 Inspection Findings - Clinton rating of 3473 MWt. The licensee entered the issue into their CAP as AR 02656128 and is preparing a technical change package to update the USAR.
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| The inspectors determined that the failure to update the USAR in accordance with 10 CFR 50.71(e), Periodic Update of the FSAR, with the design requirements for the condensate and feedwater system for a reactor thermal power rating of 3473 MWt was a performance deficiency. The performance deficiency was determined to be minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012; however, the reactor oversight programs significance determination process does not specifically consider the regulatory process impact in its assessment of licensee performance, therefore, it was necessary to address this violation which impeded the NRCs ability to regulate using traditional enforcement to adequately deter non-compliance. The inspectors reviewed this issue in accordance with NRC IMC 0612 and the NRC Enforcement Policy.
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| Violations of 10 CFR 50.71(e) are dispositioned using the traditional enforcement process because they are considered to be violations that potentially impede or impact the regulatory process. The inspectors reviewed Section 6.1.d.3 of the NRC Enforcement Policy and determined this violation was Severity Level IV because the licensees failure to update the USAR as required by 10 CFR 50.71(e) had not yet resulted in any unacceptable change to the facility or procedures. No cross cutting aspect was assigned because traditional enforcement violations are not assessed for cross cutting aspects.
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| Inspection Report# : 2016002 (pdf)
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| Significance: N/A Mar 31, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Report a Condition that Could Have Prevented Fulfillment of a Safety Function The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.72(b)(3)(v) for failing to report an event or condition, that at the time of discovery could have prevented the fulfillment of a safety function, to the NRC within eight hours. Specifically, control room operators placed both divisions of reactor water cleanup differential flow instruments in bypass, which rendered the instruments inoperable and resulted in a loss of the isolation function.
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| The licensee entered this issue into the CAP as AR 02645140 and created an action to submit an licensee event report under 10 CFR 50.73(a)(2)(v).
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| The inspectors determined that the failure to report an event or condition, that at the time of discovery could have prevented the fulfillment of a safety function, to the NRC within 8 hours as required by 10 CFR 50.72(b)(3)(v) was a performance deficiency. The inspectors reviewed this issue in accordance with IMC 0612 and the Enforcement Manual. Violations of 10 CFR 50.72 are dispositioned using the traditional enforcement process because they are considered to be violations that potentially impede or impact the regulatory process. The inspectors reviewed Section 6.9.d.9 of the NRC Enforcement Policy and determined this violation was Severity Level IV because the licensees failure to make the report, as required by 10 CFR 50.72, did not cause the NRC to reconsider a regulatory position or undertake substantial further inquiry. No cross-cutting aspect was assigned because cross-cutting aspects are not assigned to traditional enforcement only violations.
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| Inspection Report# : 2016001 (pdf)
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| Significance: N/A Mar 31, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Report Condition Prohibited by Technical Specifications The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.73(a)(2)(i)(B) for failing to report to the NRC, within 60 days of discovery, a condition prohibited by the plants TS. Specifically, the licensee failed to notify the NRC of two instances where they failed to comply with TS 3.3.6.1 and TS 3.3.6.2 and enter the limiting condition for operation action statements when required. The licensee entered this issue into their CAP as AR 02619114 and subsequently issued a licensee event report on March 16, 2016.
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| Page 17 of 18
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| 4Q/2016 Inspection Findings - Clinton The inspectors determined that the failure to report a condition prohibited by the plants TS as required by 10 CFR 50.73(a)(2)(i)(B), within 60 days of discovery, was a performance deficiency. The inspectors reviewed this issue in accordance with IMC 0612 and the Enforcement Manual. Violations of 10 CFR 50.73 are dispositioned using the traditional enforcement process because they are considered to be violations that potentially impede or impact the regulatory process. The inspectors reviewed Section 6.9.d.9 of the NRC Enforcement Policy and determined this violation was Severity Level IV because the licensees failure to make the report, as required by 10 CFR 50.73, did not cause the NRC to reconsider a regulatory position or undertake substantial further inquiry. No cross-cutting aspect was assigned because cross-cutting aspects are not assigned to traditional enforcement only violations.
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| Inspection Report# : 2016001 (pdf)
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| Last modified : February 01, 2017 Page 18 of 18
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| NRC: Clinton - Quarterly Plant Inspection Findings Home > Nuclear Reactors > Operating Reactors > Reactor Oversight Process > Plant Summaries > Clinton > Quarterly Plant Inspection Findings Clinton - Quarterly Plant Inspection Findings 2Q/2017 - Plant Inspection Findings On this page:
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| * Initiating Events
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| * Mitigating Systems
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| * Barrier Integrity
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| * Emergency Preparedness
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| * Occupational Radiation Safety
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| * Public Radiation Safety
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| * Security Initiating Events Significance: Mar 31, 2017 Identified By: Self-Revealing Item Type: NCV Non-Cited Violation FAILURE TO DEVELOP AND REVIEW A WORKER TAG OUT The inspectors documented a self-revealed finding of very low safety significance and associated non-cited violation of Technical Specification 5.4.1, "Procedures," for the licensee's failure to develop and review a worker tag out in accordance with station procedure OP-AA-109-10, "Clearance and Tagging," Revision 12. Specifically, the licensee failed to identify the effect of a worker tag out on the in-service steam jet air ejector suction valve, which caused condenser vacuum to degrade resulting in the operators entering the off normal procedure for loss of condenser vacuum. The licensee entered this issue into their corrective action program as action request (AR) 03980495. As corrective actions, the operations department issued a standing order to require worker tag outs to be challenged by a second senior reactor operator.
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| The performance deficiency was determined to be more than minor because it impacted the Initiating Events cornerstone attribute of configuration control and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during power operations. Specifically, the failure to properly develop the worker tag out caused the condenser vacuum to degrade, challenging the operators to quickly diagnose the issue and take action to avoid a turbine trip. The finding was screened against the Initiating Events cornerstone and determined to be of very low safety significance because it did not cause a reactor trip or the loss of mitigation equipment relied upon to transition the plant from the onset of a trip to a stable shutdown condition. The inspectors determined that this finding affected the cross-cutting area of human performance in the aspect of avoid complacency, where individuals recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Individuals implement appropriate error reductions tools. Specifically, the operations department failed to implement appropriate error reduction tools such as questioning attitude and thorough work product reviews to ensure the worker tag out considered all potential effects to other plant equipment. [H.12]
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| Inspection Report# : 2017001 (pdf)
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| Page 1 of 12
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| NRC: Clinton - Quarterly Plant Inspection Findings Significance: Sep 30, 2016 Identified By: Self-Revealing Item Type: NCV Non-Cited Violation Failure to Prevent Recurrence of a Significant Condition Adverse to Quality The inspectors documented a self-revealing finding of very low safety significance and an NCV of 10 CFR 50, Appendix B, Criterion XVI for the licensee's failure to take corrective action to preclude repetition of a significant condition adverse to quality (SCAQ). After identifying IGSCC on main steam flex hoses in 2007 and concluding the leakage constituted a SCAQ, the licensee's corrective actions to prevent recurrence failed to prevent pressure boundary leakage at the same location in 2016. The licensee entered this issue into their corrective action program as AR 02670593. The affected hoses were replaced. The licensee is also developing a design change to address at least one of the three factors that contributes to IGSCC.
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| The inspectors determined that the licensee's failure to take corrective actions to prevent recurrence of an SCAQ was a performance deficiency and more than minor because if left uncorrected pressure boundary leakage could become a more significant concern. Specifically, pressure boundary leakage is not allowed by TS and any leakage requires the plant to be shutdown. The finding was screened as low safety significant because it did not result in exceeding the RCS leak rate for a small LOCA and did not affect systems used to mitigate a LOCA. No cross cutting aspect was assigned as the original issue occurred greater than three years ago and was not reflective of current performance.
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| Inspection Report# : 2016003 (pdf)
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| Mitigating Systems Significance: Apr 21, 2017 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Perform Required Surveillances on Multiple Fire Dampers (Section 1R05.2b)
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| Green. The inspectors identified a finding of very-low safety significance (Green), and an associated Non-Cited Violation of License Condition 2.C(f) for the licensee's failure to adequately implement surveillance procedures and work processes associated with fire barrier damper inspections. Specifically, the licensee failed to perform fire barrier damper inspections for 15 fire dampers once every 48 months (plus an additional 25 percent grace period) as required by the Fire Protection Program. The licensee entered the issue into their Corrective Action Program, and will inspect the fire barrier dampers during the next refueling outage.
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| The inspectors determined that the performance deficiency was more-than-minor because the licensee's failure to inspect the fire barrier dampers could result in not identifying degraded dampers which could affect their ability to prevent a fire from spreading from one fire area to another. The finding was of very-low safety significance because the failure to inspect the fire barrier dampers did not impact the plant's ability to reach and maintain safe-shutdown. The finding has a cross-cutting aspect in the area of Human Performance, Work Management because the licensee failed to execute a work order to inspect the fire dampers in accordance with the required frequency in Procedure CPS 9601.01 and instead improperly extended the frequency of the fire damper inspections. (Section 1R05.2b) [H.5]
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| Inspection Report# : 2017008 (pdf)
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| Significance: Mar 31, 2017 Identified By: NRC Item Type: NCV Non-Cited Violation PLANT BARRIER CONTROL PROGRAM FAILED TO COMPENSATE FOR AN IMPACTED FLOOD BARRIER Page 2 of 12
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| NRC: Clinton - Quarterly Plant Inspection Findings The inspectors identified a finding of very low safety significance and an associated non-cited violation of Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings,"
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| for the failure to implement the plant barrier control program for an impacted flood barrier. Specifically, the plant barrier impairment (PBI) permit, PBI-2017-02-003, for work on watertight door 1SD1-24, failed to identify the door as a flood barrier and that appropriate compensatory measures for 1SD1-24 being open for an extended period were identified or implemented in accordance with station procedure CC-AA-201, "Plant Barrier Control Program,"
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| Revision 11. The licensee entered this issue into their corrective action program as AR 03980495. The corrective actions in response to this violation were to identify appropriate compensatory measures for impairment of 1SD1-24 and incorporate them into the PBI log.
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| The performance deficiency was determined to be more than minor because it impacted the Mitigating Systems cornerstone attribute of protection against external events and adversely affected the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. With the flood barrier nonfunctional and without compensatory actions in place the residual heat removal (RHR) 'B' and RHR 'C' pumps were inoperable. The finding was screened against the Mitigating Systems cornerstone and the inspectors determined that the finding involved the loss or degradation of equipment or function specifically designated to mitigate a seismic, flooding or severe weather initiating event. The inspectors determined that the loss of this equipment or function by itself during the external initiating event would degrade one or more trains of a system that supports a risk significant system or function and would require a detailed risk evaluation. The senior reactor analyst (SRA) performed the detailed risk evaluation and concluded the finding was of very low safety significance. The inspectors determined that this finding affected the cross-cutting area of human performance in the aspect of conservative bias, where individuals use decision making practices that emphasize prudent choices over those that are simply allowable. Proposed actions are determined to be safe in order to proceed, rather than unsafe in order to stop. Specifically, during preparation of the PBI permit, the station PBI log was reviewed and actions for previous work associated with the watertight door were deemed acceptable even though the work on the door in those instances was different than the work being performed this time. [H.14]
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| Inspection Report# : 2017001 (pdf)
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| Significance: Mar 31, 2017 Identified By: NRC Item Type: NCV Non-Cited Violation FAILED TO VERIFY AN APPROPRIATE ALTERNATE METHOD OF DECAY HEAT REMOVAL The inspectors identified a finding of very low safety significance and an associated non-cited violation of 10 CFR 50.36(c)(2)(i), "Limiting conditions for operation", for failing to meet/follow the required actions for limiting condition for operation 3.9.9 and 3.4.10. Specifically, the operators failed to verify a credited alternate decay heat removal method that would satisfy the required action for the limiting condition for operation. The licensee entered this issue into their corrective action program as AR 03987440. The corrective actions in response to this violation were to identify appropriate alternate methods of decay heat removal and incorporate them into the shutdown safety management program utilized during plant outages.
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| The performance deficiency was determined to be more than minor because it impacted the Mitigating Systems cornerstone attribute of equipment performance and adversely affected the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences.
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| Specifically, with the operators failing to identify a credited alternate method of decay heat removal and taking credit for the inoperable but in service RHR shutdown cooling train, the actual available methods that could have been credited were not verified to ensure their availability to provide the required function. The finding was screened against the Mitigating Systems Screening questions and determined to be of very low safety significance because the answer to all of the applicable screening questions was "No." The inspectors determined that this finding affected the cross-cutting area of human performance in the aspect of conservative bias, where individuals use decision making practices Page 3 of 12
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| NRC: Clinton - Quarterly Plant Inspection Findings that emphasize prudent choices over those that are simply allowable. Proposed actions are determined to be safe in order to proceed, rather than unsafe in order to stop. Specifically, the senior reactor operators at the station had historically credited inoperable RHR shutdown cooling subsystems as their own alternate decay heat remove method because they believed it was allowable without determining that it was safe in order to proceed. [H.14]
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| Inspection Report# : 2017001 (pdf)
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| Significance: Mar 31, 2017 Identified By: Self-Revealing Item Type: NCV Non-Cited Violation FAILURE TO PERFORM MAINTENANCE ON RESIDUAL HEAT REMOVAL PUMP 'C' BREAKER IN ACCORDANCE WITH PROCEDURES The inspectors documented a self-revealed finding of very low safety significance and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures and Drawings," for the licensee's failure to perform maintenance on a safety related breaker in accordance with station procedure Clinton Power Station (CPS) 8410.12C001, "Westinghouse DHP Circuit Breaker Checklist," Revision 7. Specifically, the licensee failed to ensure the remaining travel on the latch check switch for the RHR 'C' pump breaker was within the acceptable range resulting in the RHR 'C' pump failing to start. The licensee entered this issue into their corrective action program as AR 03949655. The corrective actions taken by the licensee included providing coaching to the involved individuals as well as changing the procedure to include a block to record the latch check switch over travel.
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| The performance deficiency was determined to be more than minor because it impacted the Mitigating Systems cornerstone attribute of equipment performance and adversely affected the cornerstone objective of ensuring the availability, capability, and reliability of equipment that responds to initiating events. Specifically, the performance deficiency adversely impacted the operability of the RHR 'C' pump. The inspectors reviewed the Mitigating Systems screening questions and determined a detailed risk evaluation was required because question A.3 was answered yes.
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| The SRA performed the detailed risk evaluation and concluded the finding was of very low safety significance. The inspectors determined that this finding affected the cross-cutting area of human performance in the aspect of resources, where leaders ensure that personnel, equipment, procedures, and other resources are available and adequate to support nuclear safety. Specifically, the organization failed to ensure the procedure step included a block for recording the latch check switch over travel value, which led to confusion on whether the value was required to be recorded and ultimately resulted in a failure to perform the step as written. [H.1]
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| Inspection Report# : 2017001 (pdf)
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| Significance: Dec 31, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Demonstrate the Condition or Flood Seals was being Effectively Controlled The inspectors identified a finding of very low safety significance and an associated non-cited violation of 10 CFR 50.65(a)(2), "Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," for the licensee's failure to demonstrate that the condition of flood seals was being effectively controlled through appropriate preventive maintenance such that the flood seals remained capable of performing their intended function. Specifically, the licensee failed to visually inspect flood seals per ER-AA-450, "Structures Monitoring" and ER-CL-450-1007, "Clinton Surveillance Inspection Program for Seals." As corrective actions, the licensee planned to visually inspect all accessible flood seals and generate an evaluation for inaccessible seals. In addition, the licensee planned to modify ER-AA-450 to clarify the frequency of flood seal inspection.
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| The inspectors determined the licensee's failure to demonstrate that the condition of flood seals was being effectively Page 4 of 12
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| NRC: Clinton - Quarterly Plant Inspection Findings controlled by visual inspection, per ER-AA-450 and ER-CL-450-1007, was a performance deficiency. The performance deficiency was determined to be more than minor in accordance with IMC 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," dated September 7, 2012, because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, failure to monitor the condition of the flood sales in a manner sufficient to provide reasonable assurance they were capable of fulfilling the intended safety functions could adversely affect multiple mitigating systems in the event of a flood or line break. Using IMC 0609, Attachment 4, "Initial Characterization of Findings," issued October 7, 2016, and Appendix A, "The Significance Determination Process for Findings at Power," issued June 19, 2012, the finding was screened against the Mitigating Systems cornerstone and determined to be of very low safety significance because the inspectors answered no to the question "does the finding involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding, or severe weather initiating event?" The inspectors determined this finding affected the cross-cutting area of Problem Identification and Resolution, in the aspect of identification, where the organization implements a corrective action program with a low threshold for identifying issues. Individuals identify issues completely, accurately, and in a timely manner in accordance with the program. Specifically, the licensee failed to identify the flood seals still had not been inspected when they performed the Maintenance Rule Program - Structures Monitoring Assessment which credited ER-CL-450-1007 in early 2014.
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| Inspection Report# : 2016004 (pdf)
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| Significance: Dec 01, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Promptly Identify that the Incapability of the RHR Design to Support TS Operability Requirements Was a CAQ (Section 4OA2.b(1))
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| Green: The team identified a finding of very-low safety significance (Green) and an associated NCV of Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion XVI, "Corrective Action," for the licensee failure to promptly identify that the incapability of the residual heat removal (RHR) design to support Technical Specifications (TS) operability requirements was a condition adverse to quality. Specifically, when reactor water temperature was greater than 150 degrees Fahrenheit, RHR could not be realigned from shutdown cooling mode of operations to provide the TS required functions of the emergency core cooling system, suppression pool cooling, containment spray, and feedwater leakage control system. The licensee captured this issue in their Corrective Action Program (CAP) as Action Request (AR) 02742439 and AR 03948042, and planned to submit a License Amendment Request to align TS requirements with the design capabilities.
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| The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of design control and adversely affected the associated cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the performance deficiency resulted in voluntarily declaring TS functions inoperable when performing shutdown cooling operations, which did not ensure the associated mitigating systems availability or capability to respond to an initiating event. The team determined that this finding was of very low safety significance (Green). Specifically, there were no known instances where the finding: (1) represented a loss of system safety function; (2) represented an actual loss of safety function of at least a single train or two separate safety systems out of service for greater than their TS allowed outage time; (3) involved non-TS trains of equipment; (4) involved a degradation of a functional RHR auto-isolation on low reactor vessel level; (5) impacted external event protection; or (6) involved fire brigade issues. The team did not identify a cross cutting aspect associated with this finding because it did not reflect current licensee performance since the performance deficiency occurred more than 3 years ago. (Section 4OA2.b(1))
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| Page 5 of 12
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| NRC: Clinton - Quarterly Plant Inspection Findings Inspection Report# : 2016009 (pdf)
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| Significance: Sep 30, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Exceeded Technical Specification Allowed Outage Time for Main Turbine Bypass System The inspectors identified a finding of very low safety significance and an associated NCV of Technical Specification 3.7.6, "Main Turbine Bypass System" for the licensee's failure to meet the limiting conditions for operation and complete the associated required actions after making a deficient change to the turbine bypass valve surveillance testing frequency. Specifically, with the main turbine bypass system inoperable and without the Core Operating Limits Report (COLR) limits for thermal power, minimum critical power ratio (MCPR), and linear heat generation rate (LGHR) with the main turbine by pass system inoperable applied, thermal power was not reduced to less than 21.6 percent of rated thermal power within six hours. The licensee entered this issue into their corrective action program as AR 02690657.
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| The licensee restored compliance by applying the COLR limits for reactor thermal power, MCPR and LGHR.
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| The inspectors determined the failure to meet the limiting conditions for operation and complete the associated required actions prior to the end of the specified completion times was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was screened against the Mitigating Systems cornerstone and determined to be of very low safety significance because all of the associated questions in IMC 0609, Appendix A, were answered no. The inspectors determined this finding affected the cross-cutting area of human performance, in the aspect of change management, where leaders use a systematic process for evaluating and implementing change so that nuclear safety remains the overriding priority because the licensee's change management process was not fully utilized by senior management when evaluating and implementing a change to the turbine bypass valve surveillance testing frequency. (H.3)
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| Inspection Report# : 2016003 (pdf)
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| Significance: Sep 30, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Perform a 50.59 Screening for Changing the Frequency of Exercising the Turbine Bypass Valves The inspectors identified a Severity Level IV NCV of 10 CFR 50.59 4(d)(1), "Changes, Tests, and Experiments," and an associated Green finding for the licensee's failure to perform a written evaluation which provided the bases for determining that changing the turbine bypass valve surveillance testing frequency from every 31 days, as specified in the Updated Safety Analysis Report, to once a year did not require a license amendment. The licensee has entered this issue into their corrective action program as AR 02720163. The licensee is currently evaluating the issue in accordance with their procedure for changes to the facility.
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| The inspectors determined that the licensee's failure to perform a written evaluation to provide the basis for the determination that a change to the facility, a change to a procedure, or a change to a test or experiment did not require a license amendment was a performance deficiency. The performance deficiency was more than minor in accordance with IMC 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," dated September 7, 2012, because, it was associated with the procedure quality attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using IMC 0609, Attachment 4, "Initial Characterization of Page 6 of 12
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| NRC: Clinton - Quarterly Plant Inspection Findings Findings," and Appendix A, "The Significance Determination Process for Findings At-Power," issued June 19, 2012, the finding was screened against the Mitigating Systems cornerstone and determined to be of very low safety significance because all of the associated questions in IMC 0609, Appendix A, were answered no. Violations of 10 CFR 50.59 are dispositioned using the traditional enforcement process because they are considered to be violations that potentially impede or impact the regulatory process. The inspectors reviewed Section 6.1.d.2 of the NRC Enforcement Policy and determined this violation was Severity Level IV because the resulting changes were evaluated by the SDP as having very low safety significance. The inspectors determined this finding affected the cross-cutting area of human performance, in the aspect of consistent process, where individuals use a consistent, systematic approach to make decisions. The licensee made a decision to proceed with implementation of a change to the turbine bypass valve surveillance testing frequency after a plant oversight committee review in lieu of following their consistent, systematic process for evaluating changes to the USAR. (H.13)
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| Inspection Report# : 2016003 (pdf)
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| Significance: Aug 09, 2016 Identified By: NRC Item Type: FIN Finding Failure to have hose configurations that were verified to be able to ensure a timely and successful implementation of a FLEX strategy Green. Two examples of a finding of very low safety significance was identified by the inspectors for the licensee's failure to have hose configurations that were verified to be able to ensure a timely and successful implementation of a flexible response (FLEX) strategy. Specifically, the licensee did not ensure through evaluations, calculations, analyses or any other means that the strategy for maintaining core cooling, containment heat removal and Spent Fuel Pool (SFP) cooling during a Beyond-Design-Basis External Event (BDBEE) flooding scenario would be capable of fulfilling its function. No violation of NRC requirements were identified.
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| The performance deficiency is more than minor because it was associated with the mitigating systems cornerstone objective attribute of protection against external factors, specifically the BDBEE flood hazard, and it adversely affected the cornerstone attribute of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Issues identified through TI-191 are evaluated through a cross-regional panel using IMC 0609, Appendix M, "Significance Determination Process Using Qualitative Criteria," as informed by draft Appendix O, "Post Fukushima Mitigation Strategies Significance Determination Process." The finding was determined to be of very low safety significance (Green). The inspectors concluded that the cause of the finding involved a cross-cutting component in the Human Performance area of Design Margins because the organization did not ensure the selected strategy contained the required verification that it could be successfully implemented. [H.6]
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| Inspection Report# : 2016007 (pdf)
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| Barrier Integrity Significance: Dec 01, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Non Conservative Control Room Radiological Habitability Assessment (Section 1R21.3.b(1))
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| Green. The team identified a finding of very-low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for the licensee failure to use a technically appropriate analytical methodology in the control room radiological habitability calculation. Specifically, the licensee used a methodology that inappropriately characterized the control room heating, ventilation and air conditioning (HVAC) system outside air Page 7 of 12
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| NRC: Clinton - Quarterly Plant Inspection Findings intake design resulting in a calculated control room dose following a loss of coolant accident that exceeded the applicable limit. The licensee captured this issue in their CAP as AR 02742442, completed an operability evaluation, and issued an NRC event notification.
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| The performance deficiency was determined to be more-than-minor because it was associated with the Barrier Integrity cornerstone attribute of design control and affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the performance deficiency resulted in the control room expected dose following a loss of coolant accident to exceed the applicable limits prompting an operability evaluation. The finding screened as of very-low safety significance (Green) because it only represented a degradation of the radiological barrier function provided for the control room.
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| Specifically, the finding did not affect the control room barrier function against smoke or a toxic atmosphere. The team did not identify a cross-cutting aspect associated with this finding because it was not confirmed to reflect current performance due to the age of the performance deficiency. Specifically, the affected calculations were performed more than 3 years ago. (Section 1R21.3.b(1))
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| Inspection Report# : 2016009 (pdf)
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| Significance: Dec 01, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Scope SFP Temperature and Level Instruments into the Maintenance Rule Program (Section 1R21.3.b(2))
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| Green. The team identified a finding of very-low safety significance (Green) and an associated NCV of Paragraph (b)
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| (2)(i) of 10 CFR 50.65, "Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," for the licensee failure to scope non-safety related mitigating structure, systems, and components (SSCs) used within an emergency operating procedure (EOP) into Maintenance Rule Program. Specifically, an EOP used spent fuel pool (SFP) low-level and high-temperature parameters as distinct entry criteria but the associated components were not included in the scope of the Maintenance Rule Program. The licensee captured the team concerns in their CAP as AR 02736193, performed an extent of condition to identify any other SSC addition to the EOPs requiring them to be added to the Maintenance Rule Program scope, and initiated plans to incorporate the affected SSCs into the Maintenance Rule Program scope.
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| The performance deficiency was determined to be more-than-minor because it was associated with the Barrier Integrity cornerstone attribute of SSC performance and affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, a key aspect of the Maintenance Rule is to ensure that maintenance activities are performed in a manner that provide reasonable assurance that SSCs within its scope perform reliably and are capable of providing their intended Maintenance Rule function(s). In the case of the SFP temperature instruments, the licensee was not performing preventive maintenance to ensure that degradation, such as instrument drift, did not adversely affect their ability to detect and alarm EOP entry conditions such that mitigating actions could be implemented to preserve secondary containment. The finding screened as of very-low safety significance (Green) because it only represented a degradation of the radiological barrier function provided for the control room. Specifically, the finding did not cause SFP temperature to exceed the maximum analyzed limit, a detectible release of radionuclides, water inventory to decrease below the analyzed limit, or an adverse effect to the SFP neutron absorber or fuel loading pattern. The team determined that the finding had a cross cutting aspect in the area of human performance because the licensee did not use a systematic process for evaluating and implementing changes when updating the affected EOP in 2015. (Section 1R21.3.b(2)) [H.3]
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| Inspection Report# : 2016009 (pdf)
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| Page 8 of 12
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| NRC: Clinton - Quarterly Plant Inspection Findings Significance: N/A Dec 01, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Amend the UFSAR Indicating Choice to Comply with 10 CFR 50.68(b) (Section 1R21.3.b(3))
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| The team identified a Severity Level-IV NCV of 10 CFR 50.68, "Criticality Accident Requirements," Paragraph (b)(8),
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| for the licensee failure to amend the Updated Final Safety Analysis Report (UFSAR) to indicate they chose to comply with 10 CFR 50.68(b). Specifically, in 2005, the licensee chose to comply with 10 CFR 50.68(b) but did not amend the UFSAR following the issuance of the associated license amendment. The licensee captured this issue in their CAP as AR 02741851, reasonably confirmed compliance with 10 CFR 50.68(b) requirements (1) through (7) was maintained, and initiated plans to update the UFSAR to specifically indicate that Clinton Power Station chose to comply with 10 CFR 50.68(b).
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| The Significance Determination Process does not specifically consider the impact to the regulatory process in its assessment of licensee performance. Therefore, it was necessary to address this violation, which potentially impacts the NRC's ability to regulate, using traditional enforcement to adequately deter non-compliance. Specifically, failure to update the UFSAR challenges the regulatory process because it serves as a reference document used, in part, for recurring safety analyses, evaluating License Amendment Request, and in preparation for and conduct of inspection activities. The team determined the traditional enforcement violation was a Severity Level-IV violation in accordance with Section 6.1.d.3 of the Enforcement Policy because the un-updated UFSAR had not been used to evaluate a facility or procedure change that resulted in a condition evaluated as having low-to-moderate or greater safety significance by the Significance Determination Process. However, it had a material impact on safety or licensed activities. Specifically, the un-updated UFSAR could be used to perform evaluations of facility or procedure changes, which would have the potential to result in unacceptable conditions and/or regulatory decisions. Traditional enforcement violations are not assessed for cross-cutting aspects. (Section 1R21.3.b(3))
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| Inspection Report# : 2016009 (pdf)
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| Significance: Dec 01, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Verify the Adequacy of Design Assumptions Related to Time Critical Operator Actions (Section 1R21.6.b(1))
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| Green. The team identified a finding of very-low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for the licensee failure to verify the adequacy of design assumptions related to time critical operator actions made in calculations associated with the control room HVAC and RHR emergency SFP cooling functions. Subsequently, it was determined that operators did not fully understand the control room HVAC system operational demands and that the operational assumptions of the RHR emergency SFP cooling design were unrealistic. The licensee captured these issues into the CAP as AR 02739012, AR 03943566, and AR 02741909; reasonably demonstrated that SFP makeup sources would be available to cope with a prolonged loss of SFP cooling; conducted operator training; and provided refined procedural guidance to ensure the control room HVAC system would be operated consistent with the design assumptions.
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| The performance deficiency was determined to be more-than-minor because it was associated with the Barrier Integrity cornerstone attribute of human performance and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events.
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| Specifically, the pilot validations of the control room HVAC system operational assumptions demonstrated a significant reduction in margin due to, in part, a lack of operator understanding of the operational assumptions.
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| Additionally, a preliminary review of procedures associated with SFP cooling and RHR determined the operational assumptions of the calculation related to RHR emergency SFP cooling were not bounding. The team determined that this finding was of very low safety significance (Green). Specifically, the control room HVAC system finding example only represented a degradation of the radiological barrier function provided for the control room in that it did not affect Page 9 of 12
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| NRC: Clinton - Quarterly Plant Inspection Findings the control room barrier function against smoke or a toxic atmosphere. In addition, the finding example related to emergency SFP cooling did not cause SFP temperature to exceed the maximum analyzed limit, a detectible release of radionuclides, water inventory to decrease below the analyzed limit, or an adverse effect to the SFP neutron absorber or fuel loading pattern. The team determined that the finding had a cross-cutting aspect in the area of Human Performance because the operation and engineering organizations did not effectively communicate and coordinate their respective roles in developing the control room HVAC system validation in a manner that supported nuclear safety. (Section 1R21.6.b(1)) [H.4]
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| Inspection Report# : 2016009 (pdf)
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| Significance: Dec 01, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Follow the Operability Determination Process Following the Identification of a Control Room HVAC System Design Issue (Section 4OA2.b(2))
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| Green: The team identified a finding of very-low safety significance (Green), and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, "Instruction, Procedures, and Drawings," for the licensee failure to follow the operability evaluation procedure after the identification of a significant design error associated with the control room HVAC system. Specifically, the licensee did not identify the affected safety function, and promptly restore or confirm system operability. The licensee captured these issues into the CAP as AR 03948266 and performed a preliminary engineering evaluation using another alternative analytical methodology that reasonably determined the control room HVAC system remained operable.
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| The performance deficiency was determined to be more-than-minor because it was associated with the Barrier Integrity cornerstone attribute of human performance and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events.
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| Specifically, the performance deficiency resulted in a condition where reasonable doubt on the operability of the control room HVAC system remained following the identification of a significant design error. The finding screened as of very low safety significance (Green) because it only represented a degradation of the radiological barrier function provided for the control room. Specifically, the finding did not affect the control room barrier function against smoke or a toxic atmosphere. The team identified that the finding had a cross-cutting aspect in the area of Human Performance because the licensee did not provide training to maintain a knowledgeable workforce that would facilitate an adequate implementation of the operability evaluation process following the identification of a non-conforming design-related issue. (Section 4OA2.b(2)) [H.9]
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| Inspection Report# : 2016009 (pdf)
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| Significance: Sep 30, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Scope Fuel Building Ventilation Pressure Control into Maintenance Rule The inspectors identified a finding of very low safety significance and an NCV of 10 CFR 50.65 (b) for the licensee's failure to scope a non-safety related structure, system and component (SSC), whose function is used in one or more Emergency Operating Procedures (EOP) and whose failure could cause actuation of a safety-related system, into maintenance rule. Specifically, the licensee failed to scope the non-safety related fuel building pressure control function into their maintenance rule program. The licensee has entered this issue into their corrective action program as AR 02716300. The licensee is scoping the pressure control function of fuel building ventilation into maintenance rule.
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| The inspectors determined that the licensee's failure to scope a non-safety related system whose function is used in one Page 10 of 12
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| NRC: Clinton - Quarterly Plant Inspection Findings or more EOPs and whose failure caused the actuation of a safety-related system into maintenance rule was a performance deficiency. The performance deficiency was determined to be more than minor because it affects the SSC and barrier performance attribute of the Barrier Integrity cornerstone and adversely affects the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. The finding was determined to be of very low safety significance because the inspectors answered yes to the question "does the finding only represent a degradation of the radiological barrier function provided for the control room, or auxiliary building, or spent fuel pool, SBGT system (BWR)?". The inspectors determined this finding affected the cross-cutting area of human performance, in the aspect of avoiding complacency, because the licensee identified water intrusion of the fuel building pressure sensing line was a longstanding, latent, known problem and failed to recognize and appropriately challenge how the function was scoped into maintenance rule. (H.12)
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| Inspection Report# : 2016003 (pdf)
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| Significance: Sep 30, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Spent Fuel Pool Liner Design not Verified per Code The inspectors identified a finding of very low safety significance and a NCV of 10 CFR Part 50, Appendix B, Criterion III for the failure of the licensee's design control measures to provide for the verifying or checking of the adequacy of design of the spent fuel pool liner. Specifically, calculations involving the liner had not been verified or checked to ensure the design basis requirements of ASME Boiler and Pressure Vessel Code, Section III, Division II, were included. The licensee initiated AR 02690744 and initiated actions to restore compliance.
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| This performance deficiency was more than minor because if left uncorrected it could lead to a more significant safety concern if independent spent fuel storage installation loading was conducted. The inspectors determined the finding was of very low safety significance because each of the Barrier Integrity screening questions was answered no. The inspectors determined this issue was cross cutting in the Human Performance, Design Margin area because the licensee failed to carefully guard their design margins and ensure the margins were only changed through a systematic and rigorous process.
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| Inspection Report# : 2016003 (pdf)
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| Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Security The security cornerstone is an important component of the ROP, which includes various security inspection activities the NRC uses to verify licensee compliance with Commission regulations and thus ensure public health and safety. The Commission determined in the staff requirements memorandum (SRM) for SECY-04-0191, "Withholding Sensitive Unclassified Information Concerning Nuclear Power Reactors from Public Disclosure," dated November 9, 2004, that specific information related to findings and performance indicators associated with the security cornerstone will not be publicly available to ensure that security-related information is not provided to a possible adversary. Security inspection report cover letters will be available on the NRC Web site; however, security-related information on the details of inspection finding(s) will not be displayed.
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| Miscellaneous Page 11 of 12
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| NRC: Clinton - Quarterly Plant Inspection Findings Significance: N/A Dec 31, 2016 Identified By: Self-Revealing Item Type: NCV Non-Cited Violation Dry Cask Storage Procedures were not Adequate to Ensure Correct Field Configuration A self-revealed violation of very low safety significance (Severity Level IV) of 10 CFR 72.150, "Instructions, Procedures, and Drawings," was identified for the failure of the licensee to ensure that ISFSI procedures contained the appropriate level of detail for the circumstances such that important loading activities would be satisfactorily accomplished. Specifically, procedure HPP-2226-200, Revision 0, "MPC Loading at Clinton," was not adequate to ensure that the Multi-Purpose Canister (MPC) was correctly oriented in the transfer cask (HI-TRAC) and procedure HPP-2226-300, Revision 4, "MPC Sealing at Clinton," was not adequate to ensure that two thermocouples were appropriately installed during the hydrostatic test of the MPC.
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| The licensee documented these issues in its corrective action program and took timely corrective actions.
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| The violation was determined to be of more than minor significance using Inspection Manual Chapter (IMC) 0612, "Power Reactor Inspection Reports," Appendix E, "Examples of Minor Issues." Example 4e is applicable to this violation in that the MPC was incorrectly oriented in the transfer cask and then loaded with spent fuel in this incorrect configuration. Example 4b is also applicable to this violation in that unexpected leakage occurred during the hydrostatic test as a result of the failure to install the thermocouples. The violation screened as having very low safety significance. Cross cutting aspects are not assigned to traditional enforcement violations.
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| Inspection Report# : 2016004 (pdf)
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| Current data as of : August 03, 2017 Page Last Reviewed/Updated Wednesday, August 10, 2016 Page 12 of 12
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| NRC: Clinton - Quarterly Plant Inspection Findings Page 1 of 16 Home > Nuclear Reactors > Operating Reactors > Reactor Oversight Process > Plant Summaries> Clinton > Quarterly Plant Inspection Findings Clinton - Quarterly Plant Inspection Findings 2Q/2017 - Plant Inspection Findings On this page:
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| * Initiating Events
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| * Mitigating Systems
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| * Barrier Integrity
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| * Emergency Preparedness
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| * Occupational Radiation Safety
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| * Public Radiation Safety
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| * Security Initiating Events Significance: Jun 30, 2017 Identified By: Self-Revealing Item Type: NCV Non-Cited Violation UNEXPECTED START OF THE DIVISION 3 EMERGENCY DIESEL GENERATOR The inspectors documented a self-revealed finding of very low safety significance and an associated non-cited violation of Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," for the licensee's failure to follow steps in Work Order (WO) 04640788 while performing troubleshooting on blown power transformer fuses in the division 3 emergency diesel start circuitry. Specifically, the electricians opened test switches in the wrong electrical cubicle resulting in the unexpected start of the division 3 emergency diesel generator and a loss of power to the 1C1 bus from an offsite source. The licensee entered this issue into their corrective action program (CAP) as Action Request (AR) 04012393. As corrective actions, the licensee performed a human performance review to identify the reasons the procedure was not followed and restored power to the 1C1 safety bus.
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| The performance deficiency was determined to be more than minor because it impacted the Initiating Events cornerstone attribute of human performance and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations.
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| Specifically, the failure of the electrical maintenance technicians to follow their procedures resulted in a loss of power to the 1C1 electrical bus. The finding was screened against the Initiating Events cornerstone and determined to be of very low safety significance because the loss of power to the 1C1 bus occurred while Clinton was in a refueling outage when the high pressure core spray system was removed from service and not being relied upon for shutdown safety defense in depth. The loss of the 1C1 bus did not affect decay heat removal from the core, did not affect reactor coolant inventory, and the event occurred while the refuel cavity was flooded up for refueling operations. The inspectors determined that this finding affected the cross-cutting area of human performance in the aspect of avoid complacency where individuals implement appropriate error reduction tools. Specifically, as documented in the licensee's human performance review, the electricians performing the work did not utilize any human performance tools to flag the https://www.nrc.gov/reactors/operating/oversight/clin/clin-pim.html 10/19/2017
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| NRC: Clinton - Quarterly Plant Inspection Findings Page 2 of 16 equipment to be operated and improperly performed the concurrent verification of the component to be manipulated.
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| [H.12]
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| Inspection Report# : 2017002 (pdf)
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| Significance: Jun 30, 2017 Identified By: NRC Item Type: NCV Non-Cited Violation ROOT CAUSE EVALUATION FAILED TO IDENTIFY CORRECTIVE ACTION TO PRECLUDE REPETITION The inspectors identified a finding of very low safety significance and an associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion II, "Quality Assurance Program," for the failure to implement a quality assurance program procedure. Specifically, the licensee failed to document a root cause and develop a corrective action to preclude repetition for the 1A bus transformer failure in accordance with quality assurance procedure PI-AA-125-1001, "Root Cause Analysis Manual." The licensee entered this issue into their CAP as AR 01594407. The corrective actions in response to this issue were to revise the root cause report with a root cause of insulation degradation of the phase windings over time and develop a corrective action to prevent recurrence by using Doble testing to ensure indication of transformer insulation degradation was discovered prior to failure.
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| The performance deficiency was determined to be more than minor because if left uncorrected the performance deficiency had the potential to lead to a more significant safety concern. Specifically, the root cause and corrective actions to prevent recurrence were not identified until the licensee was prompted by the inspectors. As a result, additional transformer failures could have occurred. The finding was screened against the Initiating Events cornerstone and determined to be of very low safety significance because the finding did not involve the complete or partial loss of a support system that contributes to the likelihood of or cause an initiating event nor did it affect mitigation equipment.
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| The inspectors determined this finding affected the cross-cutting area of human performance, in the aspect of resources, where leaders ensure that personnel, equipment, procedures and other resources are available and adequate to support nuclear safety. Specifically, the licensee's station procedure did not provide guidance on when a corrective action to preclude repetition is required, regardless of whether a risk assessment was performed. [H.1]
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| Inspection Report# : 2017002 (pdf)
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| Significance: Mar 31, 2017 Identified By: Self-Revealing Item Type: NCV Non-Cited Violation FAILURE TO DEVELOP AND REVIEW A WORKER TAG OUT The inspectors documented a self-revealed finding of very low safety significance and associated non-cited violation of Technical Specification 5.4.1, "Procedures," for the licensee's failure to develop and review a worker tag out in accordance with station procedure OP-AA-109-10, "Clearance and Tagging," Revision 12. Specifically, the licensee failed to identify the effect of a worker tag out on the in-service steam jet air ejector suction valve, which caused condenser vacuum to degrade resulting in the operators entering the off normal procedure for loss of condenser vacuum. The licensee entered this issue into their corrective action program as action request (AR) 03980495. As corrective actions, the operations department issued a standing order to require worker tag outs to be challenged by a second senior reactor operator.
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| The performance deficiency was determined to be more than minor because it impacted the Initiating Events cornerstone attribute of configuration control and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during power operations. Specifically, the failure to properly develop the worker tag out caused the condenser vacuum to degrade, challenging the operators to https://www.nrc.gov/reactors/operating/oversight/clin/clin-pim.html 10/19/2017
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| NRC: Clinton - Quarterly Plant Inspection Findings Page 3 of 16 quickly diagnose the issue and take action to avoid a turbine trip. The finding was screened against the Initiating Events cornerstone and determined to be of very low safety significance because it did not cause a reactor trip or the loss of mitigation equipment relied upon to transition the plant from the onset of a trip to a stable shutdown condition. The inspectors determined that this finding affected the cross-cutting area of human performance in the aspect of avoid complacency, where individuals recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Individuals implement appropriate error reductions tools. Specifically, the operations department failed to implement appropriate error reduction tools such as questioning attitude and thorough work product reviews to ensure the worker tag out considered all potential effects to other plant equipment. [H.12]
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| Inspection Report# : 2017001 (pdf)
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| Significance: Sep 30, 2016 Identified By: Self-Revealing Item Type: NCV Non-Cited Violation Failure to Prevent Recurrence of a Significant Condition Adverse to Quality The inspectors documented a self-revealing finding of very low safety significance and an NCV of 10 CFR 50, Appendix B, Criterion XVI for the licensee's failure to take corrective action to preclude repetition of a significant condition adverse to quality (SCAQ). After identifying intergranular stress corrosion cracking (IGSCC) on main steam flex hoses in 2007 and concluding the leakage constituted an SCAQ, the licensee's corrective actions to prevent recurrence failed to prevent pressure boundary leakage at the same location in 2016. The licensee entered this issue into their corrective action program as AR 02670593. The affected hoses were replaced. The licensee is also developing a design change to address at least one of the three factors that contributes to IGSCC.
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| The inspectors determined that the licensee's failure to take corrective actions to prevent recurrence of an SCAQ was a performance deficiency and more than minor because if left uncorrected pressure boundary leakage could become a more significant concern. Specifically, pressure boundary leakage is not allowed by TS and any leakage requires the plant to be shutdown. The finding was screened as low safety significant because it did not result in exceeding the RCS leak rate for a small LOCA and did not affect systems used to mitigate a LOCA. No cross cutting aspect was assigned as the original issue occurred greater than three years ago and was not reflective of current performance.
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| Inspection Report# : 2016003 (pdf)
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| Mitigating Systems Significance: Jun 30, 2017 Identified By: NRC Item Type: NCV Non-Cited Violation FAILURE OF OPERATORS TO MEET TIME CRITICAL OPERATOR ACTIONS The inspectors identified a finding of very low safety significance and an associated non-cited violation of Title 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for the failure to assure that applicable regulatory requirements and the design basis was correctly translated into specifications, drawings, procedures, and instructions and that design control measures provided for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Specifically, the licensee failed to assure/validate operators were able to complete the standby liquid control time critical action for an anticipated transient without a scram specified in their licensing documents.
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| The licensee entered this issue into their CAP as AR 03980202. As corrective actions, the licensee determined the scram choreography required to complete the time critical action in the specified time, initiated a standing order to inform the operating crews, processed a procedure change for the anticipated transient without scram choreography and performed an evaluation to determine the impact of initiating the standby liquid control system at 172 seconds.
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| https://www.nrc.gov/reactors/operating/oversight/clin/clin-pim.html 10/19/2017
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| NRC: Clinton - Quarterly Plant Inspection Findings Page 4 of 16 The performance deficiency was determined to be more than minor because the finding was associated with the procedure quality attribute of the Mitigating Systems cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, with the operators initiating standby liquid control at 172 seconds instead of 120 seconds, the accident analysis calculations were required to be re-performed to assure the accident analysis requirements were met. The finding was screened against the Mitigating Systems cornerstone and determined to be of very low safety significance because the inspectors were able to answer all of the associated screening questions "No." The inspectors determined that this finding is not indicative of current performance and therefore did not assign a cross-cutting aspect.
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| Inspection Report# : 2017002 (pdf)
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| Significance: Jun 30, 2017 Identified By: NRC Item Type: FIN Finding FAILURE TO PERFORM PREVENTIVE MAINTENANCE ON A SAFETY-RELATED BREAKER CUBICLE The inspectors identified a finding of very low safety significance for the licensee's failure to perform maintenance on a safety-related motor control center cubicle. Specifically, the licensee failed to perform thermography on the division 1 shutdown service water pump room cooler breaker cubicle in accordance with the maintenance strategy/template without providing justification for differing from the template as required by MA-AA-716-210, "Performance Centered Maintenance Process," Revision 3. This resulted in the division 1 shutdown service water pump room cooler fan failing because of a high resistance connection that went undetected. The licensee entered this issue into their CAP as AR 02667822. As corrective actions, the licensee replaced the thermal overload relays and created a preventative maintenance action to perform thermography on this equipment on a periodic basis.
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| This performance deficiency was determined to be more than minor because it impacted the Mitigating Systems cornerstone attribute of equipment performance and adversely affected the cornerstone objective of ensuring the availability, capability and reliability of equipment that responds to initiating events. Specifically, the room cooler fan failure directly impacted the operability of the division 1 shutdown service water pump and the division 1 emergency diesel generator which are safety-related, risk significant systems. The finding was screened against the Mitigating Systems cornerstone and determined to be of very low safety significance because the inspectors were able to answer all of the associated screening questions "No." The inspectors determined that this finding is not indicative of current plant performance and therefore did not assign a cross-cutting aspect.
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| Inspection Report# : 2017002 (pdf)
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| Significance: Apr 21, 2017 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Perform Required Surveillances on Multiple Fire Dampers (Section 1R05.2b)
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| Green. The inspectors identified a finding of very-low safety significance (Green), and an associated Non-Cited Violation of License Condition 2.C(f) for the licensee's failure to adequately implement surveillance procedures and work processes associated with fire barrier damper inspections. Specifically, the licensee failed to perform fire barrier damper inspections for 15 fire dampers once every 48 months (plus an additional 25 percent grace period) as required by the Fire Protection Program. The licensee entered the issue into their Corrective Action Program, and will inspect the fire barrier dampers during the next refueling outage.
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| The inspectors determined that the performance deficiency was more-than-minor because the licensee's failure to inspect the fire barrier dampers could result in not identifying degraded dampers which could affect their ability to prevent a fire from spreading from one fire area to another. The finding was of very-low safety significance because the https://www.nrc.gov/reactors/operating/oversight/clin/clin-pim.html 10/19/2017
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| NRC: Clinton - Quarterly Plant Inspection Findings Page 5 of 16 failure to inspect the fire barrier dampers did not impact the plant's ability to reach and maintain safe-shutdown. The finding has a cross-cutting aspect in the area of Human Performance, Work Management because the licensee failed to execute a work order to inspect the fire dampers in accordance with the required frequency in Procedure CPS 9601.01 and instead improperly extended the frequency of the fire damper inspections. (Section 1R05.2b) [H.5]
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| Inspection Report# : 2017008 (pdf)
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| Significance: Mar 31, 2017 Identified By: NRC Item Type: NCV Non-Cited Violation PLANT BARRIER CONTROL PROGRAM FAILED TO COMPENSATE FOR AN IMPACTED FLOOD BARRIER The inspectors identified a finding of very low safety significance and an associated non-cited violation of Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings,"
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| for the failure to implement the plant barrier control program for an impacted flood barrier. Specifically, the plant barrier impairment (PBI) permit, PBI-2017-02-003, for work on watertight door 1SD1-24, failed to identify the door as a flood barrier and that appropriate compensatory measures for 1SD1-24 being open for an extended period were identified or implemented in accordance with station procedure CC-AA-201, "Plant Barrier Control Program,"
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| Revision 11. The licensee entered this issue into their corrective action program as AR 03980495. The corrective actions in response to this violation were to identify appropriate compensatory measures for impairment of 1SD1-24 and incorporate them into the PBI log.
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| The performance deficiency was determined to be more than minor because it impacted the Mitigating Systems cornerstone attribute of protection against external events and adversely affected the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. With the flood barrier nonfunctional and without compensatory actions in place the residual heat removal (RHR) 'B' and RHR 'C' pumps were inoperable. The finding was screened against the Mitigating Systems cornerstone and the inspectors determined that the finding involved the loss or degradation of equipment or function specifically designated to mitigate a seismic, flooding or severe weather initiating event. The inspectors determined that the loss of this equipment or function by itself during the external initiating event would degrade one or more trains of a system that supports a risk significant system or function and would require a detailed risk evaluation. The senior reactor analyst (SRA) performed the detailed risk evaluation and concluded the finding was of very low safety significance. The inspectors determined that this finding affected the cross-cutting area of human performance in the aspect of conservative bias, where individuals use decision making practices that emphasize prudent choices over those that are simply allowable. Proposed actions are determined to be safe in order to proceed, rather than unsafe in order to stop. Specifically, during preparation of the PBI permit, the station PBI log was reviewed and actions for previous work associated with the watertight door were deemed acceptable even though the work on the door in those instances was different than the work being performed this time. [H.14]
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| Inspection Report# : 2017001 (pdf)
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| Significance: Mar 31, 2017 Identified By: NRC Item Type: NCV Non-Cited Violation FAILED TO VERIFY AN APPROPRIATE ALTERNATE METHOD OF DECAY HEAT REMOVAL The inspectors identified a finding of very low safety significance and an associated non-cited violation of 10 CFR 50.36(c)(2)(i), "Limiting conditions for operation", for failing to meet/follow the required actions for limiting condition for operation 3.9.9 and 3.4.10. Specifically, the operators failed to verify a credited alternate decay heat removal method that would satisfy the required action for the limiting condition for operation. The licensee entered this issue https://www.nrc.gov/reactors/operating/oversight/clin/clin-pim.html 10/19/2017
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| NRC: Clinton - Quarterly Plant Inspection Findings Page 6 of 16 into their corrective action program as AR 03987440. The corrective actions in response to this violation were to identify appropriate alternate methods of decay heat removal and incorporate them into the shutdown safety management program utilized during plant outages.
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| The performance deficiency was determined to be more than minor because it impacted the Mitigating Systems cornerstone attribute of equipment performance and adversely affected the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences.
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| Specifically, with the operators failing to identify a credited alternate method of decay heat removal and taking credit for the inoperable but in service RHR shutdown cooling train, the actual available methods that could have been credited were not verified to ensure their availability to provide the required function. The finding was screened against the Mitigating Systems Screening questions and determined to be of very low safety significance because the answer to all of the applicable screening questions was "No." The inspectors determined that this finding affected the cross-cutting area of human performance in the aspect of conservative bias, where individuals use decision making practices that emphasize prudent choices over those that are simply allowable. Proposed actions are determined to be safe in order to proceed, rather than unsafe in order to stop. Specifically, the senior reactor operators at the station had historically credited inoperable RHR shutdown cooling subsystems as their own alternate decay heat remove method because they believed it was allowable without determining that it was safe in order to proceed. [H.14]
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| Inspection Report# : 2017001 (pdf)
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| Significance: Mar 31, 2017 Identified By: Self-Revealing Item Type: NCV Non-Cited Violation FAILURE TO PERFORM MAINTENANCE ON RESIDUAL HEAT REMOVAL PUMP 'C' BREAKER IN ACCORDANCE WITH PROCEDURES The inspectors documented a self-revealed finding of very low safety significance and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures and Drawings," for the licensee's failure to perform maintenance on a safety related breaker in accordance with station procedure Clinton Power Station (CPS) 8410.12C001, "Westinghouse DHP Circuit Breaker Checklist," Revision 7. Specifically, the licensee failed to ensure the remaining travel on the latch check switch for the RHR 'C' pump breaker was within the acceptable range resulting in the RHR 'C' pump failing to start. The licensee entered this issue into their corrective action program as AR 03949655. The corrective actions taken by the licensee included providing coaching to the involved individuals as well as changing the procedure to include a block to record the latch check switch over travel.
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| The performance deficiency was determined to be more than minor because it impacted the Mitigating Systems cornerstone attribute of equipment performance and adversely affected the cornerstone objective of ensuring the availability, capability, and reliability of equipment that responds to initiating events. Specifically, the performance deficiency adversely impacted the operability of the RHR 'C' pump. The inspectors reviewed the Mitigating Systems screening questions and determined a detailed risk evaluation was required because question A.3 was answered yes.
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| The SRA performed the detailed risk evaluation and concluded the finding was of very low safety significance. The inspectors determined that this finding affected the cross-cutting area of human performance in the aspect of resources, where leaders ensure that personnel, equipment, procedures, and other resources are available and adequate to support nuclear safety. Specifically, the organization failed to ensure the procedure step included a block for recording the latch check switch over travel value, which led to confusion on whether the value was required to be recorded and ultimately resulted in a failure to perform the step as written. [H.1]
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| Inspection Report# : 2017001 (pdf) https://www.nrc.gov/reactors/operating/oversight/clin/clin-pim.html 10/19/2017
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| NRC: Clinton - Quarterly Plant Inspection Findings Page 7 of 16 Significance: Dec 31, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Demonstrate the Condition of Flood Seals was being Effectively Controlled The inspectors identified a finding of very low safety significance and an associated non-cited violation of 10 CFR 50.65(a)(2), "Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," for the licensee's failure to demonstrate that the condition of flood seals was being effectively controlled through appropriate preventive maintenance such that the flood seals remained capable of performing their intended function. Specifically, the licensee failed to visually inspect flood seals per ER-AA-450, "Structures Monitoring" and ER-CL-450-1007, "Clinton Surveillance Inspection Program for Seals." As corrective actions, the licensee planned to visually inspect all accessible flood seals and generate an evaluation for inaccessible seals. In addition, the licensee planned to modify ER-AA-450 to clarify the frequency of flood seal inspection.
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| The inspectors determined the licensee's failure to demonstrate that the condition of flood seals was being effectively controlled by visual inspection, per ER-AA-450 and ER-CL-450-1007, was a performance deficiency. The performance deficiency was determined to be more than minor in accordance with IMC 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," dated September 7, 2012, because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, failure to monitor the condition of the flood sales in a manner sufficient to provide reasonable assurance they were capable of fulfilling the intended safety functions could adversely affect multiple mitigating systems in the event of a flood or line break. Using IMC 0609, Attachment 4, "Initial Characterization of Findings," issued October 7, 2016, and Appendix A, "The Significance Determination Process for Findings at Power," issued June 19, 2012, the finding was screened against the Mitigating Systems cornerstone and determined to be of very low safety significance because the inspectors answered no to the question "does the finding involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding, or severe weather initiating event?" The inspectors determined this finding affected the cross-cutting area of Problem Identification and Resolution, in the aspect of identification, where the organization implements a corrective action program with a low threshold for identifying issues. Individuals identify issues completely, accurately, and in a timely manner in accordance with the program. Specifically, the licensee failed to identify the flood seals still had not been inspected when they performed the Maintenance Rule Program - Structures Monitoring Assessment which credited ER-CL-450-1007 in early 2014.
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| Inspection Report# : 2016004 (pdf)
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| Significance: Dec 01, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Promptly Identify that the Incapability of the RHR Design to Support TS Operability Requirements Was a CAQ (Section 4OA2.b(1))
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| Green: The team identified a finding of very-low safety significance (Green) and an associated NCV of Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion XVI, "Corrective Action," for the licensee failure to promptly identify that the incapability of the residual heat removal (RHR) design to support Technical Specifications (TS) operability requirements was a condition adverse to quality. Specifically, when reactor water temperature was greater than 150 degrees Fahrenheit, RHR could not be realigned from shutdown cooling mode of operations to provide the TS required functions of the emergency core cooling system, suppression pool cooling, containment spray, and feedwater leakage control system. The licensee captured this issue in their Corrective Action Program (CAP) as Action Request (AR) 02742439 and AR 03948042, and planned to submit a License Amendment Request to align TS https://www.nrc.gov/reactors/operating/oversight/clin/clin-pim.html 10/19/2017
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| NRC: Clinton - Quarterly Plant Inspection Findings Page 8 of 16 requirements with the design capabilities.
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| The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of design control and adversely affected the associated cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the performance deficiency resulted in voluntarily declaring TS functions inoperable when performing shutdown cooling operations, which did not ensure the associated mitigating systems availability or capability to respond to an initiating event. The team determined that this finding was of very low safety significance (Green). Specifically, there were no known instances where the finding: (1) represented a loss of system safety function; (2) represented an actual loss of safety function of at least a single train or two separate safety systems out of service for greater than their TS allowed outage time; (3) involved non-TS trains of equipment; (4) involved a degradation of a functional RHR auto-isolation on low reactor vessel level; (5) impacted external event protection; or (6) involved fire brigade issues. The team did not identify a cross cutting aspect associated with this finding because it did not reflect current licensee performance since the performance deficiency occurred more than 3 years ago. (Section 4OA2.b(1))
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| Inspection Report# : 2016009 (pdf)
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| Significance: Sep 30, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Exceeded Technical Specification Allowed Outage Time for Main Turbine Bypass System The inspectors identified a finding of very low safety significance and an associated NCV of Technical Specification 3.7.6, "Main Turbine Bypass System" for the licensee's failure to meet the limiting conditions for operation and complete the associated required actions after making a deficient change to the turbine bypass valve surveillance testing frequency. Specifically, with the main turbine bypass system inoperable and without the Core Operating Limits Report (COLR) limits for thermal power, minimum critical power ratio (MCPR), and linear heat generation rate (LGHR) with the main turbine by pass system inoperable applied, thermal power was not reduced to less than 21.6 percent of rated thermal power within six hours. The licensee entered this issue into their corrective action program as AR 02690657.
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| The licensee restored compliance by applying the COLR limits for reactor thermal power, MCPR and LGHR.
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| The inspectors determined the failure to meet the limiting conditions for operation and complete the associated required actions prior to the end of the specified completion times was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was screened against the Mitigating Systems cornerstone and determined to be of very low safety significance because all of the associated questions in IMC 0609, Appendix A, were answered no. The inspectors determined this finding affected the cross-cutting area of human performance, in the aspect of change management, where leaders use a systematic process for evaluating and implementing change so that nuclear safety remains the overriding priority because the licensee's change management process was not fully utilized by senior management when evaluating and implementing a change to the turbine bypass valve surveillance testing frequency. (H.3)
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| Inspection Report# : 2016003 (pdf)
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| Significance: Sep 30, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Perform a 50.59 Screening for Changing the Frequency of Exercising the Turbine Bypass Valves https://www.nrc.gov/reactors/operating/oversight/clin/clin-pim.html 10/19/2017
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| NRC: Clinton - Quarterly Plant Inspection Findings Page 9 of 16 The inspectors identified a Severity Level IV NCV of 10 CFR 50.59 4(d)(1), "Changes, Tests, and Experiments," and an associated Green finding for the licensee's failure to perform a written evaluation which provided the bases for determining that changing the turbine bypass valve surveillance testing frequency from every 31 days, as specified in the Updated Safety Analysis Report, to once a year did not require a license amendment. The licensee has entered this issue into their corrective action program as AR 02720163. The licensee is currently evaluating the issue in accordance with their procedure for changes to the facility.
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| The inspectors determined that the licensee's failure to perform a written evaluation to provide the basis for the determination that a change to the facility, a change to a procedure, or a change to a test or experiment did not require a license amendment was a performance deficiency. The performance deficiency was more than minor in accordance with IMC 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," dated September 7, 2012, because, it was associated with the procedure quality attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using IMC 0609, Attachment 4, "Initial Characterization of Findings," and Appendix A, "The Significance Determination Process for Findings At-Power," issued June 19, 2012, the finding was screened against the Mitigating Systems cornerstone and determined to be of very low safety significance because all of the associated questions in IMC 0609, Appendix A, were answered no. Violations of 10 CFR 50.59 are dispositioned using the traditional enforcement process because they are considered to be violations that potentially impede or impact the regulatory process. The inspectors reviewed Section 6.1.d.2 of the NRC Enforcement Policy and determined this violation was Severity Level IV because the resulting changes were evaluated by the SDP as having very low safety significance. The inspectors determined this finding affected the cross-cutting area of human performance, in the aspect of consistent process, where individuals use a consistent, systematic approach to make decisions. The licensee made a decision to proceed with implementation of a change to the turbine bypass valve surveillance testing frequency after a plant oversight committee review in lieu of following their consistent, systematic process for evaluating changes to the USAR. (H.13)
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| Inspection Report# : 2016003 (pdf)
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| Significance: Aug 09, 2016 Identified By: NRC Item Type: FIN Finding Failure to have hose configurations that were verified to be able to ensure a timely and successful implementation of a FLEX strategy Green. Two examples of a finding of very low safety significance was identified by the inspectors for the licensee's failure to have hose configurations that were verified to be able to ensure a timely and successful implementation of a flexible response (FLEX) strategy. Specifically, the licensee did not ensure through evaluations, calculations, analyses or any other means that the strategy for maintaining core cooling, containment heat removal and Spent Fuel Pool (SFP) cooling during a Beyond-Design-Basis External Event (BDBEE) flooding scenario would be capable of fulfilling its function. No violation of NRC requirements were identified.
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| The performance deficiency is more than minor because it was associated with the mitigating systems cornerstone objective attribute of protection against external factors, specifically the BDBEE flood hazard, and it adversely affected the cornerstone attribute of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Issues identified through TI-191 are evaluated through a cross-regional panel using IMC 0609, Appendix M, "Significance Determination Process Using Qualitative Criteria," as informed by draft Appendix O, "Post Fukushima Mitigation Strategies Significance Determination Process." The finding was determined to be of very low safety significance (Green). The inspectors concluded that the cause of the finding involved a cross-cutting component in the Human Performance area of Design Margins because the organization did https://www.nrc.gov/reactors/operating/oversight/clin/clin-pim.html 10/19/2017
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| NRC: Clinton - Quarterly Plant Inspection Findings Page 10 of 16 not ensure the selected strategy contained the required verification that it could be successfully implemented. [H.6]
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| Inspection Report# : 2016007 (pdf)
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| Barrier Integrity Significance: Jun 30, 2017 Identified By: NRC Item Type: VIO Violation FAILURE TO PERFORM ADEQUATE EVALUATION OF CRANE RAIL CLIPS The inspectors identified a finding of very-low safety significance and an associated cited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for the failure to properly verify the adequacy of design of the fuel building crane and crane support structure elements. Specifically, calculations involving the crane rail clips and clip bolts had multiple technical errors and failed to adequately demonstrate that the design met the design basis requirements. The licensee initiated corrective actions by documenting the deficiency in AR 4001089 and performed an evaluation demonstrating that the functionality of the crane was maintained.
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| The finding was determined to be more-than-minor because it was associated with the design control attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective of maintaining the functionality of the spent fuel pool (SFP) cooling system. Specifically, crane rail clip bolts were required to ensure structural integrity of structures, systems, and components described in the Updated Safety Analysis Report, when subjected to design loads as part of safe load handling of heavy loads near the SFP and to ensure integrity of the spent fuel cask. In accordance with IMC 0609, "Significance Determination Process," Attachment 0609.04, "Initial Characterization of Findings,"
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| Table 2, the inspectors determined the finding affected the Barrier Integrity cornerstone because it was associated with SFP/fuel handling activities. Based on answering "No" to questions A through F in Table 3, the inspectors determined the finding could be evaluated using Appendix A, "The Significance Determination Process for Findings At-Power,"
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| Exhibit 3, for the Barrier Integrity cornerstone screening questions. Based on the crane remaining functional, the inspectors answered "No" to Questions D.1 through D.4 because the finding did not adversely affect decay heat removal capabilities, did not result from fuel handling errors, did not result in loss of SFP inventory, and did not affect the SFP neutron absorber or fuel bundle misplacement; therefore, the finding screened as having very-low safety significance. The finding was cross-cutting in the resolution aspect of the problem identification and resolution area because the licensee failed to take effective corrective actions in a timely manner to address issues identified earlier in the rail clip evaluations. [P.3]
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| Inspection Report# : 2017002 (pdf)
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| Significance: Jun 30, 2017 Identified By: Self-Revealing Item Type: NCV Non-Cited Violation FAILURE TO PROVIDE SUFFICIENT WORK INSTRUCTIONS FOR PERFORMING MAINTENANCE ON THE CONTROL ROOM VENTILATION SYSTEM CHARCOAL FILTER The inspectors documented a self-revealed finding of very low safety significance and an associated non-cited violation of 10 of CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," for the failure of the licensee to provide sufficient work instructions for performing maintenance on the control room ventilation charcoal filter bed.
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| Specifically, the work order used to change out the charcoal filter bed (Work Order 01494189) contained only the minimum required amount of charcoal to place in the bed. Sometime after filling the bed April 6, 2015, the charcoal settled, resulting in the 'B' control room ventilation system being declared inoperable after failing a surveillance test.
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| The licensee entered this issue into their CAP as AR 03995612. As corrective actions, the licensee is revising the WO instructions and Clinton Power Station Procedure 9866.03 to require that charcoal be filled completely to the bottom of https://www.nrc.gov/reactors/operating/oversight/clin/clin-pim.html 10/19/2017
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| NRC: Clinton - Quarterly Plant Inspection Findings Page 11 of 16 the deluge piping to allow for settling.
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| The performance deficiency was determined to be more than minor because it impacted the Barrier Integrity cornerstone attribute of procedure quality and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events.
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| Specifically, the failure to provide sufficient guidance in the work order regarding the quantity of charcoal to be installed resulted in the 'B' control room ventilation system failing a surveillance test and being declared inoperable.
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| The finding was screened against the Barrier Integrity cornerstone and determined to be of very low safety significance because the finding only represents a degradation of a radiological barrier function provided for the control room. The inspectors determined that this finding affected the cross-cutting area of human performance in the aspect of design margins, where the organization operates and maintains equipment within design margins. Special attention is placed on maintaining fission product barriers, defense in depth, and safety-related equipment. Specifically, when performing maintenance on the charcoal bed, the licensee failed to recognize that filling the charcoal to the minimum bed level provided no margin if settling occurred. [H.6]
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| Inspection Report# : 2017002 (pdf)
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| Significance: Dec 01, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Non Conservative Control Room Radiological Habitability Assessment (Section 1R21.3.b(1))
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| Green. The team identified a finding of very-low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for the licensee failure to use a technically appropriate analytical methodology in the control room radiological habitability calculation. Specifically, the licensee used a methodology that inappropriately characterized the control room heating, ventilation and air conditioning (HVAC) system outside air intake design resulting in a calculated control room dose following a loss of coolant accident that exceeded the applicable limit. The licensee captured this issue in their CAP as AR 02742442, completed an operability evaluation, and issued an NRC event notification.
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| The performance deficiency was determined to be more-than-minor because it was associated with the Barrier Integrity cornerstone attribute of design control and affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the performance deficiency resulted in the control room expected dose following a loss of coolant accident to exceed the applicable limits prompting an operability evaluation. The finding screened as of very-low safety significance (Green) because it only represented a degradation of the radiological barrier function provided for the control room.
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| Specifically, the finding did not affect the control room barrier function against smoke or a toxic atmosphere. The team did not identify a cross-cutting aspect associated with this finding because it was not confirmed to reflect current performance due to the age of the performance deficiency. Specifically, the affected calculations were performed more than 3 years ago. (Section 1R21.3.b(1))
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| Inspection Report# : 2016009 (pdf)
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| Significance: Dec 01, 2016 Identified By: NRC Item Type: NCV Non-Cited Violation Failure to Scope SFP Temperature and Level Instruments into the Maintenance Rule Program (Section 1R21.3.b(2))
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| Green. The team identified a finding of very-low safety significance (Green) and an associated NCV of Paragraph (b)
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| (2)(i) of 10 CFR 50.65, "Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," for the licensee failure to scope non-safety related mitigating structure, systems, and components (SSCs) used within an https://www.nrc.gov/reactors/operating/oversight/clin/clin-pim.html 10/19/2017
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| NRC: Clinton - Quarterly Plant Inspection Findings Page 12 of 16 emergency operating procedure (EOP) into Maintenance Rule Program. Specifically, an EOP used spent fuel pool (SFP) low-level and high-temperature parameters as distinct entry criteria but the associated components were not included in the scope of the Maintenance Rule Program. The licensee captured the team concerns in their CAP as AR 02736193, performed an extent of condition to identify any other SSC addition to the EOPs requiring them to be added to the Maintenance Rule Program scope, and initiated plans to incorporate the affected SSCs into the Maintenance Rule Program scope.
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| The performance deficiency was determined to be more-than-minor because it was associated with the Barrier Integrity cornerstone attribute of SSC performance and affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, a key aspect of the Maintenance Rule is to ensure that maintenance activities are performed in a manner that provide reasonable assurance that SSCs within its scope perform reliably and are capable of providing their intended Maintenance Rule function(s). In the case of the SFP temperature instruments, the licensee was not performing preventive maintenance to ensure that degradation, such as instrument drift, did not adversely affect their ability to detect and alarm EOP entry conditions such that mitigating actions could be implemented to preserve secondary containment. The finding screened as of very-low safety significance (Green) because it only represented a degradation of the radiological barrier function provided for the control room. Specifically, the finding did not cause SFP temperature to exceed the maximum analyzed limit, a detectible release of radionuclides, water inventory to decrease below the analyzed limit, or an adverse effect to the SFP neutron absorber or fuel loading pattern. The team determined that the finding had a cross cutting aspect in the area of human performance because the licensee did not use a systematic process for evaluating and implementing changes when updating the affected EOP in 201}}
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