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05000456/FIN-2011012-032011Q2BraidwoodIncorrect Installation of Annunciator System WiringA finding of very low safety significance was self-revealed at Braidwood Station when licensee personnel failed to properly install portions of the annunciator system circuitry in accordance with design specifications. Specifically, wiring in the annunciator system clock circuitry (the portion of the circuitry that allows annunciators to change status) was incorrectly installed, which resulted in an unexpected loss of all Braidwood Unit 2 control room annunciators on March 24, 2011. The licensee entered the issue into the corrective action program (CAP) as IR 1192465, corrected the wiring to provide the intended function, and revised procedures used to energize and de-energize the system. The finding was determined to be more than minor because it was associated with the Equipment Performance 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, annunciator system redundancy was adversely affected and when the annunciator panels were de-energized, the ability of operators to identify and respond to abnormal plant conditions was degraded. Because the finding was not a design deficiency, did not result in a loss of safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event, the inspectors concluded that the finding was of very low safety significance (Green). The inspectors did not identify a cross-cutting aspect associated with this finding because it was not indicative of current performance.
05000456/FIN-2011012-042011Q2BraidwoodUntimely Declaration of a Notice of Unusual EventThe inspectors identified a finding of very low safety significance (Green) and an associated non-cited violation of 10 CFR 50.54(q) at Braidwood Station after licensee personnel failed to promptly declare a Notice of Unusual Event in accordance with the Braidwood Emergency Plan. Specifically, on March 24, 2011, contrary to the Braidwood Station Radiological Emergency Plan Annex, the licensee did not declare Emergency Action Level (EAL) MU6 (Unusual Event) within 15 minutes of indications of a loss of greater than 75 percent of Unit 2 main control room annunciators. Corrective actions included implementation of Standing Order 11-007; additional training; and procedures revisions, which were all intended to clarify the function of the annunciator test push buttons in determining whether a loss of annunciators has occurred. The finding was more than minor because it was associated with the Emergency Response Organization Performance attribute of the Emergency Preparedness cornerstone, and affected the cornerstone objective of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Using the emergency preparedness significance determination process, Sheet 2, Actual Event Implementation Problem, the inspectors determined the finding was of very low safety significance (Green) because the licensee failed to implement a risk significant planning standard (10 CFR 50.47(b)(4)) during an actual Notice of Unusual Event. This finding was associated with a cross-cutting aspect in the Resources component of the Human Performance cross-cutting area because the licensee did not ensure that procedures were accurate and adequate to assure nuclear safety. Specifically, when provided with sufficient evidence that the annunciators were not properly responding, licensee personnel delayed implementation of the Emergency Plan until further information was obtained. This was due to inaccurate and conflicting procedures and a lack of knowledge of the annunciator system.
05000454/FIN-2011015-022011Q2ByronFailure to Adequately Document and Justify Continued Operability of the Auxiliary Feedwater SystemA finding of very low safety significance was identified at the Braidwood and Byron Stations by the inspectors when licensee personnel failed to adequately document and justify continued operability of the auxiliary feedwater (AF) system. Specifically, licensee evaluations of known voids in the AF alternate source suction piping did not provide an adequate technical basis to support operability of the AF pumps during a suction swap-over scenario. Subsequently, the licensee filled the voids and a Root Cause Evaluation (RCE) was initiated under Issue Report (IR) 1194196 (Braidwood) and IR 1194324 (Byron). The RCE was initiated to determine why prior opportunities for discovery of the inadequate void acceptance basis were missed and to develop associated corrective actions. The inspectors determined the finding was more than minor because, if left uncorrected, the failure to recognize conditions that could render equipment inoperable had the potential to lead to a more significant safety concern. Because the finding was not a design deficiency, did not result in a loss of safety function, and did not screen as potentially risk-significant due to a seismic, flooding, or severe weather initiating event, the inspectors concluded that the finding was of very low safety significance (Green). This finding was associated with a cross-cutting aspect in the Decision-Making component of the Human Performance cross-cutting area because the licensee did not use conservative assumptions and did not verify the validity of underlying assumptions in their evaluations of the AF suction piping voids.
05000456/FIN-2011012-022011Q2BraidwoodFailure to Adequately Document and Justify Continued Operability of the Auxiliary Feedwater SystemA finding of very low safety significance was identified at the Braidwood and Byron Stations by the inspectors when licensee personnel failed to adequately document and justify continued operability of the auxiliary feedwater (AF) system. Specifically, licensee evaluations of known voids in the AF alternate source suction piping did not provide an adequate technical basis to support operability of the AF pumps during a suction swap-over scenario. Subsequently, the licensee filled the voids and a Root Cause Evaluation (RCE) was initiated under Issue Report (IR) 1194196 (Braidwood) and IR 1194324 (Byron). The RCE was initiated to determine why prior opportunities for discovery of the inadequate void acceptance basis were missed and to develop associated corrective actions. The inspectors determined the finding was more than minor because, if left uncorrected, the failure to recognize conditions that could render equipment inoperable had the potential to lead to a more significant safety concern. Because the finding was not a design deficiency, did not result in a loss of safety function, and did not screen as potentially risk-significant due to a seismic, flooding, or severe weather initiating event, the inspectors concluded that the finding was of very low safety significance (Green). This finding was associated with a cross-cutting aspect in the Decision-Making component of the Human Performance cross-cutting area because the licensee did not use conservative assumptions and did not verify the validity of underlying assumptions in their evaluations of the AF suction piping voids.
05000454/FIN-2011015-012011Q2ByronDesign of Auxiliary Feedwater System Included Voids in Safety-Related Alternate Suction FlowpathsA URI was identified by the inspectors during the review of voided sections of AF alternate suction piping affecting all AF trains at Braidwood and Byron. Specifically, the inspectors questioned the acceptability of the voids and the potential impact on the AF systems, and an adequate technical justification from the licensee was not readily available. The AF system provides decay heat removal by cooling the steam generators following a reactor shutdown. The normal supply of water to the AF system is from the non-safety related condensate storage tank (CST), which contains chemically treated water. In the event the CST becomes empty or damaged, the alternate water source is the safety related SX system, which provides lake water at Braidwood and river water at Byron. The design of the AF systems at Braidwood and Byron included a section of SX supply to AF piping that was maintained voided between two valves with a partially open drain. The purpose of this configuration was to prevent SX water intrusion into the steam generators through valve leak-by, which would have an adverse chemical effect since SX water is not chemically treated to the same standards as CST water. On January 31, 2011, NRC inspectors at Byron questioned the existence of voided sections of piping in the SX supply to the AF system. At Byron, the NRC questions were entered into the CAP as IR 1172938. In response to the question identified in the subject IR, the Byron licensee concluded that the AF system was operable with the voids present based on a 1993 Byron Engineering letter addressed to the Braidwood and Byron Station Managers. This letter documented that the AF pumps would not be adversely affected by the ingestion of the assumed voids through the pumps following a swap-over from the CST to the SX water supply. As a basis for this conclusion, the letter referenced a 1987 telephone conversation with the pump vendor and a Duke Engineering Services letter, neither of which were attached to the Byron Engineering letter. The inspectors at Byron questioned whether this conclusion was valid since no formal documentation supporting the conclusion was included with the 1993 Byron Engineering letter. Despite an exhaustive effort at Byron, neither a record of the telephone conversation with the pump vendor or a copy of the Duke Engineering Services letter that provided the basis for the conclusion in the Byron Engineering letter could be found. At Byron, in the absence of formal documentation supporting the conclusion that the AF pumps would not be adversely impacted by the ingestion of the voids following a swap over from the CST to the SX water supply, previously installed vent valves in the voided piping sections were used to fill the voids in all four AF trains on Unit 1 and Unit 2 on February 15, which eliminated current operability questions. Event Notification (EN) 446708 for Byron was submitted to the NRC on March 30, 2011, in accordance with 10 CFR 50.72(b)(3)(ii)(B).
05000461/FIN-2011002-062011Q1ClintonApparent Interaction Between Non-Safety Related and Safety-Related Portions of the NSPS Causing Spurious Component ActuationsCPS UFSAR Section 7.2.1.1.4.8 states, in part, that the STS was designed to meet the separation requirements of Regulatory Guide 1.75 and interfaces are by means of high impedance isolation devices to ensure that failures in the STS will not propagate to the safety equipment. The UFSAR specifically noted that any STS failure would not degrade the NSPS function since STS was isolated from NSPS, eliminating failure propagation. From initial licensing in 1987 and prior to August 26, 2010, the licensee failed to ensure that the safety-related portion of the NSPS was isolated from the non-safety related STS as specified in its design basis. As a result, safety-related components (i.e., primary containment isolation valves) repositioned independent of operator action or a valid signal on March 15, 2010, and during August 24 through 26, 2010. The failure to ensure that the design basis was correctly translated into the design specifications for the STS is a Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control. The NRC determined that this violation resulted from matters not reasonably within the licensees control; that is, the failure to meet the requirements could not be readily identified and, therefore, addressed. Therefore, in accordance with the Enforcement Policy, and after consultation with the Director of the Office of Enforcement and the Region III Regional Administrator, the NRC has decided to exercise enforcement discretion in accordance with Section 3.5 of the NRC Enforcement Policy and to refrain from issuing enforcement action for the violation. In accordance with the NRCs Reactor Oversight Process, this condition will not be considered in the assessment process or the NRCs Action Matrix.
05000461/FIN-2011002-052011Q1ClintonLicensee-Identified ViolationTechnical Specification 5.4.1.a for failure to implement procedures required to conduct timely reviews of job progress and implement actions necessary to reduce workers exposure. Specifically, work in progress reviews for jobs greater than 5 rem were not completed as directed by licensee procedure RP-AA-401, Operational ALARA Planning and Controls; and, therefore, did not implement additional actions necessary to reduce workers exposure. The issue was entered in the licensees corrective action program as AR 01056002. The finding is more than minor because it impacted 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 a full evaluation into the cause for additional exposure was not performed nor were exposure mitigation efforts prescribed. Therefore, additional exposure was received by the plant staff. The inspector determined that this finding did not involve: (1) an ALARA finding; (2) an overexposure; (3) a substantial potential for overexposure; or (4) an impaired ability to assess doses. Consequently, the inspector concluded that the SDP assessment for this finding was of very low safety significance.
05000461/FIN-2011002-032011Q1ClintonInadequate Testing Controls to Perform Surveillance Testing of Hydrogen Igniters in the Primary Containment and DrywellThe 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. 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)).
05000461/FIN-2011002-022011Q1ClintonFailure to Meet Surveillance Testing Requirement for Hydrogen Igniters in Accessible Areas of the Primary Containment and DrywellThe 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 Y 1700 degrees Fahrenheit (aF) 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. 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.
05000461/FIN-2011002-012011Q1ClintonFailure to Control Transient Combustible Materials in Accordance with Fire Protection ProgramThe 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. 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)).
05000455/FIN-2011011-012011Q1ByronSelf-Revealing Failure of the 2A Diesel Generator Upper Lube Oil CoolerA preliminary finding of low to moderate safety significance (White) and an apparent violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed when the 2A Diesel Generator (D/G) was required to be shutdown during routine monthly surveillance testing on November 17, 2010, when a flange connection on a spool piece connected to the upper lube oil cooler failed, resulting in a significant oil leak. The cause of the failure was that Work Order 1206254, Clean Tube Side of Lube Oil Coolers, did not contain appropriate acceptance criteria to ensure proper reassembly of the spool piece for the upper lube oil cooler following maintenance on January 17, 2010. Specifically, the work order package did not contain a final torque verification to ensure that the spool piece flange bolts were torqued to required values, which resulted in the leak. The licensee entered this issue into the correction action program as Issue Report (IR) 1141591, properly re-installed the spool piece, and returned the 2A D/G to service on November 21, 2010. The inspectors determined that this finding was more than minor, because it was associated with the Human Performance attribute of the Mitigating Systems cornerstone and impacted the cornerstone objective of ensuring the availability of systems that respond to initiating events to prevent undesirable consequences. The NRC assessed this finding through a Phase 3 Risk Evaluation of the Significance Determination Process and made a preliminary determination that it was an issue of low to moderate safety significance (White). The cause of this finding was related to the Work Practices component of the Human Performance cross-cutting area since licensee personnel proceeded in the face of uncertainty or unexpected circumstances during the upper lube oil cooler maintenance activity.
05000306/FIN-2010009-012010Q3Prairie IslandFailure to Design Diesels to Survive Tornado Borne MissilesThe inspectors identified an NCV of 10 CFR 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to design the D1/D2 diesel generators to survive impact from the design basis missiles. 10 CFR 50, Appendix B, Criterion III states, in part, that Measures shall be established to assure that applicable regulatory requirements and the design basis...for those systems, structures, and components to which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions. Contrary to this requirement, on July 28, 1994, the licensee approved a calculation that used evaluation methodologies that were not included in the license for the facility. The licensee evaluated the condition and concluded D1/D2 remained operable but non-conforming. The inspectors determined that the failure to design the facility to withstand the impact of the design basis missile was a performance deficiency that warranted a significance evaluation. Using IMC 0612, the inspectors determined the failure to design the D1/D2 diesel to survive an impact from the design basis missile was more than minor because it is associated with the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events. The inspectors consulted with the Senior Reactor Analyst (SRA) and determined that the risk associated with the condition was green. No cross-cutting aspect was assigned because the performance deficiency from 1994 was not representative of current performance.
05000440/FIN-2010001-012010Q1PerryFailure to Perform Adequate Evaluation of Crane Support Structure ElementsA finding of very low safety significance and an associated NCV of Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors for failure to provide adequate design control measures for crane support structure elements which included bridge crane rail, bridge crane rail clips, bridge crane rail clip studs, leveling plate and leveling plate anchors. Specifically, for evaluation of these structural elements, the licensee failed to demonstrate Seismic Category I compliance in accordance with their design and licensing basis and failed to evaluate the structural elements for resulting reaction forces from the Fuel Handling Building crane. The licensee documented these issues in CRs 11-88791; 11-90252; 10-86582; and 11-04124. 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 SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Table 4a for the Barrier Integrity cornerstone. Based on answering No to all the questions in the Barrier Integrity Cornerstone column of Table 4a, the finding was determined to be of very low safety significance (Green). The inspectors identified a Human Performance, Work Practices (H.4.c) cross-cutting aspect associated with this finding, in that the licensee did not ensure effective supervisory and management oversight of work activities, including contractors, such that nuclear safety was supported. Specifically, the licensee failed to have adequate oversight of design calculations and documentation for establishing structural adequacy of the rail, rail clips, rail clip bolts, leveling plate and leveling plate anchors
05000306/FIN-2009010-022009Q3Prairie IslandFailure to Ensure Design Measures Were Appropriately Established for the Unit 2 Component Cooling Water SystemAn inspector identified apparent violation of 10 CFR Part 50,Appendix B, Criterion III, Design Control, was identified due to the licensees failure to establish design control measures to ensure that the design basis for the Unit 2 CCW system was correctly translated into specifications, drawings, procedures, and instructions. Specifically, the licensee failed to ensure that the safety-related function of the CCW system was maintained following initiating events (such as high energy line break, seismic or tornado events) in the turbine building. This issue has been preliminarily determined to be of low to moderate safety significance (White).This issue was entered into the licensees corrective action program as corrective action document 1145695. Upon identifying this issue, the licensee immediately declared the Unit 2 CCW system inoperable and entered Technical Specification 3.0.3. The Technical Specification was exited following the closure of several system isolation valves approximately 2 hours later. The closure of the isolation valves prevented the Unit 2 CCW system from being vulnerable to failure following events in the turbine building. This finding was determined to be more than minor because it impacted the design control and external events aspects of the Mitigating Systems Cornerstone. The finding also impacted the Mitigating Systems Cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The initiating events in the turbine building could cause the CCW piping to fail. Loss of CCW inventory affects both trains of CCW based on the piping arrangement. The loss of both trains of CCW required a phase 3 significance determination. The results of the phase 3 assessment showed a delta core damage frequency of 3.2E-6, White. The cause of this finding was related to the cross-cutting element of Human Performance, Decision Making because the licensee failed to make safety-significant and risk-significant decisions using a systematic process to ensure that safety was maintained (H.1(a)). Since both the Unit 1 and Unit 2 cross-cutting aspects are from the same performance deficiency and are separated based on the risk determination, the aspect of H.1(a) counts as one cross-cutting aspect in this report.
05000454/FIN-2007009-032008Q1ByronInadequate Design Margins for Continued Operation of Sx Riser PipesThe team identified an apparent violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, associated with the licensees failure to verify the adequacy of the methodology and design inputs used to support licensee decisions to accept the degraded 0B, 0E and 0H essential service water system riser pipes for continued service. Specifically, the licensee failed to evaluate for compressive loads (e.g., buckling), use the applicable Code allowable stress, apply Code equations which account for thermal loads, and failed to correctly apply equations for checking the pipe functional capability. Consequently, the licensee failed to establish adequate design margins for continued service of the 0E, 0H and 0B essential service water system riser which resulted in extended plant operation with degraded SX riser pipes. The cause of this apparent violation was related to the Resources Component (Item H.2(a) of IMC 305) for the cross-cutting area of Human Performance, because the licensee failed to maintain plant safety by maintenance of design margins. Specifically, these degraded riser pipes remained in-service without establishing adequate design margins in the engineering evaluations to justify continued service. The licensee subsequently completed a plant shutdown and replaced the degraded portions of these essential service water system riser pipes. The finding associated with this apparent violation was greater than minor because the degraded essential service water piping condition resulted in an increase in the likelihood of the loss of the essential service water system due to pipe failures, which directly affected the Initiating Events Cornerstone. It was also associated with the Equipment Performance 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). The finding associated with this apparent violation was assessed using a Phase 3 analysis in accordance with NRC Inspection Manual Chapter 0609 Appendix M, Significance Determination Process Using Qualitative Criteria, and is preliminarily determined to have low to moderate safety significance (White)
05000454/FIN-2007009-022008Q1ByronFailure to Implement Timely Corrective Actions for Degraded Sx Riser PipingThe team identified an apparent violation of 10 CFR 50 Appendix B, Criterion XVI, Corrective action, associated with the licensees failure to take timely corrective actions after identification of the corroded essential service water system riser pipes. Specifically, the licensee failed to take timely actions to remove the external corrosion layer present on the riser pipes to support sufficient wall thickness measurements to assess the significance of the pipe wall loss. Consequently, the licensee operated the plant for an extended period of time with a substantial loss of pipe wall on the essential service water riser piping while corrosion proceeded to the point that a through-wall leak developed on the 0C essential service water riser pipe. The cause of this apparent violation was related to the Decision Making Component (Item H.1(b) of IMC 305) for the cross-cutting area of Human Performance, because the licensee failed to make conservative assumptions in decisions affecting the integrity of the essential service water riser piping. The presumption of pipe integrity was not based on sufficient information to be able to demonstrate that the proposed action/decision to leave these risers in service was safe. The licensee subsequently completed a plant shutdown and replaced the degraded portions of these essential service water system riser pipes. The finding associated with this apparent violation was greater than minor because the degraded essential service water piping condition resulted in an increase in the likelihood of the loss of the essential service water system due to pipe failures, which directly affected the Initiating Events Cornerstone. It was also associated with the Equipment Performance 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). The finding associated with this apparent violation was assessed using a Phase 3 analysis in accordance with NRC Inspection Manual Chapter 0609 Appendix M, Significance Determination Process Using Qualitative Criteria, and is preliminarily determined to have low to moderate safety significance (White)
05000254/FIN-2006014-012006Q2Quad CitiesFailure to Establish Measures to Ensure That the Unit 1 ERV Actuators Remained Suitable for Operation While Operating at EPU Power LevelsAn apparent violation (AV) of 10 CFR 50, Appendix B, Criterion III, Design Control, having a preliminary low to moderate safety significance (White), was identified in January 2006 following the discovery that two of the Unit 1 electromatic relief valves (ERVs) would not have performed their safety function. Increased vibrations experienced while operating at extended power uprate (EPU) power levels resulted in the degradation of multiple ERV actuator components which rendered the valves inoperable. The inspectors determined that the licensee implemented the Unit 1 EPU in November 2002, but failed to verify that the ERV actuator design was suitable for operation at the increased vibration levels experienced at EPU power levels. Organizational weaknesses at the station and corporate levels contributed to the licensees failure to identify this issue prior to, or immediately following, EPU implementation. The finding was determined to be more than minor because it impacted the Mitigating Systems cornerstone. In addition, the attributes of design control and equipment performance were adversely impacted by the failure of the ERV actuators. The finding was preliminarily determined to be of low to moderate safety significance following the performance of a case-specific Phase 3 SDP evaluation. The inspectors determined that this finding also affected the cross-cutting area of problem identification and resolution because the licensee failed to fully evaluate historical and predictive information regarding higher than expected main steam line vibrations. Corrective actions included replacing the Unit 1 ERV actuators in January 2006, installing new ERV actuators designed to withstand the increased vibrations experienced during EPU operations in May 2006, and installing an additional modification to reduce the overall main steam line vibration levels. Additional corrective actions were in progress to address the organizational aspects that contributed to this issue. (Section 4OA2)
05000266/FIN-2005017-012005Q3Point BeachLicensee's failure to self-identify the untimely declaration of an Alert classification during an August 2002 emergency preparedness (EP) drill.The NRC also identified an apparent violation of 10 CFR 50.9, Completeness and Accuracy of Information, associated with incomplete and inaccurate information the licensee provided to the NRC in a falsified critique record associated with the August 2002 EP drill. The licensee provided the falsified critique record to NRC inspectors on November 20, 2002. Specifically, the falsified critique record for the August 2002 EP drill indicated that the licensee had self-identified the untimely declaration of an Alert emergency classification. However, the OI investigation determined that the EP Manager and the EP Coordinator deliberately altered the critique record to indicate that the untimely Alert classification declaration was self-identified by the licensee as a part of its formal critique process. The information is material to the NRC because, the NRC relies, in part, on the licensees conduct and self-critiquing of EP drills and exercises to ensure the licensee maintains an effective emergency preparedness and response capability. In a letter to the NRC, dated May 16, 2003, the licensee documented the corrective actions it had taken based upon its own internal investigation of the EP Manager and the EP Coordinators November 2002 deliberate falsification of the August 2002 EP drill and providing of the falsified record to the NRC. Based upon information developed during the NRC inspections and investigation and provided in your letter dated May 16, 2003, we believe that we have sufficient information to make a final significance determination for the preliminary White Finding and to determine the appropriate significance and enforcement actions for the apparent violations. However, before we make a final decision on these matters, we are providing you an opportunity to present to the NRC your perspectives on the facts used by the NRC to arrive at the finding and its significance, and the apparent violations and their significance at a combined regulatory and predecisional enforcement conference (conference) or through the submittal to the NRC of your position on the finding and the apparent violations in writing. If you choose to request a conference, it should be held within 30 days of the receipt of this letter and we encourage you to submit supporting documentation at least one week prior to the conference in an effort to make the conference more efficient and effective. If a conference is held, that portion of the conference associated with the White Finding and the associated apparent violation will be open for public observation. The portion of the conference associated with the 10 CFR 50.9 apparent violation will be closed for public observation because it involves an OI investigation. If you decide to submit only a written response, such submittal should be sent to the NRC within 30 days of the receipt of this letter.
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