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{{#Wiki_filter:1Q/2000 Inspection Findings - Kewaunee                                                                                                  Page 1 of 11 Kewaunee Initiating Events Mitigating Systems Significance:          Feb 21, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform Thorough 10 CFR 50.59 Safety Evaluation The inspectors identified a Non-Cited Violation for failure to perform an adequate 10 CFR 50.59 safety evaluation associated with emergency operating procedure changes to address component cooling water pump dead-head operational concerns. The safety evaluation did not evaluate the potential for initiating a loss-of-coolant accident via the reactor coolant loop seals during conditions of a complete loss of component cooling water.
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Inspection Report# : 2001017(pdf)
Significance: N/A Sep 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ESTABLISH CONTINGENCY PLANS FOR ORANGE RISK CONDITION The inspectors identified the failure to establish contingency plans during a planned high risk plant configuration. Contrary to administrative requirements, the licensee approved an orange risk condition during a refueling outage with no contingency plans to mitigate the consequences of a loss of spent fuel pool cooling with a full core offload in the pool. A Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was identified. The finding was of very low safety significance because although the licensee had not approved appropriate contingency actions for the orange risk condition, the licensee subsequently rescheduled the planned maintenance to eliminate the orange risk condition.
Inspection Report# : 2001013(pdf)
Significance:          Aug 24, 2001 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE CORRECTIVE ACTIONS FOR PREVIOUS PROBLEM WITH THE CONTROL OF SAFETY-RELATED MATERIALS A Non-Cited Violation of Criterion XVI, "Corrective Action," of Appendix B of 10 CFR Part 50 was identified for ineffective corrective actions for a problem with the control of the storage of consumable materials, such as thread sealant, used in safety-related applications. These ineffective actions subsequently resulted in the inadequate control of the storage of grease used in safety-related breakers. This issue was more than minor because if left uncorrected, could under the same condition become a more significant safety concern. In that this issue could credibly affect the operability, availability, reliability, or function of a system or train in a mitigating system, it is a Green finding.
Inspection Report# : 2001012(pdf)
Significance: N/A Aug 24, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO CONDUCT BIENNIAL SURVIELLANCE OF SAFETY-RELATED PROCEDURES PER TS 6.8.c A Non-Cited Violation of Technical Specification 6.8.c was identified for the failure to perform a biennial surveillance of safety-related procedues.
This issue was more than minor because if left uncorrected, could under the same condition become a more significant safety concern. However, since no specific cornerstone had been impacted, this finding is designated as No Color.
Inspection Report# : 2001012(pdf)
Significance:          Aug 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation NON-RATED FIRE BARRIER
 
1Q/2000 Inspection Findings - Kewaunee                                                                                                  Page 2 of 11 A Non-Cited Violation [of 10 CFR Part 50, Appendix R, Section III.G.2.a] was identified for failure to provide a 3-hour rated fire barrier to separate redundant trains of safe shutdown equipment. This finding was of very low safety significance because the licensee tested a replica of the fire barrier and demonstrated that the fire barrier provided protection for at least 60 minutes, which was sufficient for the hazards in the area.
Inspection Report# : 2001011(pdf)
Significance:          Jul 20, 2001 Identified By: NRC Item Type: FIN Finding LICENSED OPERATOR REQUALIFICATION EXAMINATION RESULTS The inspectors identified that two of eight crews examined during the licensee's calendar year 2001 licensed operator requalification operating test had failed. The finding was of very low safety significance because both crews that had failed received remedial training prior to being returned to shift, and the results of the licensee's operator licensing requalification operating test given in calendar year 2000 indicated that only one crew, out of a total of eight crews tested, had failed.
Inspection Report# : 2001011(pdf)
Significance:          Jun 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO TRACK UNAVAILABILITY OF SYSTEMS REQUIRED DURING SHUTDOWN OPERATION 10 CFR 50.65(a)(1), required, in part, that the licensee monitor the performance or condition of SSCs within the scope of the rule as defined by 10 CFR 50.65(b), against licensee-established goals, in a manner sufficient to provide reasonable assurance that such SSCs are capable of fulfilling their intended functions. 10 CFR 50.65(a)(2) stated, in part, that monitoring as specified in 10 CFR 50.65(a)(1) was not required where it had been demonstrated that the performance or condition of an SSC was being effectively controlled through the performance of appropriate preventive maintenance, such that the SSC remained capable of performing its intended function. Contrary to the above, the licensee failed to demonstrate that the performance or condition of systems required to be available during shutdown conditions and within the scope of the rule had been effectively controlled through the performance of appropriate preventive maintenance and did not monitor against licensee-established goals.
Specifically, the licensee failed to monitor the unavailability of systems required during shutdown operation.
Inspection Report# : 2001009(pdf)
Significance: N/A Jun 30, 2001 Identified By: NRC Item Type: FIN Finding FAILURE TO BALANCE RELIABILITY AND AVAILABILITY AS REQUIRED BY 10 CFR 50.65(a)(3)
The inspectors identified a failure to evaluate whether adjustments were necessary such that there would be an appropriate balance between systems' availability and reliability in accordance with 10 CFR 50.65(a)(3) of the maintenance rule. The inspectors identified that the licensee did not have an administrative process to track maintenance rule functional failures and maintenance preventible maintenance functional failures. As a result, reliability and availability could not be balanced as required by the Maintenance Rule periodic evaluation. The safety significance of the specific finding was very low because it did not affect the operability of the systems, and the licensee entered the finding in the corrective action program. However, this finding was considered to be of regulatory concern in the area of maintenance rule implementation due to the extent of the problems with the Maintenance Rule Program.
Inspection Report# : 2001009(pdf)
Significance:          Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IDENTIFY DEFICIENT CONDITION OF VALVE AFW-1B The inspectors identified that the licensee failed to promptly identify and correct the B' train auxiliary feedwater pump discharge check valve which was stuck in an intermediate position. A Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was issued. The finding was of very low safety significance because, although the check valve was stuck in an intermediate position, the time that it was known to have been stuck was less than the technical specification allowed outage time for one train of auxiliary feedwater to be out of service (less than 72 hours). Additionally, the other two trains of auxiliary feedwater were each capable of 100 percent decay heat removal.
Inspection Report# : 2001009(pdf)
Significance:          Jun 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO ESTABLISH MAINTENANCE RULE (a)(1) GOALS
 
1Q/2000 Inspection Findings - Kewaunee                                                                                                      Page 3 of 11 10 CFR 50.65(a)(1), requires, in part, that the licensee monitor the performance or condition of SSCs within the scope of the rule as defined by 10 CFR 50.65(b), against licensee-established goals, in a manner sufficient to provide reasonable assurance that such structures, systems, and components, are capable of fulfilling their intended functions. Such goals shall be established commensurate with safety. When the performance or condition of a structure, system, or component does not meet established goals, appropriate corrective action shall be taken. Contrary to the above, from 1996, the licensee did not take appropriate corrective actions when the performance of those systems in (a)(1) did not meet licensee established goals. Specifically, the licensee determined timely and appropriate corrective actions had not been taken for five systems that had been in (a)(1) category for approximately 3 years to 5 years: component cooling (entered (a)(1) on April 23, 1997), control room air conditioning (July 24, 1996), station and instrument air (July 3, 1997), auxiliary building air ventilation (July 31, 1997), and control rod drive (August 6, 1998). This issue is in the licensee's corrective action system as KAP WO 01-3323. The inspectors evaluated the risk significance of this issue using the Significance Determination Process. The inspectors did not identify where this failure resulted in a total loss of a risk significant SSC. Therefore, this issue was screened as Green (very low risk significance) after a Phase 1 Significance Determination Process review. Although the risk significance of this issue was low, the inspectors concluded that this was more than a minor concern because the failure to recognize and correct ineffective maintenance practices resulted in risk significant systems in (a)(1) for years with no improvement in performance. The NRC tracking number for this issue is 50-305/01-09-02.
Inspection Report# : 2001009(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation FIRE DOOR FUSIBLE LINKS.
On February 20, 2001, the licensee determined that the installed fusible link arrangement on roll-up fire Doors 279 and 281, which separated both trains of service water pumps, would not actuate as designed to ensure that the doors would automatically close to provide a 3-hour fire barrier, contrary to 10 CFR Part 50, Appendix R, Section III.G.2.a which required, in part, separation of cables and equipment of redundant trains by a fire barrier having a 3-hour rating.
Inspection Report# : 2001006(pdf)
Significance:        Mar 12, 2001 Identified By: Licensee Item Type: NCV NonCited Violation INADEQUATE SMOKE DETECTOR COVERAGE IN FIRE ZONE TU-95B.
Licensee identified violation of licensee's operating license that the licensee failed to install a detector in each beam pocket in Fire Zone TU-95B.
Inspection Report# : 2001002(pdf)
Significance:        Feb 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO TEST FIRE DOOR IN ACCORDANCE WITH FIRE PLAN.
The inspectors identified a non-cited violation for failure to properly test a fire door in accordance with the facility's fire protection program plan. The finding was of very low safety significance because, although the fire door separated both trains of service water pumps and did not fully close as designed when subsequently tested, the fire loading in the area was insufficient to credibly impact more than two of the four service water pumps in the area.
Inspection Report# : 2001004(pdf)
Significance: N/A Nov 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO COMPLETE COMPONENT RE-TEST REQUIREMENTS PRIOR TO RETURNING EQUIPMENT TO AN OPERABLE STATUS.
No Color. The inspectors identified a Non-Cited Violation for failure to complete component retest requirements following maintenance performed on the B train control room air conditioner compressor condenser. The unit had been returned to an operable status prior to the retest requirements being completed as prescribed in the associated maintenance procedure. This issue was determined to be a violation of the licensee's Operational Quality Assurance Program Manual, Section 8, "Maintenance Planning and Control." Although the risk associated with this finding was very low and did not affect any cornerstones, the inspectors noted that this finding was similar to previous NRC-identified findings and therefore was of greater than minor significance and warranted documentation. (Section 1R19).
Inspection Report# : 2000020(pdf)
Significance:        Sep 29, 2000 Identified By: NRC
 
1Q/2000 Inspection Findings - Kewaunee                                                                                                  Page 4 of 11 Item Type: NCV NonCited Violation FAILURE TO IDENTIFY CORROSION AS POTENTIAL FAILURE MECHANISM.
The inspectors identified that the licensee failed to identify corrosion as a potential failure mechanism in the operability determination for a carbon steel key in the service water system. Thus, the licensee failed to quantify the corrosion rate and therefore did not adequately evaluate the expected service life of the carbon steel key. One non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was identified.
Inspection Report# : 2000019(pdf)
Significance:        Sep 29, 2000 Identified By: NRC Item Type: NCV NonCited Violation PRESSURE RATING OF AUXILIARY FEEDWATER STEAM TRAPS.
The inspectors identified that a root cause evaluation for a 1996 equipment issue in the turbine-driven auxiliary feedwater system was not completed until 1999. The evaluation stated that the internals of the steam traps were designed to operate at pressures up to a maximum 600 psig but that the traps were exposed to pressures up to 1025 psig. A corrective action item to initiate a design change request to replace the steam traps with a different model rated for the design pressure of the system was described in the evaluation. However, the inspectors identified that the design change request had never been initiated and the KAP ( Kewaunee corrective action document) had been closed. As a result, the corrective action item for this design problem was lost. In addition, operability of the system had never been formally evaluated despite the identification that the system design requirements were not met. The licensee subsequently determined that the steam traps remained operable and was planning to initiate the design change to correct the problem. One non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was identified.
Inspection Report# : 2000019(pdf)
Significance:        Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation DESIGN CONTROL VIOLATION FOR AUXILIARY FEEDWATER STRAINER MESH SIZE.
The inspectors questioned the mesh size of the strainers, which were installed in the suction of the three auxiliary feedwater (AFW) pumps. As a result of the inspectors' questions, license personnel inspected the strainers on August 21, 2000, and found the strainers to have 1/16 inch openings. A note was later found on Figure 10.2-3 of the UFSAR that indicated that the AFW suction strainer size was 1/8 inch. The smaller openings would not support the use of service water as a safety related source for AFW and as a result all three trains of AFW were declared inoperable. This condition had apparently existed for approximately 25 years and was identified as a non-cited violation of Criterion III, "Design Control," of 10 CFR 50, Appendix B.
Inspection Report# : 2000012(pdf)
Significance: N/A Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation RETRIEVAL OF SERVICE WATER SYSTEM DESIGN INFORMATION.
In many cases, design basis information for the service water system was difficult if not impossible to locate. Licensee personnel wrote KAP WO 00-002566 to enter the problem in the corrective action program.
Inspection Report# : 2000012(pdf)
Significance:        Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PROPERLY CONTROL CORRECTED TEST DATA NECESSARY FOR DESIGN CALCULATIONS.
An example of a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," was identified in the handling of service water system flow test data, which was subsequently used in calculations. Gauge readings corrected for post test calibration checks, gauge reading corrections for elevation considerations, and flow values corrected for pump degradation were contained in spreadsheets in the possession of an individual staff member, but not currently packaged with raw test data, and not bearing evidence of a formal review and control process. The connection between the test data, which had been vaulted, and the values used in the calculation, could not be made without use of the uncontrolled spreadsheet.
Inspection Report# : 2000012(pdf)
Significance:        Aug 09, 2000 Identified By: NRC
 
1Q/2000 Inspection Findings - Kewaunee                                                                                                  Page 5 of 11 Item Type: NCV NonCited Violation IMPROPER DESIGN CALCULATION IDENTIFICATION, NON-CONSERVATIVE ASSUMPTIONS, CALCULATION ERRORS, AND DUPLICATE CALCULATIONS.
An example of a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," was identified because of inadequate control of design calculations. The control failures included improper identification of calculations, non-conservative assumptions, calculation errors, and duplicate or superceded calculations not properly identified or canceled. The failure to follow the established design control process increased the potential for errors in the design and operation of the service water system. Because the system was subsequently demonstrated to be capable of removing the design heat load, the actual significance was low and this finding screened out as having very low risk significance.
Inspection Report# : 2000012(pdf)
Significance:          Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO DOCUMENT IN THE CORRECTIVE ACTION PROGRAM THE USE OF INCORRECT MATERIAL IN A SERVICE WATER PUMP KEY.
An example of a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was identified because of inadequate corrective action to correct an incorrect coupling adjust nut set screw and a low strength "soft" key material, which had contributed to a pump shaft failure. Licensee personnel had known of the "soft" key material since July 21, 1999. The "soft" key material was found in other service water pumps but had not been removed from all pumps. As of July 25, 2000, licensee personnel had not documented the existence of the "soft" key material in the corrective action program.
Inspection Report# : 2000012(pdf)
Significance:          Jul 07, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO COMPLETE COMPONENT RETEST REQUIREMENTS IN ACCORDANCE WITH PROCEDURE.
On June 26, 2000, during a review of post maintenance testing requirements following maintenance performed on the control room post accident system charcoal filter heat detector, the inspectors identified that maintenance technicians had not completed the component re-test requirements, as required by a preventative maintenance procedure prior to the system being returned to an operable status. On July 7, the inspectors identified a second example of failing to complete component re-test requirements following maintenance on the zone special ventilation system charcoal filter heat detector as required. The issue was considered to be of very low safety significance based on the determination that although the licensee had not completed all of the component retest requirements prior to returning the equipment to service, the subsequent testing determined that the equipment was in an operable status. The failure to complete the component retest requirements in accordance with site procedures was identified as a Non-Cited Violation.
Inspection Report# : 2000014(pdf)
Significance:          Jun 22, 2000 Identified By: NRC Item Type: FIN Finding FAILURE TO MEET SINGLE FAILURE CRITERIA FOR RESIDUAL HEAT REMOVAL VALVE CIRCUITRY.
The licensee identified that the circuitry associated with the residual heat removal system discharge to safety injection system suction isolation valves did not meet single failure criteria. The inspectors noted that this design requirement was identified in the facility's updated safety analysis report. The licensee subsequently implemented a temporary change to the facility. The inspectors reviewed the issue and identified that the facility had been operating outside of its design basis, which was reportable to the NRC. The licensee subsequently made a one hour non-emergency report to the NRC. Since there was no actual loss of safety function to the system, this issue was screened as very low risk significance.
Inspection Report# : 2000008(pdf)
Significance:          Jun 22, 2000 Identified By: NRC Item Type: FIN Finding MAINTENANCE RULE FAILURES ASSOCIATED WITH REACTOR HEAD VENT VALVE.
The inspectors reviewed the licensee's implementation of the maintenance rule for failures associated with a reactor head vent valve. The licensee's corrective action documents identified a potential maintenance rule functional failure but the completed evaluation of the problem did not document the final determination. However, the inspectors identified that the repeated failures may have been prevented if maintenance activities such as valve disassembly and cleaning had been performed. In this case, maintenance rule reliability goals were not exceeded. The licensee had documented similar maintenance rule program deficiencies and developed a corrective action program to address the deficiencies. Although programmatic deficiencies exist, since no maintenance rule reliability criteria had been exceeded, this issue was considered of very low risk
 
1Q/2000 Inspection Findings - Kewaunee                                                                                                    Page 6 of 11 significance.
Inspection Report# : 2000008(pdf)
Significance:        Jun 22, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO INITIATE KEWAUNEE ASSESSMENT PROCESS DOCUMENT REGARDING REFUELING WATER STORAGE TANK LOW-LOW LEVEL ALARM INOPERABILITY.
The inspectors identified that the refueling water storage tank low-low level alarm which was actuating five percent higher than normal had not been documented in a Kewaunee Assessment Process form by the licensee, and therefore had not received an operability evaluation. This failure was identified as contrary to site administrative procedures. Following the licensee's documentation of the problem, the inspectors identified that the associated operability evaluation considered the acceptability of an operator workaround to address the issue, but did not address any safety implications or consequences of the alarm actuating early. A subsequent operability evaluation by the licensee was evaluated as adequate by the inspectors. Since the subsequent operability evaluation was adequate and it was determined that no safety mitigation equipment was adversely affected by the early actuation of the alarm, this issue was considered of very low risk significance. A non-cited violation (NCV) was identified for failing to document a non-conforming condition, contrary to site administrative procedure requirements.
Inspection Report# : 2000008(pdf)
Significance:        Jun 22, 2000 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO TEST ADDITIONAL RELIEF VALVES IN ACCORDANCE WITH TECHINCAL SPECIFICATIONS.
The licensee identified that the suction relief valve for an auxiliary feedwater pump may have failed its relief test criteria, but did not process the documented deficiency until several weeks later. The licensee then expanded the scope of the relief testing to the suction relief valves associated with the other auxiliary feedwater pumps to meet technical specification requirements. Since any one train of auxiliary feedwater was capable of supplying 100 percent of the decay heat removal requirements, this issue was screened as very low risk significance. However, the time delay in complying with technical specification requirements for testing other relief valves was identified as an NCV.
Inspection Report# : 2000008(pdf)
Significance:        May 22, 2000 Identified By: NRC Item Type: FIN Finding FIRE EXTINGUISHERS NOT LOCATED IN ALL AREAS OF CONTAINMENT BUILDING AT BEGINNING OF REFUELING OUTAGE.
During a walkdown of the containment building, the inspectors identified that portable fire extinguishers were not located in the containment basement at the beginning of the plant refueling outage. Additionally, site fire protection procedures required that responsible fire protection personnel perform inspections of selected plant areas to ensure that the quantity of combustible material was minimized. However, the procedure did not list the containment as an area to be inspected and the procedures did not require the placement and location of portable fire suppression equipment inside containment during the refueling outage. Due to a low number of work activities ongoing at the time, this issue was screened as Green (very low risk significance).
Inspection Report# : 2000007(pdf)
Significance:        May 22, 2000 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO INSTALL RAYCHEM HEAT SHRINK MATERIAL IN ACCORDANCE WITH PROCEDURE REQUIREMENTS.
The licensee identified that two Raychem electrical cable splices utilized in environmentally qualified (EQ) safety-related equipment had not been installed in accordance with EQ requirements. These splices were associated with pressurizer level transmitters and were installed in 1984. The licensee subsequently performed extensive EQ testing of the splices to determine the qualification of the splices' as-found configurations. Test results indicated that the splices would have been able to perform their intended function in a harsh environment inside containment. This issue was considered to be of low safety significance based on the successful EQ testing of the as-found splices' configurations and was screened as Green (very low risk significance). One non-cited violation was identified.
Inspection Report# : 2000007(pdf)
Barrier Integrity
 
1Q/2000 Inspection Findings - Kewaunee                                                                                                  Page 7 of 11 Significance:        Jun 22, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation FAILURE TO UPDATE COMPUTER ALARM FOR CURRENT AXIAL FLUX DISTRIBUTION TARGET BAND.
The licensee identified, following plant startup, that a computer alarm had not been updated properly to alarm if axial flux distribution deviated outside of the flux distribution target band. This condition was contrary to technical specification requirements. The licensee reviewed the axial flux distribution history since the startup and determined that at no time was the flux distribution outside of the target band. Since the axial flux distribution was never outside of the target band, this issue was screened as very low risk significance. An NCV was identified for failing to comply with technical specification requirements for monitoring axial flux distributions.
Inspection Report# : 2000008(pdf)
Significance:        May 22, 2000 Identified By: Licensee Item Type: FIN Finding TECHNICAL SPECIFICATIONS INTERPRETATION FOR TESTING REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES.
The NRC determined that the licensee's practice of testing reactor coolant system pressure isolation Valve SI-22B prior to entering the cold shutdown condition was contrary to Technical Specification requirements 4.2.a.3.a. Technical Specification 4.2.a.3.a required that periodic leakage testing of Valve SI-22B be accomplished prior to reaching operating mode after the plant was placed in cold shutdown. This issue was considered to be of low safety significance because of a subsequent successful valve test and was screened as Green (very low risk significance).
Enforcement discretion was applied to this item in accordance with Section VII.B.6 of the Enforcement Policy.
Inspection Report# : 2000007(pdf)
Significance:        May 19, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PROPERLY IMPLEMENT FLAW ACCEPTANCE CRITERIA FOR SLEEVE WELD INSPECTIONS.
During 1998 inservice inspection examinations, the licensee failed to properly implement the flaw acceptance criteria for laser welded sleeve inspection within two steam generator tubes. The safety significance was very low based on the absence of adverse consequences, and May 2000 in-situ pressure testing where both welds exhibited zero leakage at normal operating pressure, main steam line break pressure, and three times normal operating differential pressures. As such, this issue was characterized as a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion IX, Control of Special Processes. To correct the error, the licensee plugged both tubes (Section 1RO8).
Inspection Report# : 2000011(pdf)
Emergency Preparedness Significance:        Mar 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation SUPPLEMENTAL INSPECTION OF WHITE ERO AUGMENTATION FINDING AND RESULTING GREEN FINDING.
This supplemental inspection was performed by the NRC to evaluate the licensees evaluation associated with the failure to conduct successful quarterly, off-hours, unannounced staff augmentation drills during the second, third, and fourth quarters of 1999 and the second quarter of 2000.
This performance issue was previously characterized as having low to moderate risk significance (White) in NRC Inspection Report No. 50-305/2000015(DRS). During this supplemental inspection, performed in accordance with Inspection Procedure95002, the inspector concluded that the licensee performed a comprehensive evaluation of the unsuccessful staff augmentation drills. The licensees evaluation identifiedtwo root causes which resulted in the unsuccessful drills and in the staffs inability tocorrect thedeficiencies: (1) Management has not effectively acted to provide increased depth and flexibility in the emergency response organization following a reduction in staffing several years ago; and (2)
Management has accepted an adverse trend of test failures without requiring investigation into the root causes. The inspector reviewed the licensees corrective actions, both completed and planned, and concluded that the programmatic corrective actions appeared to address the identified root causes. In particular, the licensee assigned certain positions to an on-call rotation to ensure personnel were capable of augmenting in a timely manner, and the licensee was progressing in training additional staff to increase the depth of personnel assigned to key emergency response positions. In addition, the licensee was continuing its efforts in improving its corrective action program. The inspector reviewed the licensees immediate response to the issue and identified that one of the licensees initial corrective actions resulted in a Non-Cited Violation of regulatory requirements. To obtain a timely response of a key emergency response position (severe accident management - core hydraulics), the emergency preparedness staff effectively changed the emergency plan without revising the necessary procedures and without formally assessing the impact of that change. The staff instructed and trained personnel to respond to a location other than the Technical Support Center, which was
 
1Q/2000 Inspection Findings - Kewaunee                                                                                                    Page 8 of 11 contrary to the licensees current implementing procedures. While this change enabled the licensee to augment its staff in a timely manner, the change was not performed in accordance with NRC requirements. In order to make such a change, the licensees emergency plan required that the change be formally assessed to ensure that it did not reduce the effectiveness of the plan or any other implementing procedure. Since the issue did not result in a failure to meet an emergency preparedness planning standard, the failure to adequately implement the emergency plan was determined to be a violation of very low safety significance (Green) (Section 02.3(a)). Due to the licensees acceptable performance in assessing the emergency response augmentation drill deficiencies, the White finding associated with this issue will only be considered in assessing plant performance for a total of four quarters in accordance with the guidance in IMC 0305, Operating Reactor Assessment Program. Implementation of the licensees corrective actions will be reviewed during a future inspection.
Inspection Report# : 2001007(pdf)
Significance:          Mar 30, 2001 Identified By: NRC Item Type: VIO Violation FAILURE TO CORRECT SELF-IDENTIFIED ERO AUGMENTATION DRILL DEFICIENCIES.
During a baseline inspection of the emergency preparedness program conducted on August 14 - September 21, 2000, the NRC identified a preliminary White issue and potential violation for the licensees failure to successfully correct deficiencies identified during staff augmentation drills and to demonstrate timely staff augmentation in 1999 and 2000. The issue was unresolved pending the outcome of the NRCs final significance determination. On January 30, 2001, the NRC conducted a regulatory conference with the licensee and subsequently issued the licensee a White finding and Notice of Violation (Enforcement Action No. 00-214) associated with the performance issue.
Inspection Report# : 2001007(pdf)
Significance:          Mar 23, 2001 Identified By: NRC Item Type: NCV NonCited Violation POST-ACCIDENT SAMPLING SYSTEM CONTAINMENT AIR SAMPLE PANEL TESTING.
A Non-Cited Violation of Technical Specification 6.14 was identified for the failure to implement a program that ensured the capability to obtain and analyze containment atmosphere samples under accident samples using the cantainment air sampling panel (CASP). Although the CASP was installed, as was indicated in the emergency plan, the licensee had neither developed procedures nor had tested its capability to obtain a containment atmosphere sample using the CASP. The licensee could not recall if and when containment air samples were last obtained using the CASP.
Inspection Report# : 2001006(pdf)
Significance: N/A Feb 27, 2001 Identified By: NRC Item Type: FIN Finding SUPPLEMENTAL INSPECTION OF YELLOW ANS PERFORMANCE INDICATOR AND ASSOCIATED CORRECTIVE ACTION PROGRAM DEFICIENCIES.
This supplemental inspection was performed by the NRC to assess the licensee's evaluation associated with a Yellow performance indicator for the Alert and Notification System (ANS) and the associated Yellow finding related to the licensee's corrective action program. During this supplemental inspection, performed in accordance with Inspection Procedure 95002, the inspectors concluded that the licensee performed comprehensive evaluations of the performance problems associated with the ANS and its corrective action program. These evaluations identified primary root causes and contributing causes for both issues. Along with the electronics/hardware problems, the licensee identified the primary root causes for the ANS to be the failure to make changes to the system via a change control process and the failure of the activation procedure to provide for alternate/backup activation methods and to provide clear success criteria. In the case of the corrective action program, the licensee concluded that the primary root causes were plant management's inadequate risk evaluation regarding decisions affecting the corrective action program and the development of a culture in the licensee's organization, which minimized the importance of information from outside organizations. In particular, the licensee focused on low-cost power operation and failed to recognize the value of changes and improvements which had occurred throughout the industry, such as the value and expansion of the corrective action program. In the case of the Yellow ANS performance indicator, the licensee had completed several significant corrective actions to address the root causes and contributing causes identified in its evaluation. The inspectors found that the corrective actions appeared appropriate to address the underlying root causes and that ANS testing data indicated an improving trend in the NRC performance indicator. The licensee also performed comprehensive assessments of the emergency preparedness program, quality assurance program, plant operations, and other plant programs to determine the extent of condition (re. the root causes described above). Based on these evaluations, the licensee began to implement significant actions to correct the deficiencies in the corrective action program and other weaknesses identified. Generally, the inspectors observed progress in the licensee's initial implementation of these corrective actions. Due to the licensee's acceptable performance in assessing the Yellow ANS performance indicator and the associated Yellow finding, the Yellow finding will not be considered in assessing future plant performance.
Inspection Report# : 2001005(pdf)
Significance:          Sep 21, 2000 Identified By: NRC
 
1Q/2000 Inspection Findings - Kewaunee                                                                                                  Page 9 of 11 Item Type: FIN Finding INADEQUATE ROOT CAUSE EVALUATION FOR YELLOW ALERT AND NOTIFICATION SYSTEM PERFORMANCE INDICATOR.
The licensee's evaluation of the Yellow Alert and Notification (siren) System Performance Indicator (PI) was inadequate. The inspector concluded that the licensee's evaluation was not performed at the depth necessary to identify the root causes of the siren performance problems and, instead, only identified the symptoms of the root causes. Specifically, the inspector identified the following substantive weaknesses in the licensee's evaluation of the siren system performance, which appeared to result from systemic corrective action program deficiencies within this cornerstone:
* The licensee's evaluation was not of sufficient depth to clearly identify the root causes associated with the decline in siren system performance.
* Licensee management did not provide well-understood and clear guidance/expectations for performing root cause evaluations.
* The licensee's evaluation of the quality assurance program was narrowly focused and was not critical of its role in failing to identify and correct the siren performance problems.
* The licensee did not establish a priority for each of the long-term corrective actions in accordance with the associated significance or risk.
* The licensee did not have any formal provisions for measuring the effectiveness of its corrective actions.
* Within the licensee's evaluation, the licensee had not evaluated common causes or the extent of the condition. Due to the corrective action program performance deficiencies within this cornerstone, we have been unable to conclude that the performance issues that resulted in the yellow PI have been addressed. Therefore, we are issuing a yellow finding that corresponds to the original issues that resulted in a yellow PI. Additional inspection effort will be focused on the licensee's further evaluation of the siren reliability root causes and the continuing corrective action program implementation deficiencies identified during this inspection.
Inspection Report# : 2000017(pdf)
Significance: N/A Apr 05, 2000 Identified By: NRC Item Type: FIN Finding LICENSEE FAILED TO IDENTIFY THE FULL SCOPE OF PROBLEMS WITH THE ALERT AND NOTIFICATION SYSTEM PERFORMANCE.
The inspectors concluded that the licensee's assessment was not sufficiently comprehensive to identify the full scope of problems associated with the Alert and Notification System (ANS) performance program. As a result, licensee corrective actions generally were focused on the equipment problem rather than all root causes and contributing factors. For example, the inspectors identified that: (1) management oversight of the ANS performance program was limited; (2) an audit failed to identify degrading ANS performance as a concern; (3) annual preventive maintenance was not consistently performed on the system; (4) the corrective action program was not used consistently to document ANS problems; and (5) maintenance procedures and records were deficient. Collectively, these problems indicate that the ANS performance program lacked sufficient structure and oversight.
Inspection Report# : 2000006(pdf)
Occupational Radiation Safety Significance:        Oct 02, 2001 Identified By: NRC Item Type: NCV NonCited Violation HIGH RADIATION AREA ACCESS CONTROLS Non-Cited Violation of Technical Specification 6.13 and an associated Green Finding for failure to 'barricade' three ladders that provided entry to high radiation areas (less than 1000 mrem/hour) located on the steam generator/pressurizer platforms.
Inspection Report# : 2001014(pdf)
Significance:        May 19, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO POST A VERY HIGH RADIATION AREA.
The inspectors identified a noncited violation for the failure to post a very high radiation area in accordance with 10 CFR 20.1902(c). Although the area was not adequately posted, the licensee had provided physical controls and barriers that were consistent with its requirements for a very high radiation area. Based on the adequacy of these controls, the potential for an overexposure from the inadvertent entry of personnel into the area was low. Consequently, this finding was determined to be of very low safety significance (Section 20S1.1).
Inspection Report# : 2000009(pdf)
Public Radiation Safety
 
1Q/2000 Inspection Findings - Kewaunee                                                                                                Page 10 of 11 Physical Protection Significance: N/A Jul 21, 2000 Identified By: NRC Item Type: FIN Finding CORRECTIVE ACTION ON SEARCH ISSUES DID NOT WORK.
The inspector determined that the licensee's effectiveness of implemented corrective actions for a previously identified inspection finding regarding an inadequate vehicle search was not totally effective in preventing recurrence. Previous corrective action was not adequately focused (Section 3PP2.2).
Inspection Report# : 2000013(pdf)
Significance:        Jul 21, 2000 Identified By: NRC Item Type: NCV NonCited Violation SEARCHES INADEQUATE (REPEAT).
The inspector identified a Non-Cited violation by observing that, a security officer failed to search an easily accessible compartment on one vehicle.
The failure resulted from human error because the officer did not observe the access panel to the compartment during the vehicle search process.
When searched, no prohibited items were found. Corrective actions were implemented. The inspector determined that the licensee's effectiveness of implemented corrective actions for a previously identified inspection finding regarding an inadequate vehicle search was not totally effective in preventing recurrence. Previous corrective action was not adequately focused.
Inspection Report# : 2000013(pdf)
Miscellaneous Significance: N/A Feb 21, 2002 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Implement Required Fire Watch Following Completion of Hot Work Activities A licensee-identified violation was reviewed by the inspectors. Corrective actions taken or planned by the licensee appeared reasonable.
Inspection Report# : 2001017(pdf)
Significance: N/A Aug 24, 2001 Identified By: NRC Item Type: FIN Finding IDENTIFICATION AND RESOLUTION OF PROBLEMS The team concluded that the licensee was generally effective at identifying problems and putting them into the corrective action program. The program itself contained all the necessary attributes of an acceptable corrective action program and was generally successful in correcting identified issues. However, the team noted that, although licensee management had taken efforts to ensure that issues were resolved in accordance with program guidance and requirements, additional efforts appeared necessary to ensure timely resolution of issues. A positive program initiative was the establishment of positions in each of the major plant departments to serve as liaisons between the departments and the corrective action program and to assist with self-assessments. However, examples were identified by the inspectors of problems with the licensee's identification and resolution of problems, prioritization and evaluation of issues, and the effectiveness of corrective actions. Included in these examples were the routine granting of due date extensions for problem evaluation and corrective action implementation, failure to perform a Technical Specification-required biennial surveillance of safety-related procedures, and ineffective corrective actions that resulted in the lack of proper controls over the storage of grease used in safety-related breakers. Based on a review of records and discussions with plant staff, the inspectors concluded that workers at the site felt free to input safety issues into the corrective action program.
Inspection Report# : 2001012(pdf)
Significance: N/A Nov 09, 2000 Identified By: NRC Item Type: FIN Finding FAILURE TO TAKE ADEQUATE CORRECTIVE ACTIONS TO ENSURE COMPONENT RE-TEST REQUIREMENTS COMPLETED ACCORDING TO MAINTENANCE PROCEDURES.
No Color. The inspectors determined that a negative performance trend had developed in the licensee's ability to identify and promptly take appropriate corrective actions to prevent recurrence based on two previously identified examples (NCV 50-305/2000014-01) and one example identified during this inspection period (NCV 50-305/00-20-01). All three examples related to the licensee returning safety-related equipment to service prior to completing all required post-maintenance retesting. While the risk of the individual examples was very low, the licensee had failed to ensure that all retest requirements had been completed before returning safety-related equipment to service. These findings collectively indicated a problem with the licensee's ability to provide timely and adequate corrective actions to prevent recurrence. (Section 4OA2).
Inspection Report# : 2000020(pdf)
 
1Q/2000 Inspection Findings - Kewaunee                                                                                                Page 11 of 11 Significance: N/A Sep 29, 2000 Identified By: NRC Item Type: NCV NonCited Violation NO PROCEDURAL GUIDANCE FOR DETERMINING IF CONDITIONS ADVERSE TO QUALITY ARE SIGNIFICANT.
The inspectors reviewed the quality assurance (QA) manual requirements against Kewaunee's implementing procedures and identified that two QA manual requirements were not being implemented. Specifically, Program Requirement 3.1.9 which stated that directives and procedures shall provide for the review of conditions adverse to quality to determine if the conditions are significant in nature. This requirement paralleled 10 CFR Part 50, Appendix B, Criterion XVI, which requires that the cause of significant conditions adverse to quality be determined and corrective actions taken to prevent recurrence. The inspectors reviewed the Nuclear Administrative Directive (NAD 11.08) and the procedure (GNP 11.08.01) governing the KAP (Kewaunee corrective action program) process and found no procedure requirements for identifying significant conditions adverse to quality. This finding does not directly affect a cornerstone. As a result, this issue was not evaluated with the Significance Determination Process and was not assigned a color. One example of a non-cited violation of 10 CFR Part 50, Appendix B, Criterion II, "Quality Assurance Programs," was identified.
Inspection Report# : 2000019(pdf)
Significance: N/A Sep 29, 2000 Identified By: NRC Item Type: FIN Finding EFFECTIVENESS OF CORRECTIVE ACTION PROGRAM.
Based on the results of this inspection, the NRC concluded that the corrective action program at Kewaunee showed significant weaknesses and inconsistencies across all of the procedural elements inspected. These weaknesses existed across departments and affected multiple cornerstones in the strategic performance areas of Reactor Safety, Radiation Safety, and Safeguards. Of particular note was the lack of procedures for determining the significance of conditions adverse to quality and for trending of issues and the complete lack of trending within your corrective action program. We also identified a lack of urgency in correcting issues which resulted in repeat examples occurring and, coupled with a poor tracking system, a tendency for issues to be dropped. While none of the specific examples identified by the team were of high risk significance when looked at in isolation, in the aggregate they were similar in nature to prior issues in the emergency preparedness area that rose to a higher significance level and contributed to a degraded cornerstone. While we concluded that the station had fostered an environment in which personnel freely identified conditions adverse to quality without fear of discrimination or retaliation, we also concluded that significant weaknesses with, and inconsistent implementation of, the station corrective action program resulted in multiple examples where station personnel did not enter deficiencies into the station's formal corrective action program.
Inspection Report# : 2000019(pdf)
Significance: N/A Sep 29, 2000 Identified By: NRC Item Type: NCV NonCited Violation NO PROCEDURAL GUIDANCE FOR TRENDING CONDITIONS ADVERSE TO QUALITY.
The inspectors reviewed the quality assurance (QA) manual requirements against Kewaunee's implementing procedures and identified that two QA manual requirements were not being implemented. Specifically, QA Program Requirement, 3.1.10, stated that directives and procedures shall provide for analyzing trends of conditions adverse to quality. Once identified these trends were required to be considered significant conditions adverse to quality. The inspectors found that conditions adverse to quality were not defined in the KAP (Kewaunee corrective action program) procedures and that no procedure existed for trending. This finding does not directly affect a cornerstone. As a result, this issue was not evaluated with the Significance Determination Process and was not assigned a color. One example of a non-cited violation of 10 CFR Part 50, Appendix B, Criterion II, "Quality Assurance Program," was identified.
Inspection Report# : 2000019(pdf)
Significance: N/A Jun 22, 2000 Identified By: NRC Item Type: FIN Finding CONTROL ROOM OPERATIONS HUMAN PERFORMANCE ISSUES.
The inspectors interviewed operators to evaluate their awareness of degraded control room indications and alarms, and their ability to adequately take manual actions based on degraded alarm functions. The inspectors identified, during interviews, that there was a lack of awareness by operators of a degraded refueling water storage tank low-low level alarm which would be potentially confusing to operators and therefore increase the risk associated with initiating long term sump recirculation.
Inspection Report# : 2000008(pdf)
Last modified : April 01, 2002
 
2Q/2000 Inspection Findings - Kewaunee                                                                                                    Page 1 of 11 Kewaunee Initiating Events Mitigating Systems Significance:        Jun 22, 2000 Identified By: NRC Item Type: FIN Finding FAILURE TO MEET SINGLE FAILURE CRITERIA FOR RESIDUAL HEAT REMOVAL VALVE CIRCUITRY.
The licensee identified that the circuitry associated with the residual heat removal system discharge to safety injection system suction isolation valves did not meet single failure criteria. The inspectors noted that this design requirement was identified in the facility's updated safety analysis report. The licensee subsequently implemented a temporary change to the facility. The inspectors reviewed the issue and identified that the facility had been operating outside of its design basis, which was reportable to the NRC. The licensee subsequently made a one hour non-emergency report to the NRC. Since there was no actual loss of safety function to the system, this issue was screened as very low risk significance.
Inspection Report# : 2000008(pdf)
Significance:        Jun 22, 2000 Identified By: NRC Item Type: FIN Finding MAINTENANCE RULE FAILURES ASSOCIATED WITH REACTOR HEAD VENT VALVE.
The inspectors reviewed the licensee's implementation of the maintenance rule for failures associated with a reactor head vent valve. The licensee's corrective action documents identified a potential maintenance rule functional failure but the completed evaluation of the problem did not document the final determination. However, the inspectors identified that the repeated failures may have been prevented if maintenance activities such as valve disassembly and cleaning had been performed. In this case, maintenance rule reliability goals were not exceeded. The licensee had documented similar maintenance rule program deficiencies and developed a corrective action program to address the deficiencies. Although programmatic deficiencies exist, since no maintenance rule reliability criteria had been exceeded, this issue was considered of very low risk significance.
Inspection Report# : 2000008(pdf)
Significance:        Jun 22, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO INITIATE KEWAUNEE ASSESSMENT PROCESS DOCUMENT REGARDING REFUELING WATER STORAGE TANK LOW-LOW LEVEL ALARM INOPERABILITY.
The inspectors identified that the refueling water storage tank low-low level alarm which was actuating five percent higher than normal had not been documented in a Kewaunee Assessment Process form by the licensee, and therefore had not received an operability evaluation. This failure was identified as contrary to site administrative procedures. Following the licensee's documentation of the problem, the inspectors identified that the associated operability evaluation considered the acceptability of an operator workaround to address the issue, but did not address any safety implications or consequences of the alarm actuating early. A subsequent operability evaluation by the licensee was evaluated as adequate by the inspectors. Since the subsequent operability evaluation was adequate and it was determined that no safety mitigation equipment was adversely affected by the early actuation of the alarm, this issue was considered of very low risk significance. A non-cited violation (NCV) was identified for failing to document a non-conforming condition, contrary to site administrative procedure requirements.
Inspection Report# : 2000008(pdf)
Significance:        Jun 22, 2000 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO TEST ADDITIONAL RELIEF VALVES IN ACCORDANCE WITH TECHINCAL SPECIFICATIONS.
The licensee identified that the suction relief valve for an auxiliary feedwater pump may have failed its relief test criteria, but did not process the
 
2Q/2000 Inspection Findings - Kewaunee                                                                                                  Page 2 of 11 documented deficiency until several weeks later. The licensee then expanded the scope of the relief testing to the suction relief valves associated with the other auxiliary feedwater pumps to meet technical specification requirements. Since any one train of auxiliary feedwater was capable of supplying 100 percent of the decay heat removal requirements, this issue was screened as very low risk significance. However, the time delay in complying with technical specification requirements for testing other relief valves was identified as an NCV.
Inspection Report# : 2000008(pdf)
Significance:          May 22, 2000 Identified By: NRC Item Type: FIN Finding FIRE EXTINGUISHERS NOT LOCATED IN ALL AREAS OF CONTAINMENT BUILDING AT BEGINNING OF REFUELING OUTAGE.
During a walkdown of the containment building, the inspectors identified that portable fire extinguishers were not located in the containment basement at the beginning of the plant refueling outage. Additionally, site fire protection procedures required that responsible fire protection personnel perform inspections of selected plant areas to ensure that the quantity of combustible material was minimized. However, the procedure did not list the containment as an area to be inspected and the procedures did not require the placement and location of portable fire suppression equipment inside containment during the refueling outage. Due to a low number of work activities ongoing at the time, this issue was screened as Green (very low risk significance).
Inspection Report# : 2000007(pdf)
Significance:          May 22, 2000 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO INSTALL RAYCHEM HEAT SHRINK MATERIAL IN ACCORDANCE WITH PROCEDURE REQUIREMENTS.
The licensee identified that two Raychem electrical cable splices utilized in environmentally qualified (EQ) safety-related equipment had not been installed in accordance with EQ requirements. These splices were associated with pressurizer level transmitters and were installed in 1984. The licensee subsequently performed extensive EQ testing of the splices to determine the qualification of the splices' as-found configurations. Test results indicated that the splices would have been able to perform their intended function in a harsh environment inside containment. This issue was considered to be of low safety significance based on the successful EQ testing of the as-found splices' configurations and was screened as Green (very low risk significance). One non-cited violation was identified.
Inspection Report# : 2000007(pdf)
Significance:          Feb 21, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform Thorough 10 CFR 50.59 Safety Evaluation The inspectors identified a Non-Cited Violation for failure to perform an adequate 10 CFR 50.59 safety evaluation associated with emergency operating procedure changes to address component cooling water pump dead-head operational concerns. The safety evaluation did not evaluate the potential for initiating a loss-of-coolant accident via the reactor coolant loop seals during conditions of a complete loss of component cooling water.
Inspection Report# : 2001017(pdf)
Significance: N/A Sep 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ESTABLISH CONTINGENCY PLANS FOR ORANGE RISK CONDITION The inspectors identified the failure to establish contingency plans during a planned high risk plant configuration. Contrary to administrative requirements, the licensee approved an orange risk condition during a refueling outage with no contingency plans to mitigate the consequences of a loss of spent fuel pool cooling with a full core offload in the pool. A Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was identified. The finding was of very low safety significance because although the licensee had not approved appropriate contingency actions for the orange risk condition, the licensee subsequently rescheduled the planned maintenance to eliminate the orange risk condition.
Inspection Report# : 2001013(pdf)
Significance: N/A Aug 24, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO CONDUCT BIENNIAL SURVIELLANCE OF SAFETY-RELATED PROCEDURES PER TS 6.8.c A Non-Cited Violation of Technical Specification 6.8.c was identified for the failure to perform a biennial surveillance of safety-related procedues.
This issue was more than minor because if left uncorrected, could under the same condition become a more significant safety concern. However, since no specific cornerstone had been impacted, this finding is designated as No Color.
 
2Q/2000 Inspection Findings - Kewaunee                                                                                                  Page 3 of 11 Inspection Report# : 2001012(pdf)
Significance:          Aug 24, 2001 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE CORRECTIVE ACTIONS FOR PREVIOUS PROBLEM WITH THE CONTROL OF SAFETY-RELATED MATERIALS A Non-Cited Violation of Criterion XVI, "Corrective Action," of Appendix B of 10 CFR Part 50 was identified for ineffective corrective actions for a problem with the control of the storage of consumable materials, such as thread sealant, used in safety-related applications. These ineffective actions subsequently resulted in the inadequate control of the storage of grease used in safety-related breakers. This issue was more than minor because if left uncorrected, could under the same condition become a more significant safety concern. In that this issue could credibly affect the operability, availability, reliability, or function of a system or train in a mitigating system, it is a Green finding.
Inspection Report# : 2001012(pdf)
Significance:          Aug 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation NON-RATED FIRE BARRIER A Non-Cited Violation [of 10 CFR Part 50, Appendix R, Section III.G.2.a] was identified for failure to provide a 3-hour rated fire barrier to separate redundant trains of safe shutdown equipment. This finding was of very low safety significance because the licensee tested a replica of the fire barrier and demonstrated that the fire barrier provided protection for at least 60 minutes, which was sufficient for the hazards in the area.
Inspection Report# : 2001011(pdf)
Significance:          Jul 20, 2001 Identified By: NRC Item Type: FIN Finding LICENSED OPERATOR REQUALIFICATION EXAMINATION RESULTS The inspectors identified that two of eight crews examined during the licensee's calendar year 2001 licensed operator requalification operating test had failed. The finding was of very low safety significance because both crews that had failed received remedial training prior to being returned to shift, and the results of the licensee's operator licensing requalification operating test given in calendar year 2000 indicated that only one crew, out of a total of eight crews tested, had failed.
Inspection Report# : 2001011(pdf)
Significance: N/A Jun 30, 2001 Identified By: NRC Item Type: FIN Finding FAILURE TO BALANCE RELIABILITY AND AVAILABILITY AS REQUIRED BY 10 CFR 50.65(a)(3)
The inspectors identified a failure to evaluate whether adjustments were necessary such that there would be an appropriate balance between systems' availability and reliability in accordance with 10 CFR 50.65(a)(3) of the maintenance rule. The inspectors identified that the licensee did not have an administrative process to track maintenance rule functional failures and maintenance preventible maintenance functional failures. As a result, reliability and availability could not be balanced as required by the Maintenance Rule periodic evaluation. The safety significance of the specific finding was very low because it did not affect the operability of the systems, and the licensee entered the finding in the corrective action program. However, this finding was considered to be of regulatory concern in the area of maintenance rule implementation due to the extent of the problems with the Maintenance Rule Program.
Inspection Report# : 2001009(pdf)
Significance:          Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IDENTIFY DEFICIENT CONDITION OF VALVE AFW-1B The inspectors identified that the licensee failed to promptly identify and correct the B' train auxiliary feedwater pump discharge check valve which was stuck in an intermediate position. A Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was issued. The finding was of very low safety significance because, although the check valve was stuck in an intermediate position, the time that it was known to have been stuck was less than the technical specification allowed outage time for one train of auxiliary feedwater to be out of service (less than 72 hours). Additionally, the other two trains of auxiliary feedwater were each capable of 100 percent decay heat removal.
Inspection Report# : 2001009(pdf)
 
2Q/2000 Inspection Findings - Kewaunee                                                                                                  Page 4 of 11 Significance:        Jun 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO TRACK UNAVAILABILITY OF SYSTEMS REQUIRED DURING SHUTDOWN OPERATION 10 CFR 50.65(a)(1), required, in part, that the licensee monitor the performance or condition of SSCs within the scope of the rule as defined by 10 CFR 50.65(b), against licensee-established goals, in a manner sufficient to provide reasonable assurance that such SSCs are capable of fulfilling their intended functions. 10 CFR 50.65(a)(2) stated, in part, that monitoring as specified in 10 CFR 50.65(a)(1) was not required where it had been demonstrated that the performance or condition of an SSC was being effectively controlled through the performance of appropriate preventive maintenance, such that the SSC remained capable of performing its intended function. Contrary to the above, the licensee failed to demonstrate that the performance or condition of systems required to be available during shutdown conditions and within the scope of the rule had been effectively controlled through the performance of appropriate preventive maintenance and did not monitor against licensee-established goals.
Specifically, the licensee failed to monitor the unavailability of systems required during shutdown operation.
Inspection Report# : 2001009(pdf)
Significance:        Jun 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO ESTABLISH MAINTENANCE RULE (a)(1) GOALS 10 CFR 50.65(a)(1), requires, in part, that the licensee monitor the performance or condition of SSCs within the scope of the rule as defined by 10 CFR 50.65(b), against licensee-established goals, in a manner sufficient to provide reasonable assurance that such structures, systems, and components, are capable of fulfilling their intended functions. Such goals shall be established commensurate with safety. When the performance or condition of a structure, system, or component does not meet established goals, appropriate corrective action shall be taken. Contrary to the above, from 1996, the licensee did not take appropriate corrective actions when the performance of those systems in (a)(1) did not meet licensee established goals. Specifically, the licensee determined timely and appropriate corrective actions had not been taken for five systems that had been in (a)(1) category for approximately 3 years to 5 years: component cooling (entered (a)(1) on April 23, 1997), control room air conditioning (July 24, 1996), station and instrument air (July 3, 1997), auxiliary building air ventilation (July 31, 1997), and control rod drive (August 6, 1998). This issue is in the licensee's corrective action system as KAP WO 01-3323. The inspectors evaluated the risk significance of this issue using the Significance Determination Process. The inspectors did not identify where this failure resulted in a total loss of a risk significant SSC. Therefore, this issue was screened as Green (very low risk significance) after a Phase 1 Significance Determination Process review. Although the risk significance of this issue was low, the inspectors concluded that this was more than a minor concern because the failure to recognize and correct ineffective maintenance practices resulted in risk significant systems in (a)(1) for years with no improvement in performance. The NRC tracking number for this issue is 50-305/01-09-02.
Inspection Report# : 2001009(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation FIRE DOOR FUSIBLE LINKS.
On February 20, 2001, the licensee determined that the installed fusible link arrangement on roll-up fire Doors 279 and 281, which separated both trains of service water pumps, would not actuate as designed to ensure that the doors would automatically close to provide a 3-hour fire barrier, contrary to 10 CFR Part 50, Appendix R, Section III.G.2.a which required, in part, separation of cables and equipment of redundant trains by a fire barrier having a 3-hour rating.
Inspection Report# : 2001006(pdf)
Significance:        Mar 12, 2001 Identified By: Licensee Item Type: NCV NonCited Violation INADEQUATE SMOKE DETECTOR COVERAGE IN FIRE ZONE TU-95B.
Licensee identified violation of licensee's operating license that the licensee failed to install a detector in each beam pocket in Fire Zone TU-95B.
Inspection Report# : 2001002(pdf)
Significance:        Feb 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO TEST FIRE DOOR IN ACCORDANCE WITH FIRE PLAN.
 
2Q/2000 Inspection Findings - Kewaunee                                                                                                      Page 5 of 11 The inspectors identified a non-cited violation for failure to properly test a fire door in accordance with the facility's fire protection program plan. The finding was of very low safety significance because, although the fire door separated both trains of service water pumps and did not fully close as designed when subsequently tested, the fire loading in the area was insufficient to credibly impact more than two of the four service water pumps in the area.
Inspection Report# : 2001004(pdf)
Significance: N/A Nov 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO COMPLETE COMPONENT RE-TEST REQUIREMENTS PRIOR TO RETURNING EQUIPMENT TO AN OPERABLE STATUS.
No Color. The inspectors identified a Non-Cited Violation for failure to complete component retest requirements following maintenance performed on the B train control room air conditioner compressor condenser. The unit had been returned to an operable status prior to the retest requirements being completed as prescribed in the associated maintenance procedure. This issue was determined to be a violation of the licensee's Operational Quality Assurance Program Manual, Section 8, "Maintenance Planning and Control." Although the risk associated with this finding was very low and did not affect any cornerstones, the inspectors noted that this finding was similar to previous NRC-identified findings and therefore was of greater than minor significance and warranted documentation. (Section 1R19).
Inspection Report# : 2000020(pdf)
Significance:        Sep 29, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IDENTIFY CORROSION AS POTENTIAL FAILURE MECHANISM.
The inspectors identified that the licensee failed to identify corrosion as a potential failure mechanism in the operability determination for a carbon steel key in the service water system. Thus, the licensee failed to quantify the corrosion rate and therefore did not adequately evaluate the expected service life of the carbon steel key. One non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was identified.
Inspection Report# : 2000019(pdf)
Significance:        Sep 29, 2000 Identified By: NRC Item Type: NCV NonCited Violation PRESSURE RATING OF AUXILIARY FEEDWATER STEAM TRAPS.
The inspectors identified that a root cause evaluation for a 1996 equipment issue in the turbine-driven auxiliary feedwater system was not completed until 1999. The evaluation stated that the internals of the steam traps were designed to operate at pressures up to a maximum 600 psig but that the traps were exposed to pressures up to 1025 psig. A corrective action item to initiate a design change request to replace the steam traps with a different model rated for the design pressure of the system was described in the evaluation. However, the inspectors identified that the design change request had never been initiated and the KAP ( Kewaunee corrective action document) had been closed. As a result, the corrective action item for this design problem was lost. In addition, operability of the system had never been formally evaluated despite the identification that the system design requirements were not met. The licensee subsequently determined that the steam traps remained operable and was planning to initiate the design change to correct the problem. One non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was identified.
Inspection Report# : 2000019(pdf)
Significance: N/A Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation RETRIEVAL OF SERVICE WATER SYSTEM DESIGN INFORMATION.
In many cases, design basis information for the service water system was difficult if not impossible to locate. Licensee personnel wrote KAP WO 00-002566 to enter the problem in the corrective action program.
Inspection Report# : 2000012(pdf)
Significance:        Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation DESIGN CONTROL VIOLATION FOR AUXILIARY FEEDWATER STRAINER MESH SIZE.
The inspectors questioned the mesh size of the strainers, which were installed in the suction of the three auxiliary feedwater (AFW) pumps. As a result of the inspectors' questions, license personnel inspected the strainers on August 21, 2000, and found the strainers to have 1/16 inch openings. A note was later found on Figure 10.2-3 of the UFSAR that indicated that the AFW suction strainer size was 1/8 inch. The smaller openings would not support the use of service water as a safety related source for AFW and as a result all three trains of AFW were declared
 
2Q/2000 Inspection Findings - Kewaunee                                                                                                Page 6 of 11 inoperable. This condition had apparently existed for approximately 25 years and was identified as a non-cited violation of Criterion III, "Design Control," of 10 CFR 50, Appendix B.
Inspection Report# : 2000012(pdf)
Significance:          Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PROPERLY CONTROL CORRECTED TEST DATA NECESSARY FOR DESIGN CALCULATIONS.
An example of a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," was identified in the handling of service water system flow test data, which was subsequently used in calculations. Gauge readings corrected for post test calibration checks, gauge reading corrections for elevation considerations, and flow values corrected for pump degradation were contained in spreadsheets in the possession of an individual staff member, but not currently packaged with raw test data, and not bearing evidence of a formal review and control process. The connection between the test data, which had been vaulted, and the values used in the calculation, could not be made without use of the uncontrolled spreadsheet.
Inspection Report# : 2000012(pdf)
Significance:          Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation IMPROPER DESIGN CALCULATION IDENTIFICATION, NON-CONSERVATIVE ASSUMPTIONS, CALCULATION ERRORS, AND DUPLICATE CALCULATIONS.
An example of a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," was identified because of inadequate control of design calculations. The control failures included improper identification of calculations, non-conservative assumptions, calculation errors, and duplicate or superceded calculations not properly identified or canceled. The failure to follow the established design control process increased the potential for errors in the design and operation of the service water system. Because the system was subsequently demonstrated to be capable of removing the design heat load, the actual significance was low and this finding screened out as having very low risk significance.
Inspection Report# : 2000012(pdf)
Significance:          Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO DOCUMENT IN THE CORRECTIVE ACTION PROGRAM THE USE OF INCORRECT MATERIAL IN A SERVICE WATER PUMP KEY.
An example of a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was identified because of inadequate corrective action to correct an incorrect coupling adjust nut set screw and a low strength "soft" key material, which had contributed to a pump shaft failure. Licensee personnel had known of the "soft" key material since July 21, 1999. The "soft" key material was found in other service water pumps but had not been removed from all pumps. As of July 25, 2000, licensee personnel had not documented the existence of the "soft" key material in the corrective action program.
Inspection Report# : 2000012(pdf)
Significance:          Jul 07, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO COMPLETE COMPONENT RETEST REQUIREMENTS IN ACCORDANCE WITH PROCEDURE.
On June 26, 2000, during a review of post maintenance testing requirements following maintenance performed on the control room post accident system charcoal filter heat detector, the inspectors identified that maintenance technicians had not completed the component re-test requirements, as required by a preventative maintenance procedure prior to the system being returned to an operable status. On July 7, the inspectors identified a second example of failing to complete component re-test requirements following maintenance on the zone special ventilation system charcoal filter heat detector as required. The issue was considered to be of very low safety significance based on the determination that although the licensee had not completed all of the component retest requirements prior to returning the equipment to service, the subsequent testing determined that the equipment was in an operable status. The failure to complete the component retest requirements in accordance with site procedures was identified as a Non-Cited Violation.
Inspection Report# : 2000014(pdf)
 
2Q/2000 Inspection Findings - Kewaunee                                                                                                  Page 7 of 11 Barrier Integrity Significance:        Jun 22, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation FAILURE TO UPDATE COMPUTER ALARM FOR CURRENT AXIAL FLUX DISTRIBUTION TARGET BAND.
The licensee identified, following plant startup, that a computer alarm had not been updated properly to alarm if axial flux distribution deviated outside of the flux distribution target band. This condition was contrary to technical specification requirements. The licensee reviewed the axial flux distribution history since the startup and determined that at no time was the flux distribution outside of the target band. Since the axial flux distribution was never outside of the target band, this issue was screened as very low risk significance. An NCV was identified for failing to comply with technical specification requirements for monitoring axial flux distributions.
Inspection Report# : 2000008(pdf)
Significance:        May 22, 2000 Identified By: Licensee Item Type: FIN Finding TECHNICAL SPECIFICATIONS INTERPRETATION FOR TESTING REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES.
The NRC determined that the licensee's practice of testing reactor coolant system pressure isolation Valve SI-22B prior to entering the cold shutdown condition was contrary to Technical Specification requirements 4.2.a.3.a. Technical Specification 4.2.a.3.a required that periodic leakage testing of Valve SI-22B be accomplished prior to reaching operating mode after the plant was placed in cold shutdown. This issue was considered to be of low safety significance because of a subsequent successful valve test and was screened as Green (very low risk significance).
Enforcement discretion was applied to this item in accordance with Section VII.B.6 of the Enforcement Policy.
Inspection Report# : 2000007(pdf)
Significance:        May 19, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PROPERLY IMPLEMENT FLAW ACCEPTANCE CRITERIA FOR SLEEVE WELD INSPECTIONS.
During 1998 inservice inspection examinations, the licensee failed to properly implement the flaw acceptance criteria for laser welded sleeve inspection within two steam generator tubes. The safety significance was very low based on the absence of adverse consequences, and May 2000 in-situ pressure testing where both welds exhibited zero leakage at normal operating pressure, main steam line break pressure, and three times normal operating differential pressures. As such, this issue was characterized as a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion IX, Control of Special Processes. To correct the error, the licensee plugged both tubes (Section 1RO8).
Inspection Report# : 2000011(pdf)
Emergency Preparedness Significance: N/A Apr 05, 2000 Identified By: NRC Item Type: FIN Finding LICENSEE FAILED TO IDENTIFY THE FULL SCOPE OF PROBLEMS WITH THE ALERT AND NOTIFICATION SYSTEM PERFORMANCE.
The inspectors concluded that the licensee's assessment was not sufficiently comprehensive to identify the full scope of problems associated with the Alert and Notification System (ANS) performance program. As a result, licensee corrective actions generally were focused on the equipment problem rather than all root causes and contributing factors. For example, the inspectors identified that: (1) management oversight of the ANS performance program was limited; (2) an audit failed to identify degrading ANS performance as a concern; (3) annual preventive maintenance was not consistently performed on the system; (4) the corrective action program was not used consistently to document ANS problems; and (5) maintenance procedures and records were deficient. Collectively, these problems indicate that the ANS performance program lacked sufficient structure and oversight.
Inspection Report# : 2000006(pdf)
Significance:        Mar 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation
 
2Q/2000 Inspection Findings - Kewaunee                                                                                                    Page 8 of 11 SUPPLEMENTAL INSPECTION OF WHITE ERO AUGMENTATION FINDING AND RESULTING GREEN FINDING.
This supplemental inspection was performed by the NRC to evaluate the licensees evaluation associated with the failure to conduct successful quarterly, off-hours, unannounced staff augmentation drills during the second, third, and fourth quarters of 1999 and the second quarter of 2000.
This performance issue was previously characterized as having low to moderate risk significance (White) in NRC Inspection Report No. 50-305/2000015(DRS). During this supplemental inspection, performed in accordance with Inspection Procedure95002, the inspector concluded that the licensee performed a comprehensive evaluation of the unsuccessful staff augmentation drills. The licensees evaluation identifiedtwo root causes which resulted in the unsuccessful drills and in the staffs inability tocorrect thedeficiencies: (1) Management has not effectively acted to provide increased depth and flexibility in the emergency response organization following a reduction in staffing several years ago; and (2)
Management has accepted an adverse trend of test failures without requiring investigation into the root causes. The inspector reviewed the licensees corrective actions, both completed and planned, and concluded that the programmatic corrective actions appeared to address the identified root causes. In particular, the licensee assigned certain positions to an on-call rotation to ensure personnel were capable of augmenting in a timely manner, and the licensee was progressing in training additional staff to increase the depth of personnel assigned to key emergency response positions. In addition, the licensee was continuing its efforts in improving its corrective action program. The inspector reviewed the licensees immediate response to the issue and identified that one of the licensees initial corrective actions resulted in a Non-Cited Violation of regulatory requirements. To obtain a timely response of a key emergency response position (severe accident management - core hydraulics), the emergency preparedness staff effectively changed the emergency plan without revising the necessary procedures and without formally assessing the impact of that change. The staff instructed and trained personnel to respond to a location other than the Technical Support Center, which was contrary to the licensees current implementing procedures. While this change enabled the licensee to augment its staff in a timely manner, the change was not performed in accordance with NRC requirements. In order to make such a change, the licensees emergency plan required that the change be formally assessed to ensure that it did not reduce the effectiveness of the plan or any other implementing procedure. Since the issue did not result in a failure to meet an emergency preparedness planning standard, the failure to adequately implement the emergency plan was determined to be a violation of very low safety significance (Green) (Section 02.3(a)). Due to the licensees acceptable performance in assessing the emergency response augmentation drill deficiencies, the White finding associated with this issue will only be considered in assessing plant performance for a total of four quarters in accordance with the guidance in IMC 0305, Operating Reactor Assessment Program. Implementation of the licensees corrective actions will be reviewed during a future inspection.
Inspection Report# : 2001007(pdf)
Significance:          Mar 30, 2001 Identified By: NRC Item Type: VIO Violation FAILURE TO CORRECT SELF-IDENTIFIED ERO AUGMENTATION DRILL DEFICIENCIES.
During a baseline inspection of the emergency preparedness program conducted on August 14 - September 21, 2000, the NRC identified a preliminary White issue and potential violation for the licensees failure to successfully correct deficiencies identified during staff augmentation drills and to demonstrate timely staff augmentation in 1999 and 2000. The issue was unresolved pending the outcome of the NRCs final significance determination. On January 30, 2001, the NRC conducted a regulatory conference with the licensee and subsequently issued the licensee a White finding and Notice of Violation (Enforcement Action No. 00-214) associated with the performance issue.
Inspection Report# : 2001007(pdf)
Significance:          Mar 23, 2001 Identified By: NRC Item Type: NCV NonCited Violation POST-ACCIDENT SAMPLING SYSTEM CONTAINMENT AIR SAMPLE PANEL TESTING.
A Non-Cited Violation of Technical Specification 6.14 was identified for the failure to implement a program that ensured the capability to obtain and analyze containment atmosphere samples under accident samples using the cantainment air sampling panel (CASP). Although the CASP was installed, as was indicated in the emergency plan, the licensee had neither developed procedures nor had tested its capability to obtain a containment atmosphere sample using the CASP. The licensee could not recall if and when containment air samples were last obtained using the CASP.
Inspection Report# : 2001006(pdf)
Significance: N/A Feb 27, 2001 Identified By: NRC Item Type: FIN Finding SUPPLEMENTAL INSPECTION OF YELLOW ANS PERFORMANCE INDICATOR AND ASSOCIATED CORRECTIVE ACTION PROGRAM DEFICIENCIES.
This supplemental inspection was performed by the NRC to assess the licensee's evaluation associated with a Yellow performance indicator for the Alert and Notification System (ANS) and the associated Yellow finding related to the licensee's corrective action program. During this supplemental inspection, performed in accordance with Inspection Procedure 95002, the inspectors concluded that the licensee performed comprehensive evaluations of the performance problems associated with the ANS and its corrective action program. These evaluations identified primary root causes and contributing causes for both issues. Along with the electronics/hardware problems, the licensee identified the primary root causes for the ANS to be the failure to make changes to the system via a change control process and the failure of the activation procedure to provide for alternate/backup activation methods and to provide clear success criteria. In the case of the corrective action program, the licensee concluded that the primary root causes were plant management's inadequate risk evaluation regarding decisions affecting the corrective action program and the
 
2Q/2000 Inspection Findings - Kewaunee                                                                                                  Page 9 of 11 development of a culture in the licensee's organization, which minimized the importance of information from outside organizations. In particular, the licensee focused on low-cost power operation and failed to recognize the value of changes and improvements which had occurred throughout the industry, such as the value and expansion of the corrective action program. In the case of the Yellow ANS performance indicator, the licensee had completed several significant corrective actions to address the root causes and contributing causes identified in its evaluation. The inspectors found that the corrective actions appeared appropriate to address the underlying root causes and that ANS testing data indicated an improving trend in the NRC performance indicator. The licensee also performed comprehensive assessments of the emergency preparedness program, quality assurance program, plant operations, and other plant programs to determine the extent of condition (re. the root causes described above). Based on these evaluations, the licensee began to implement significant actions to correct the deficiencies in the corrective action program and other weaknesses identified. Generally, the inspectors observed progress in the licensee's initial implementation of these corrective actions. Due to the licensee's acceptable performance in assessing the Yellow ANS performance indicator and the associated Yellow finding, the Yellow finding will not be considered in assessing future plant performance.
Inspection Report# : 2001005(pdf)
Significance:        Sep 21, 2000 Identified By: NRC Item Type: FIN Finding INADEQUATE ROOT CAUSE EVALUATION FOR YELLOW ALERT AND NOTIFICATION SYSTEM PERFORMANCE INDICATOR.
The licensee's evaluation of the Yellow Alert and Notification (siren) System Performance Indicator (PI) was inadequate. The inspector concluded that the licensee's evaluation was not performed at the depth necessary to identify the root causes of the siren performance problems and, instead, only identified the symptoms of the root causes. Specifically, the inspector identified the following substantive weaknesses in the licensee's evaluation of the siren system performance, which appeared to result from systemic corrective action program deficiencies within this cornerstone:
* The licensee's evaluation was not of sufficient depth to clearly identify the root causes associated with the decline in siren system performance.
* Licensee management did not provide well-understood and clear guidance/expectations for performing root cause evaluations.
* The licensee's evaluation of the quality assurance program was narrowly focused and was not critical of its role in failing to identify and correct the siren performance problems.
* The licensee did not establish a priority for each of the long-term corrective actions in accordance with the associated significance or risk.
* The licensee did not have any formal provisions for measuring the effectiveness of its corrective actions.
* Within the licensee's evaluation, the licensee had not evaluated common causes or the extent of the condition. Due to the corrective action program performance deficiencies within this cornerstone, we have been unable to conclude that the performance issues that resulted in the yellow PI have been addressed. Therefore, we are issuing a yellow finding that corresponds to the original issues that resulted in a yellow PI. Additional inspection effort will be focused on the licensee's further evaluation of the siren reliability root causes and the continuing corrective action program implementation deficiencies identified during this inspection.
Inspection Report# : 2000017(pdf)
Occupational Radiation Safety Significance:        May 19, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO POST A VERY HIGH RADIATION AREA.
The inspectors identified a noncited violation for the failure to post a very high radiation area in accordance with 10 CFR 20.1902(c). Although the area was not adequately posted, the licensee had provided physical controls and barriers that were consistent with its requirements for a very high radiation area. Based on the adequacy of these controls, the potential for an overexposure from the inadvertent entry of personnel into the area was low. Consequently, this finding was determined to be of very low safety significance (Section 20S1.1).
Inspection Report# : 2000009(pdf)
Significance:        Oct 02, 2001 Identified By: NRC Item Type: NCV NonCited Violation HIGH RADIATION AREA ACCESS CONTROLS Non-Cited Violation of Technical Specification 6.13 and an associated Green Finding for failure to 'barricade' three ladders that provided entry to high radiation areas (less than 1000 mrem/hour) located on the steam generator/pressurizer platforms.
Inspection Report# : 2001014(pdf)
 
2Q/2000 Inspection Findings - Kewaunee                                                                                                Page 10 of 11 Public Radiation Safety Physical Protection Significance: N/A Jul 21, 2000 Identified By: NRC Item Type: FIN Finding CORRECTIVE ACTION ON SEARCH ISSUES DID NOT WORK.
The inspector determined that the licensee's effectiveness of implemented corrective actions for a previously identified inspection finding regarding an inadequate vehicle search was not totally effective in preventing recurrence. Previous corrective action was not adequately focused (Section 3PP2.2).
Inspection Report# : 2000013(pdf)
Significance:        Jul 21, 2000 Identified By: NRC Item Type: NCV NonCited Violation SEARCHES INADEQUATE (REPEAT).
The inspector identified a Non-Cited violation by observing that, a security officer failed to search an easily accessible compartment on one vehicle.
The failure resulted from human error because the officer did not observe the access panel to the compartment during the vehicle search process.
When searched, no prohibited items were found. Corrective actions were implemented. The inspector determined that the licensee's effectiveness of implemented corrective actions for a previously identified inspection finding regarding an inadequate vehicle search was not totally effective in preventing recurrence. Previous corrective action was not adequately focused.
Inspection Report# : 2000013(pdf)
Miscellaneous Significance: N/A Jun 22, 2000 Identified By: NRC Item Type: FIN Finding CONTROL ROOM OPERATIONS HUMAN PERFORMANCE ISSUES.
The inspectors interviewed operators to evaluate their awareness of degraded control room indications and alarms, and their ability to adequately take manual actions based on degraded alarm functions. The inspectors identified, during interviews, that there was a lack of awareness by operators of a degraded refueling water storage tank low-low level alarm which would be potentially confusing to operators and therefore increase the risk associated with initiating long term sump recirculation.
Inspection Report# : 2000008(pdf)
Significance: N/A Feb 21, 2002 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Implement Required Fire Watch Following Completion of Hot Work Activities A licensee-identified violation was reviewed by the inspectors. Corrective actions taken or planned by the licensee appeared reasonable.
Inspection Report# : 2001017(pdf)
Significance: N/A Aug 24, 2001 Identified By: NRC Item Type: FIN Finding IDENTIFICATION AND RESOLUTION OF PROBLEMS The team concluded that the licensee was generally effective at identifying problems and putting them into the corrective action program. The program itself contained all the necessary attributes of an acceptable corrective action program and was generally successful in correcting identified issues. However, the team noted that, although licensee management had taken efforts to ensure that issues were resolved in accordance with program guidance and requirements, additional efforts appeared necessary to ensure timely resolution of issues. A positive program initiative was the establishment of positions in each of the major plant departments to serve as liaisons between the departments and the corrective action program and to assist with self-assessments. However, examples were identified by the inspectors of problems with the licensee's identification and resolution of problems, prioritization and evaluation of issues, and the effectiveness of corrective actions. Included in these examples were the routine granting of due date extensions for problem evaluation and corrective action implementation, failure to perform a Technical Specification-required biennial surveillance of safety-related procedures, and ineffective corrective actions that resulted in the lack of
 
2Q/2000 Inspection Findings - Kewaunee                                                                                                Page 11 of 11 proper controls over the storage of grease used in safety-related breakers. Based on a review of records and discussions with plant staff, the inspectors concluded that workers at the site felt free to input safety issues into the corrective action program.
Inspection Report# : 2001012(pdf)
Significance: N/A Nov 09, 2000 Identified By: NRC Item Type: FIN Finding FAILURE TO TAKE ADEQUATE CORRECTIVE ACTIONS TO ENSURE COMPONENT RE-TEST REQUIREMENTS COMPLETED ACCORDING TO MAINTENANCE PROCEDURES.
No Color. The inspectors determined that a negative performance trend had developed in the licensee's ability to identify and promptly take appropriate corrective actions to prevent recurrence based on two previously identified examples (NCV 50-305/2000014-01) and one example identified during this inspection period (NCV 50-305/00-20-01). All three examples related to the licensee returning safety-related equipment to service prior to completing all required post-maintenance retesting. While the risk of the individual examples was very low, the licensee had failed to ensure that all retest requirements had been completed before returning safety-related equipment to service. These findings collectively indicated a problem with the licensee's ability to provide timely and adequate corrective actions to prevent recurrence. (Section 4OA2).
Inspection Report# : 2000020(pdf)
Significance: N/A Sep 29, 2000 Identified By: NRC Item Type: FIN Finding EFFECTIVENESS OF CORRECTIVE ACTION PROGRAM.
Based on the results of this inspection, the NRC concluded that the corrective action program at Kewaunee showed significant weaknesses and inconsistencies across all of the procedural elements inspected. These weaknesses existed across departments and affected multiple cornerstones in the strategic performance areas of Reactor Safety, Radiation Safety, and Safeguards. Of particular note was the lack of procedures for determining the significance of conditions adverse to quality and for trending of issues and the complete lack of trending within your corrective action program. We also identified a lack of urgency in correcting issues which resulted in repeat examples occurring and, coupled with a poor tracking system, a tendency for issues to be dropped. While none of the specific examples identified by the team were of high risk significance when looked at in isolation, in the aggregate they were similar in nature to prior issues in the emergency preparedness area that rose to a higher significance level and contributed to a degraded cornerstone. While we concluded that the station had fostered an environment in which personnel freely identified conditions adverse to quality without fear of discrimination or retaliation, we also concluded that significant weaknesses with, and inconsistent implementation of, the station corrective action program resulted in multiple examples where station personnel did not enter deficiencies into the station's formal corrective action program.
Inspection Report# : 2000019(pdf)
Significance: N/A Sep 29, 2000 Identified By: NRC Item Type: NCV NonCited Violation NO PROCEDURAL GUIDANCE FOR DETERMINING IF CONDITIONS ADVERSE TO QUALITY ARE SIGNIFICANT.
The inspectors reviewed the quality assurance (QA) manual requirements against Kewaunee's implementing procedures and identified that two QA manual requirements were not being implemented. Specifically, Program Requirement 3.1.9 which stated that directives and procedures shall provide for the review of conditions adverse to quality to determine if the conditions are significant in nature. This requirement paralleled 10 CFR Part 50, Appendix B, Criterion XVI, which requires that the cause of significant conditions adverse to quality be determined and corrective actions taken to prevent recurrence. The inspectors reviewed the Nuclear Administrative Directive (NAD 11.08) and the procedure (GNP 11.08.01) governing the KAP (Kewaunee corrective action program) process and found no procedure requirements for identifying significant conditions adverse to quality. This finding does not directly affect a cornerstone. As a result, this issue was not evaluated with the Significance Determination Process and was not assigned a color. One example of a non-cited violation of 10 CFR Part 50, Appendix B, Criterion II, "Quality Assurance Programs," was identified.
Inspection Report# : 2000019(pdf)
Significance: N/A Sep 29, 2000 Identified By: NRC Item Type: NCV NonCited Violation NO PROCEDURAL GUIDANCE FOR TRENDING CONDITIONS ADVERSE TO QUALITY.
The inspectors reviewed the quality assurance (QA) manual requirements against Kewaunee's implementing procedures and identified that two QA manual requirements were not being implemented. Specifically, QA Program Requirement, 3.1.10, stated that directives and procedures shall provide for analyzing trends of conditions adverse to quality. Once identified these trends were required to be considered significant conditions adverse to quality. The inspectors found that conditions adverse to quality were not defined in the KAP (Kewaunee corrective action program) procedures and that no procedure existed for trending. This finding does not directly affect a cornerstone. As a result, this issue was not evaluated with the Significance Determination Process and was not assigned a color. One example of a non-cited violation of 10 CFR Part 50, Appendix B, Criterion II, "Quality Assurance Program," was identified.
Inspection Report# : 2000019(pdf)
Last modified : April 01, 2002
 
3Q/2000 Inspection Findings - Kewaunee                                                                                                  Page 1 of 11 Kewaunee Initiating Events Mitigating Systems Significance:          Sep 29, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IDENTIFY CORROSION AS POTENTIAL FAILURE MECHANISM.
The inspectors identified that the licensee failed to identify corrosion as a potential failure mechanism in the operability determination for a carbon steel key in the service water system. Thus, the licensee failed to quantify the corrosion rate and therefore did not adequately evaluate the expected service life of the carbon steel key. One non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was identified.
Inspection Report# : 2000019(pdf)
Significance:          Sep 29, 2000 Identified By: NRC Item Type: NCV NonCited Violation PRESSURE RATING OF AUXILIARY FEEDWATER STEAM TRAPS.
The inspectors identified that a root cause evaluation for a 1996 equipment issue in the turbine-driven auxiliary feedwater system was not completed until 1999. The evaluation stated that the internals of the steam traps were designed to operate at pressures up to a maximum 600 psig but that the traps were exposed to pressures up to 1025 psig. A corrective action item to initiate a design change request to replace the steam traps with a different model rated for the design pressure of the system was described in the evaluation. However, the inspectors identified that the design change request had never been initiated and the KAP ( Kewaunee corrective action document) had been closed. As a result, the corrective action item for this design problem was lost. In addition, operability of the system had never been formally evaluated despite the identification that the system design requirements were not met. The licensee subsequently determined that the steam traps remained operable and was planning to initiate the design change to correct the problem. One non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was identified.
Inspection Report# : 2000019(pdf)
Significance:          Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation IMPROPER DESIGN CALCULATION IDENTIFICATION, NON-CONSERVATIVE ASSUMPTIONS, CALCULATION ERRORS, AND DUPLICATE CALCULATIONS.
An example of a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," was identified because of inadequate control of design calculations. The control failures included improper identification of calculations, non-conservative assumptions, calculation errors, and duplicate or superceded calculations not properly identified or canceled. The failure to follow the established design control process increased the potential for errors in the design and operation of the service water system. Because the system was subsequently demonstrated to be capable of removing the design heat load, the actual significance was low and this finding screened out as having very low risk significance.
Inspection Report# : 2000012(pdf)
Significance:          Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation DESIGN CONTROL VIOLATION FOR AUXILIARY FEEDWATER STRAINER MESH SIZE.
The inspectors questioned the mesh size of the strainers, which were installed in the suction of the three auxiliary feedwater (AFW) pumps. As a result of the inspectors' questions, license personnel inspected the strainers on August 21, 2000, and found the strainers to have 1/16 inch openings. A note was later found on Figure 10.2-3 of the UFSAR that indicated that the AFW suction strainer size was 1/8 inch. The smaller
 
3Q/2000 Inspection Findings - Kewaunee                                                                                                  Page 2 of 11 openings would not support the use of service water as a safety related source for AFW and as a result all three trains of AFW were declared inoperable. This condition had apparently existed for approximately 25 years and was identified as a non-cited violation of Criterion III, "Design Control," of 10 CFR 50, Appendix B.
Inspection Report# : 2000012(pdf)
Significance: N/A Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation RETRIEVAL OF SERVICE WATER SYSTEM DESIGN INFORMATION.
In many cases, design basis information for the service water system was difficult if not impossible to locate. Licensee personnel wrote KAP WO 00-002566 to enter the problem in the corrective action program.
Inspection Report# : 2000012(pdf)
Significance:        Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PROPERLY CONTROL CORRECTED TEST DATA NECESSARY FOR DESIGN CALCULATIONS.
An example of a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," was identified in the handling of service water system flow test data, which was subsequently used in calculations. Gauge readings corrected for post test calibration checks, gauge reading corrections for elevation considerations, and flow values corrected for pump degradation were contained in spreadsheets in the possession of an individual staff member, but not currently packaged with raw test data, and not bearing evidence of a formal review and control process. The connection between the test data, which had been vaulted, and the values used in the calculation, could not be made without use of the uncontrolled spreadsheet.
Inspection Report# : 2000012(pdf)
Significance:        Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO DOCUMENT IN THE CORRECTIVE ACTION PROGRAM THE USE OF INCORRECT MATERIAL IN A SERVICE WATER PUMP KEY.
An example of a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was identified because of inadequate corrective action to correct an incorrect coupling adjust nut set screw and a low strength "soft" key material, which had contributed to a pump shaft failure. Licensee personnel had known of the "soft" key material since July 21, 1999. The "soft" key material was found in other service water pumps but had not been removed from all pumps. As of July 25, 2000, licensee personnel had not documented the existence of the "soft" key material in the corrective action program.
Inspection Report# : 2000012(pdf)
Significance:        Jul 07, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO COMPLETE COMPONENT RETEST REQUIREMENTS IN ACCORDANCE WITH PROCEDURE.
On June 26, 2000, during a review of post maintenance testing requirements following maintenance performed on the control room post accident system charcoal filter heat detector, the inspectors identified that maintenance technicians had not completed the component re-test requirements, as required by a preventative maintenance procedure prior to the system being returned to an operable status. On July 7, the inspectors identified a second example of failing to complete component re-test requirements following maintenance on the zone special ventilation system charcoal filter heat detector as required. The issue was considered to be of very low safety significance based on the determination that although the licensee had not completed all of the component retest requirements prior to returning the equipment to service, the subsequent testing determined that the equipment was in an operable status. The failure to complete the component retest requirements in accordance with site procedures was identified as a Non-Cited Violation.
Inspection Report# : 2000014(pdf)
Significance:        Jun 22, 2000 Identified By: NRC Item Type: FIN Finding FAILURE TO MEET SINGLE FAILURE CRITERIA FOR RESIDUAL HEAT REMOVAL VALVE CIRCUITRY.
The licensee identified that the circuitry associated with the residual heat removal system discharge to safety injection system suction isolation valves did not meet single failure criteria. The inspectors noted that this design requirement was identified in the facility's updated safety analysis
 
3Q/2000 Inspection Findings - Kewaunee                                                                                                    Page 3 of 11 report. The licensee subsequently implemented a temporary change to the facility. The inspectors reviewed the issue and identified that the facility had been operating outside of its design basis, which was reportable to the NRC. The licensee subsequently made a one hour non-emergency report to the NRC. Since there was no actual loss of safety function to the system, this issue was screened as very low risk significance.
Inspection Report# : 2000008(pdf)
Significance:        Jun 22, 2000 Identified By: NRC Item Type: FIN Finding MAINTENANCE RULE FAILURES ASSOCIATED WITH REACTOR HEAD VENT VALVE.
The inspectors reviewed the licensee's implementation of the maintenance rule for failures associated with a reactor head vent valve. The licensee's corrective action documents identified a potential maintenance rule functional failure but the completed evaluation of the problem did not document the final determination. However, the inspectors identified that the repeated failures may have been prevented if maintenance activities such as valve disassembly and cleaning had been performed. In this case, maintenance rule reliability goals were not exceeded. The licensee had documented similar maintenance rule program deficiencies and developed a corrective action program to address the deficiencies. Although programmatic deficiencies exist, since no maintenance rule reliability criteria had been exceeded, this issue was considered of very low risk significance.
Inspection Report# : 2000008(pdf)
Significance:        Jun 22, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO INITIATE KEWAUNEE ASSESSMENT PROCESS DOCUMENT REGARDING REFUELING WATER STORAGE TANK LOW-LOW LEVEL ALARM INOPERABILITY.
The inspectors identified that the refueling water storage tank low-low level alarm which was actuating five percent higher than normal had not been documented in a Kewaunee Assessment Process form by the licensee, and therefore had not received an operability evaluation. This failure was identified as contrary to site administrative procedures. Following the licensee's documentation of the problem, the inspectors identified that the associated operability evaluation considered the acceptability of an operator workaround to address the issue, but did not address any safety implications or consequences of the alarm actuating early. A subsequent operability evaluation by the licensee was evaluated as adequate by the inspectors. Since the subsequent operability evaluation was adequate and it was determined that no safety mitigation equipment was adversely affected by the early actuation of the alarm, this issue was considered of very low risk significance. A non-cited violation (NCV) was identified for failing to document a non-conforming condition, contrary to site administrative procedure requirements.
Inspection Report# : 2000008(pdf)
Significance:        Jun 22, 2000 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO TEST ADDITIONAL RELIEF VALVES IN ACCORDANCE WITH TECHINCAL SPECIFICATIONS.
The licensee identified that the suction relief valve for an auxiliary feedwater pump may have failed its relief test criteria, but did not process the documented deficiency until several weeks later. The licensee then expanded the scope of the relief testing to the suction relief valves associated with the other auxiliary feedwater pumps to meet technical specification requirements. Since any one train of auxiliary feedwater was capable of supplying 100 percent of the decay heat removal requirements, this issue was screened as very low risk significance. However, the time delay in complying with technical specification requirements for testing other relief valves was identified as an NCV.
Inspection Report# : 2000008(pdf)
Significance:        May 22, 2000 Identified By: NRC Item Type: FIN Finding FIRE EXTINGUISHERS NOT LOCATED IN ALL AREAS OF CONTAINMENT BUILDING AT BEGINNING OF REFUELING OUTAGE.
During a walkdown of the containment building, the inspectors identified that portable fire extinguishers were not located in the containment basement at the beginning of the plant refueling outage. Additionally, site fire protection procedures required that responsible fire protection personnel perform inspections of selected plant areas to ensure that the quantity of combustible material was minimized. However, the procedure did not list the containment as an area to be inspected and the procedures did not require the placement and location of portable fire suppression equipment inside containment during the refueling outage. Due to a low number of work activities ongoing at the time, this issue was screened as Green (very low risk significance).
Inspection Report# : 2000007(pdf)
 
3Q/2000 Inspection Findings - Kewaunee                                                                                                  Page 4 of 11 Significance:          May 22, 2000 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO INSTALL RAYCHEM HEAT SHRINK MATERIAL IN ACCORDANCE WITH PROCEDURE REQUIREMENTS.
The licensee identified that two Raychem electrical cable splices utilized in environmentally qualified (EQ) safety-related equipment had not been installed in accordance with EQ requirements. These splices were associated with pressurizer level transmitters and were installed in 1984. The licensee subsequently performed extensive EQ testing of the splices to determine the qualification of the splices' as-found configurations. Test results indicated that the splices would have been able to perform their intended function in a harsh environment inside containment. This issue was considered to be of low safety significance based on the successful EQ testing of the as-found splices' configurations and was screened as Green (very low risk significance). One non-cited violation was identified.
Inspection Report# : 2000007(pdf)
Significance:          Feb 21, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform Thorough 10 CFR 50.59 Safety Evaluation The inspectors identified a Non-Cited Violation for failure to perform an adequate 10 CFR 50.59 safety evaluation associated with emergency operating procedure changes to address component cooling water pump dead-head operational concerns. The safety evaluation did not evaluate the potential for initiating a loss-of-coolant accident via the reactor coolant loop seals during conditions of a complete loss of component cooling water.
Inspection Report# : 2001017(pdf)
Significance: N/A Sep 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ESTABLISH CONTINGENCY PLANS FOR ORANGE RISK CONDITION The inspectors identified the failure to establish contingency plans during a planned high risk plant configuration. Contrary to administrative requirements, the licensee approved an orange risk condition during a refueling outage with no contingency plans to mitigate the consequences of a loss of spent fuel pool cooling with a full core offload in the pool. A Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was identified. The finding was of very low safety significance because although the licensee had not approved appropriate contingency actions for the orange risk condition, the licensee subsequently rescheduled the planned maintenance to eliminate the orange risk condition.
Inspection Report# : 2001013(pdf)
Significance: N/A Aug 24, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO CONDUCT BIENNIAL SURVIELLANCE OF SAFETY-RELATED PROCEDURES PER TS 6.8.c A Non-Cited Violation of Technical Specification 6.8.c was identified for the failure to perform a biennial surveillance of safety-related procedues.
This issue was more than minor because if left uncorrected, could under the same condition become a more significant safety concern. However, since no specific cornerstone had been impacted, this finding is designated as No Color.
Inspection Report# : 2001012(pdf)
Significance:          Aug 24, 2001 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE CORRECTIVE ACTIONS FOR PREVIOUS PROBLEM WITH THE CONTROL OF SAFETY-RELATED MATERIALS A Non-Cited Violation of Criterion XVI, "Corrective Action," of Appendix B of 10 CFR Part 50 was identified for ineffective corrective actions for a problem with the control of the storage of consumable materials, such as thread sealant, used in safety-related applications. These ineffective actions subsequently resulted in the inadequate control of the storage of grease used in safety-related breakers. This issue was more than minor because if left uncorrected, could under the same condition become a more significant safety concern. In that this issue could credibly affect the operability, availability, reliability, or function of a system or train in a mitigating system, it is a Green finding.
Inspection Report# : 2001012(pdf)
Significance:          Aug 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation
 
3Q/2000 Inspection Findings - Kewaunee                                                                                                  Page 5 of 11 NON-RATED FIRE BARRIER A Non-Cited Violation [of 10 CFR Part 50, Appendix R, Section III.G.2.a] was identified for failure to provide a 3-hour rated fire barrier to separate redundant trains of safe shutdown equipment. This finding was of very low safety significance because the licensee tested a replica of the fire barrier and demonstrated that the fire barrier provided protection for at least 60 minutes, which was sufficient for the hazards in the area.
Inspection Report# : 2001011(pdf)
Significance:          Jul 20, 2001 Identified By: NRC Item Type: FIN Finding LICENSED OPERATOR REQUALIFICATION EXAMINATION RESULTS The inspectors identified that two of eight crews examined during the licensee's calendar year 2001 licensed operator requalification operating test had failed. The finding was of very low safety significance because both crews that had failed received remedial training prior to being returned to shift, and the results of the licensee's operator licensing requalification operating test given in calendar year 2000 indicated that only one crew, out of a total of eight crews tested, had failed.
Inspection Report# : 2001011(pdf)
Significance: N/A Jun 30, 2001 Identified By: NRC Item Type: FIN Finding FAILURE TO BALANCE RELIABILITY AND AVAILABILITY AS REQUIRED BY 10 CFR 50.65(a)(3)
The inspectors identified a failure to evaluate whether adjustments were necessary such that there would be an appropriate balance between systems' availability and reliability in accordance with 10 CFR 50.65(a)(3) of the maintenance rule. The inspectors identified that the licensee did not have an administrative process to track maintenance rule functional failures and maintenance preventible maintenance functional failures. As a result, reliability and availability could not be balanced as required by the Maintenance Rule periodic evaluation. The safety significance of the specific finding was very low because it did not affect the operability of the systems, and the licensee entered the finding in the corrective action program. However, this finding was considered to be of regulatory concern in the area of maintenance rule implementation due to the extent of the problems with the Maintenance Rule Program.
Inspection Report# : 2001009(pdf)
Significance:          Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IDENTIFY DEFICIENT CONDITION OF VALVE AFW-1B The inspectors identified that the licensee failed to promptly identify and correct the B' train auxiliary feedwater pump discharge check valve which was stuck in an intermediate position. A Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was issued. The finding was of very low safety significance because, although the check valve was stuck in an intermediate position, the time that it was known to have been stuck was less than the technical specification allowed outage time for one train of auxiliary feedwater to be out of service (less than 72 hours). Additionally, the other two trains of auxiliary feedwater were each capable of 100 percent decay heat removal.
Inspection Report# : 2001009(pdf)
Significance:          Jun 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO TRACK UNAVAILABILITY OF SYSTEMS REQUIRED DURING SHUTDOWN OPERATION 10 CFR 50.65(a)(1), required, in part, that the licensee monitor the performance or condition of SSCs within the scope of the rule as defined by 10 CFR 50.65(b), against licensee-established goals, in a manner sufficient to provide reasonable assurance that such SSCs are capable of fulfilling their intended functions. 10 CFR 50.65(a)(2) stated, in part, that monitoring as specified in 10 CFR 50.65(a)(1) was not required where it had been demonstrated that the performance or condition of an SSC was being effectively controlled through the performance of appropriate preventive maintenance, such that the SSC remained capable of performing its intended function. Contrary to the above, the licensee failed to demonstrate that the performance or condition of systems required to be available during shutdown conditions and within the scope of the rule had been effectively controlled through the performance of appropriate preventive maintenance and did not monitor against licensee-established goals.
Specifically, the licensee failed to monitor the unavailability of systems required during shutdown operation.
Inspection Report# : 2001009(pdf)
Significance:          Jun 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation
 
3Q/2000 Inspection Findings - Kewaunee                                                                                                      Page 6 of 11 FAILURE TO ESTABLISH MAINTENANCE RULE (a)(1) GOALS 10 CFR 50.65(a)(1), requires, in part, that the licensee monitor the performance or condition of SSCs within the scope of the rule as defined by 10 CFR 50.65(b), against licensee-established goals, in a manner sufficient to provide reasonable assurance that such structures, systems, and components, are capable of fulfilling their intended functions. Such goals shall be established commensurate with safety. When the performance or condition of a structure, system, or component does not meet established goals, appropriate corrective action shall be taken. Contrary to the above, from 1996, the licensee did not take appropriate corrective actions when the performance of those systems in (a)(1) did not meet licensee established goals. Specifically, the licensee determined timely and appropriate corrective actions had not been taken for five systems that had been in (a)(1) category for approximately 3 years to 5 years: component cooling (entered (a)(1) on April 23, 1997), control room air conditioning (July 24, 1996), station and instrument air (July 3, 1997), auxiliary building air ventilation (July 31, 1997), and control rod drive (August 6, 1998). This issue is in the licensee's corrective action system as KAP WO 01-3323. The inspectors evaluated the risk significance of this issue using the Significance Determination Process. The inspectors did not identify where this failure resulted in a total loss of a risk significant SSC. Therefore, this issue was screened as Green (very low risk significance) after a Phase 1 Significance Determination Process review. Although the risk significance of this issue was low, the inspectors concluded that this was more than a minor concern because the failure to recognize and correct ineffective maintenance practices resulted in risk significant systems in (a)(1) for years with no improvement in performance. The NRC tracking number for this issue is 50-305/01-09-02.
Inspection Report# : 2001009(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation FIRE DOOR FUSIBLE LINKS.
On February 20, 2001, the licensee determined that the installed fusible link arrangement on roll-up fire Doors 279 and 281, which separated both trains of service water pumps, would not actuate as designed to ensure that the doors would automatically close to provide a 3-hour fire barrier, contrary to 10 CFR Part 50, Appendix R, Section III.G.2.a which required, in part, separation of cables and equipment of redundant trains by a fire barrier having a 3-hour rating.
Inspection Report# : 2001006(pdf)
Significance:        Mar 12, 2001 Identified By: Licensee Item Type: NCV NonCited Violation INADEQUATE SMOKE DETECTOR COVERAGE IN FIRE ZONE TU-95B.
Licensee identified violation of licensee's operating license that the licensee failed to install a detector in each beam pocket in Fire Zone TU-95B.
Inspection Report# : 2001002(pdf)
Significance:        Feb 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO TEST FIRE DOOR IN ACCORDANCE WITH FIRE PLAN.
The inspectors identified a non-cited violation for failure to properly test a fire door in accordance with the facility's fire protection program plan. The finding was of very low safety significance because, although the fire door separated both trains of service water pumps and did not fully close as designed when subsequently tested, the fire loading in the area was insufficient to credibly impact more than two of the four service water pumps in the area.
Inspection Report# : 2001004(pdf)
Significance: N/A Nov 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO COMPLETE COMPONENT RE-TEST REQUIREMENTS PRIOR TO RETURNING EQUIPMENT TO AN OPERABLE STATUS.
No Color. The inspectors identified a Non-Cited Violation for failure to complete component retest requirements following maintenance performed on the B train control room air conditioner compressor condenser. The unit had been returned to an operable status prior to the retest requirements being completed as prescribed in the associated maintenance procedure. This issue was determined to be a violation of the licensee's Operational Quality Assurance Program Manual, Section 8, "Maintenance Planning and Control." Although the risk associated with this finding was very low and did not affect any cornerstones, the inspectors noted that this finding was similar to previous NRC-identified findings and therefore was of greater than minor significance and warranted documentation. (Section 1R19).
Inspection Report# : 2000020(pdf)
 
3Q/2000 Inspection Findings - Kewaunee                                                                                                  Page 7 of 11 Barrier Integrity Significance:        Jun 22, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation FAILURE TO UPDATE COMPUTER ALARM FOR CURRENT AXIAL FLUX DISTRIBUTION TARGET BAND.
The licensee identified, following plant startup, that a computer alarm had not been updated properly to alarm if axial flux distribution deviated outside of the flux distribution target band. This condition was contrary to technical specification requirements. The licensee reviewed the axial flux distribution history since the startup and determined that at no time was the flux distribution outside of the target band. Since the axial flux distribution was never outside of the target band, this issue was screened as very low risk significance. An NCV was identified for failing to comply with technical specification requirements for monitoring axial flux distributions.
Inspection Report# : 2000008(pdf)
Significance:        May 22, 2000 Identified By: Licensee Item Type: FIN Finding TECHNICAL SPECIFICATIONS INTERPRETATION FOR TESTING REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES.
The NRC determined that the licensee's practice of testing reactor coolant system pressure isolation Valve SI-22B prior to entering the cold shutdown condition was contrary to Technical Specification requirements 4.2.a.3.a. Technical Specification 4.2.a.3.a required that periodic leakage testing of Valve SI-22B be accomplished prior to reaching operating mode after the plant was placed in cold shutdown. This issue was considered to be of low safety significance because of a subsequent successful valve test and was screened as Green (very low risk significance).
Enforcement discretion was applied to this item in accordance with Section VII.B.6 of the Enforcement Policy.
Inspection Report# : 2000007(pdf)
Significance:        May 19, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PROPERLY IMPLEMENT FLAW ACCEPTANCE CRITERIA FOR SLEEVE WELD INSPECTIONS.
During 1998 inservice inspection examinations, the licensee failed to properly implement the flaw acceptance criteria for laser welded sleeve inspection within two steam generator tubes. The safety significance was very low based on the absence of adverse consequences, and May 2000 in-situ pressure testing where both welds exhibited zero leakage at normal operating pressure, main steam line break pressure, and three times normal operating differential pressures. As such, this issue was characterized as a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion IX, Control of Special Processes. To correct the error, the licensee plugged both tubes (Section 1RO8).
Inspection Report# : 2000011(pdf)
Emergency Preparedness Significance:        Sep 21, 2000 Identified By: NRC Item Type: FIN Finding INADEQUATE ROOT CAUSE EVALUATION FOR YELLOW ALERT AND NOTIFICATION SYSTEM PERFORMANCE INDICATOR.
The licensee's evaluation of the Yellow Alert and Notification (siren) System Performance Indicator (PI) was inadequate. The inspector concluded that the licensee's evaluation was not performed at the depth necessary to identify the root causes of the siren performance problems and, instead, only identified the symptoms of the root causes. Specifically, the inspector identified the following substantive weaknesses in the licensee's evaluation of the siren system performance, which appeared to result from systemic corrective action program deficiencies within this cornerstone:
* The licensee's evaluation was not of sufficient depth to clearly identify the root causes associated with the decline in siren system performance.
* Licensee management did not provide well-understood and clear guidance/expectations for performing root cause evaluations.
* The licensee's evaluation of the quality assurance program was narrowly focused and was not critical of its role in failing to identify and correct the siren performance problems.
* The licensee did not establish a priority for each of the long-term corrective actions in accordance with the associated significance or risk.
* The licensee did not have any formal provisions for measuring the effectiveness of its corrective actions.
* Within the licensee's evaluation, the licensee had not evaluated common causes or the extent of the condition. Due to the corrective action program performance deficiencies within this cornerstone, we have been unable to conclude that the performance issues that resulted in the yellow PI have been addressed. Therefore, we are issuing a yellow finding that corresponds to the original issues that resulted in a yellow PI. Additional inspection effort will be focused on the licensee's further evaluation of the siren reliability root causes and the continuing corrective action program
 
3Q/2000 Inspection Findings - Kewaunee                                                                                                    Page 8 of 11 implementation deficiencies identified during this inspection.
Inspection Report# : 2000017(pdf)
Significance: N/A Apr 05, 2000 Identified By: NRC Item Type: FIN Finding LICENSEE FAILED TO IDENTIFY THE FULL SCOPE OF PROBLEMS WITH THE ALERT AND NOTIFICATION SYSTEM PERFORMANCE.
The inspectors concluded that the licensee's assessment was not sufficiently comprehensive to identify the full scope of problems associated with the Alert and Notification System (ANS) performance program. As a result, licensee corrective actions generally were focused on the equipment problem rather than all root causes and contributing factors. For example, the inspectors identified that: (1) management oversight of the ANS performance program was limited; (2) an audit failed to identify degrading ANS performance as a concern; (3) annual preventive maintenance was not consistently performed on the system; (4) the corrective action program was not used consistently to document ANS problems; and (5) maintenance procedures and records were deficient. Collectively, these problems indicate that the ANS performance program lacked sufficient structure and oversight.
Inspection Report# : 2000006(pdf)
Significance:          Mar 30, 2001 Identified By: NRC Item Type: VIO Violation FAILURE TO CORRECT SELF-IDENTIFIED ERO AUGMENTATION DRILL DEFICIENCIES.
During a baseline inspection of the emergency preparedness program conducted on August 14 - September 21, 2000, the NRC identified a preliminary White issue and potential violation for the licensees failure to successfully correct deficiencies identified during staff augmentation drills and to demonstrate timely staff augmentation in 1999 and 2000. The issue was unresolved pending the outcome of the NRCs final significance determination. On January 30, 2001, the NRC conducted a regulatory conference with the licensee and subsequently issued the licensee a White finding and Notice of Violation (Enforcement Action No. 00-214) associated with the performance issue.
Inspection Report# : 2001007(pdf)
Significance:          Mar 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation SUPPLEMENTAL INSPECTION OF WHITE ERO AUGMENTATION FINDING AND RESULTING GREEN FINDING.
This supplemental inspection was performed by the NRC to evaluate the licensees evaluation associated with the failure to conduct successful quarterly, off-hours, unannounced staff augmentation drills during the second, third, and fourth quarters of 1999 and the second quarter of 2000.
This performance issue was previously characterized as having low to moderate risk significance (White) in NRC Inspection Report No. 50-305/2000015(DRS). During this supplemental inspection, performed in accordance with Inspection Procedure95002, the inspector concluded that the licensee performed a comprehensive evaluation of the unsuccessful staff augmentation drills. The licensees evaluation identifiedtwo root causes which resulted in the unsuccessful drills and in the staffs inability tocorrect thedeficiencies: (1) Management has not effectively acted to provide increased depth and flexibility in the emergency response organization following a reduction in staffing several years ago; and (2)
Management has accepted an adverse trend of test failures without requiring investigation into the root causes. The inspector reviewed the licensees corrective actions, both completed and planned, and concluded that the programmatic corrective actions appeared to address the identified root causes. In particular, the licensee assigned certain positions to an on-call rotation to ensure personnel were capable of augmenting in a timely manner, and the licensee was progressing in training additional staff to increase the depth of personnel assigned to key emergency response positions. In addition, the licensee was continuing its efforts in improving its corrective action program. The inspector reviewed the licensees immediate response to the issue and identified that one of the licensees initial corrective actions resulted in a Non-Cited Violation of regulatory requirements. To obtain a timely response of a key emergency response position (severe accident management - core hydraulics), the emergency preparedness staff effectively changed the emergency plan without revising the necessary procedures and without formally assessing the impact of that change. The staff instructed and trained personnel to respond to a location other than the Technical Support Center, which was contrary to the licensees current implementing procedures. While this change enabled the licensee to augment its staff in a timely manner, the change was not performed in accordance with NRC requirements. In order to make such a change, the licensees emergency plan required that the change be formally assessed to ensure that it did not reduce the effectiveness of the plan or any other implementing procedure. Since the issue did not result in a failure to meet an emergency preparedness planning standard, the failure to adequately implement the emergency plan was determined to be a violation of very low safety significance (Green) (Section 02.3(a)). Due to the licensees acceptable performance in assessing the emergency response augmentation drill deficiencies, the White finding associated with this issue will only be considered in assessing plant performance for a total of four quarters in accordance with the guidance in IMC 0305, Operating Reactor Assessment Program. Implementation of the licensees corrective actions will be reviewed during a future inspection.
Inspection Report# : 2001007(pdf)
Significance:          Mar 23, 2001 Identified By: NRC Item Type: NCV NonCited Violation
 
3Q/2000 Inspection Findings - Kewaunee                                                                                              Page 9 of 11 POST-ACCIDENT SAMPLING SYSTEM CONTAINMENT AIR SAMPLE PANEL TESTING.
A Non-Cited Violation of Technical Specification 6.14 was identified for the failure to implement a program that ensured the capability to obtain and analyze containment atmosphere samples under accident samples using the cantainment air sampling panel (CASP). Although the CASP was installed, as was indicated in the emergency plan, the licensee had neither developed procedures nor had tested its capability to obtain a containment atmosphere sample using the CASP. The licensee could not recall if and when containment air samples were last obtained using the CASP.
Inspection Report# : 2001006(pdf)
Significance: N/A Feb 27, 2001 Identified By: NRC Item Type: FIN Finding SUPPLEMENTAL INSPECTION OF YELLOW ANS PERFORMANCE INDICATOR AND ASSOCIATED CORRECTIVE ACTION PROGRAM DEFICIENCIES.
This supplemental inspection was performed by the NRC to assess the licensee's evaluation associated with a Yellow performance indicator for the Alert and Notification System (ANS) and the associated Yellow finding related to the licensee's corrective action program. During this supplemental inspection, performed in accordance with Inspection Procedure 95002, the inspectors concluded that the licensee performed comprehensive evaluations of the performance problems associated with the ANS and its corrective action program. These evaluations identified primary root causes and contributing causes for both issues. Along with the electronics/hardware problems, the licensee identified the primary root causes for the ANS to be the failure to make changes to the system via a change control process and the failure of the activation procedure to provide for alternate/backup activation methods and to provide clear success criteria. In the case of the corrective action program, the licensee concluded that the primary root causes were plant management's inadequate risk evaluation regarding decisions affecting the corrective action program and the development of a culture in the licensee's organization, which minimized the importance of information from outside organizations. In particular, the licensee focused on low-cost power operation and failed to recognize the value of changes and improvements which had occurred throughout the industry, such as the value and expansion of the corrective action program. In the case of the Yellow ANS performance indicator, the licensee had completed several significant corrective actions to address the root causes and contributing causes identified in its evaluation. The inspectors found that the corrective actions appeared appropriate to address the underlying root causes and that ANS testing data indicated an improving trend in the NRC performance indicator. The licensee also performed comprehensive assessments of the emergency preparedness program, quality assurance program, plant operations, and other plant programs to determine the extent of condition (re. the root causes described above). Based on these evaluations, the licensee began to implement significant actions to correct the deficiencies in the corrective action program and other weaknesses identified. Generally, the inspectors observed progress in the licensee's initial implementation of these corrective actions. Due to the licensee's acceptable performance in assessing the Yellow ANS performance indicator and the associated Yellow finding, the Yellow finding will not be considered in assessing future plant performance.
Inspection Report# : 2001005(pdf)
Occupational Radiation Safety Significance:        May 19, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO POST A VERY HIGH RADIATION AREA.
The inspectors identified a noncited violation for the failure to post a very high radiation area in accordance with 10 CFR 20.1902(c). Although the area was not adequately posted, the licensee had provided physical controls and barriers that were consistent with its requirements for a very high radiation area. Based on the adequacy of these controls, the potential for an overexposure from the inadvertent entry of personnel into the area was low. Consequently, this finding was determined to be of very low safety significance (Section 20S1.1).
Inspection Report# : 2000009(pdf)
Significance:        Oct 02, 2001 Identified By: NRC Item Type: NCV NonCited Violation HIGH RADIATION AREA ACCESS CONTROLS Non-Cited Violation of Technical Specification 6.13 and an associated Green Finding for failure to 'barricade' three ladders that provided entry to high radiation areas (less than 1000 mrem/hour) located on the steam generator/pressurizer platforms.
Inspection Report# : 2001014(pdf)
Public Radiation Safety
 
3Q/2000 Inspection Findings - Kewaunee                                                                                                Page 10 of 11 Physical Protection Significance: N/A Jul 21, 2000 Identified By: NRC Item Type: FIN Finding CORRECTIVE ACTION ON SEARCH ISSUES DID NOT WORK.
The inspector determined that the licensee's effectiveness of implemented corrective actions for a previously identified inspection finding regarding an inadequate vehicle search was not totally effective in preventing recurrence. Previous corrective action was not adequately focused (Section 3PP2.2).
Inspection Report# : 2000013(pdf)
Significance:        Jul 21, 2000 Identified By: NRC Item Type: NCV NonCited Violation SEARCHES INADEQUATE (REPEAT).
The inspector identified a Non-Cited violation by observing that, a security officer failed to search an easily accessible compartment on one vehicle.
The failure resulted from human error because the officer did not observe the access panel to the compartment during the vehicle search process.
When searched, no prohibited items were found. Corrective actions were implemented. The inspector determined that the licensee's effectiveness of implemented corrective actions for a previously identified inspection finding regarding an inadequate vehicle search was not totally effective in preventing recurrence. Previous corrective action was not adequately focused.
Inspection Report# : 2000013(pdf)
Miscellaneous Significance: N/A Sep 29, 2000 Identified By: NRC Item Type: FIN Finding EFFECTIVENESS OF CORRECTIVE ACTION PROGRAM.
Based on the results of this inspection, the NRC concluded that the corrective action program at Kewaunee showed significant weaknesses and inconsistencies across all of the procedural elements inspected. These weaknesses existed across departments and affected multiple cornerstones in the strategic performance areas of Reactor Safety, Radiation Safety, and Safeguards. Of particular note was the lack of procedures for determining the significance of conditions adverse to quality and for trending of issues and the complete lack of trending within your corrective action program. We also identified a lack of urgency in correcting issues which resulted in repeat examples occurring and, coupled with a poor tracking system, a tendency for issues to be dropped. While none of the specific examples identified by the team were of high risk significance when looked at in isolation, in the aggregate they were similar in nature to prior issues in the emergency preparedness area that rose to a higher significance level and contributed to a degraded cornerstone. While we concluded that the station had fostered an environment in which personnel freely identified conditions adverse to quality without fear of discrimination or retaliation, we also concluded that significant weaknesses with, and inconsistent implementation of, the station corrective action program resulted in multiple examples where station personnel did not enter deficiencies into the station's formal corrective action program.
Inspection Report# : 2000019(pdf)
Significance: N/A Sep 29, 2000 Identified By: NRC Item Type: NCV NonCited Violation NO PROCEDURAL GUIDANCE FOR DETERMINING IF CONDITIONS ADVERSE TO QUALITY ARE SIGNIFICANT.
The inspectors reviewed the quality assurance (QA) manual requirements against Kewaunee's implementing procedures and identified that two QA manual requirements were not being implemented. Specifically, Program Requirement 3.1.9 which stated that directives and procedures shall provide for the review of conditions adverse to quality to determine if the conditions are significant in nature. This requirement paralleled 10 CFR Part 50, Appendix B, Criterion XVI, which requires that the cause of significant conditions adverse to quality be determined and corrective actions taken to prevent recurrence. The inspectors reviewed the Nuclear Administrative Directive (NAD 11.08) and the procedure (GNP 11.08.01) governing the KAP (Kewaunee corrective action program) process and found no procedure requirements for identifying significant conditions adverse to quality. This finding does not directly affect a cornerstone. As a result, this issue was not evaluated with the Significance Determination Process and was not assigned a color. One example of a non-cited violation of 10 CFR Part 50, Appendix B, Criterion II, "Quality Assurance Programs," was identified.
Inspection Report# : 2000019(pdf)
Significance: N/A Sep 29, 2000
 
3Q/2000 Inspection Findings - Kewaunee                                                                                                Page 11 of 11 Identified By: NRC Item Type: NCV NonCited Violation NO PROCEDURAL GUIDANCE FOR TRENDING CONDITIONS ADVERSE TO QUALITY.
The inspectors reviewed the quality assurance (QA) manual requirements against Kewaunee's implementing procedures and identified that two QA manual requirements were not being implemented. Specifically, QA Program Requirement, 3.1.10, stated that directives and procedures shall provide for analyzing trends of conditions adverse to quality. Once identified these trends were required to be considered significant conditions adverse to quality. The inspectors found that conditions adverse to quality were not defined in the KAP (Kewaunee corrective action program) procedures and that no procedure existed for trending. This finding does not directly affect a cornerstone. As a result, this issue was not evaluated with the Significance Determination Process and was not assigned a color. One example of a non-cited violation of 10 CFR Part 50, Appendix B, Criterion II, "Quality Assurance Program," was identified.
Inspection Report# : 2000019(pdf)
Significance: N/A Jun 22, 2000 Identified By: NRC Item Type: FIN Finding CONTROL ROOM OPERATIONS HUMAN PERFORMANCE ISSUES.
The inspectors interviewed operators to evaluate their awareness of degraded control room indications and alarms, and their ability to adequately take manual actions based on degraded alarm functions. The inspectors identified, during interviews, that there was a lack of awareness by operators of a degraded refueling water storage tank low-low level alarm which would be potentially confusing to operators and therefore increase the risk associated with initiating long term sump recirculation.
Inspection Report# : 2000008(pdf)
Significance: N/A Feb 21, 2002 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Implement Required Fire Watch Following Completion of Hot Work Activities A licensee-identified violation was reviewed by the inspectors. Corrective actions taken or planned by the licensee appeared reasonable.
Inspection Report# : 2001017(pdf)
Significance: N/A Aug 24, 2001 Identified By: NRC Item Type: FIN Finding IDENTIFICATION AND RESOLUTION OF PROBLEMS The team concluded that the licensee was generally effective at identifying problems and putting them into the corrective action program. The program itself contained all the necessary attributes of an acceptable corrective action program and was generally successful in correcting identified issues. However, the team noted that, although licensee management had taken efforts to ensure that issues were resolved in accordance with program guidance and requirements, additional efforts appeared necessary to ensure timely resolution of issues. A positive program initiative was the establishment of positions in each of the major plant departments to serve as liaisons between the departments and the corrective action program and to assist with self-assessments. However, examples were identified by the inspectors of problems with the licensee's identification and resolution of problems, prioritization and evaluation of issues, and the effectiveness of corrective actions. Included in these examples were the routine granting of due date extensions for problem evaluation and corrective action implementation, failure to perform a Technical Specification-required biennial surveillance of safety-related procedures, and ineffective corrective actions that resulted in the lack of proper controls over the storage of grease used in safety-related breakers. Based on a review of records and discussions with plant staff, the inspectors concluded that workers at the site felt free to input safety issues into the corrective action program.
Inspection Report# : 2001012(pdf)
Significance: N/A Nov 09, 2000 Identified By: NRC Item Type: FIN Finding FAILURE TO TAKE ADEQUATE CORRECTIVE ACTIONS TO ENSURE COMPONENT RE-TEST REQUIREMENTS COMPLETED ACCORDING TO MAINTENANCE PROCEDURES.
No Color. The inspectors determined that a negative performance trend had developed in the licensee's ability to identify and promptly take appropriate corrective actions to prevent recurrence based on two previously identified examples (NCV 50-305/2000014-01) and one example identified during this inspection period (NCV 50-305/00-20-01). All three examples related to the licensee returning safety-related equipment to service prior to completing all required post-maintenance retesting. While the risk of the individual examples was very low, the licensee had failed to ensure that all retest requirements had been completed before returning safety-related equipment to service. These findings collectively indicated a problem with the licensee's ability to provide timely and adequate corrective actions to prevent recurrence. (Section 4OA2).
Inspection Report# : 2000020(pdf)
Last modified : March 29, 2002
 
4Q/2000 Inspection Findings - Kewaunee                                                                                                  Page 1 of 11 Kewaunee Initiating Events Mitigating Systems Significance: N/A Nov 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO COMPLETE COMPONENT RE-TEST REQUIREMENTS PRIOR TO RETURNING EQUIPMENT TO AN OPERABLE STATUS.
No Color. The inspectors identified a Non-Cited Violation for failure to complete component retest requirements following maintenance performed on the B train control room air conditioner compressor condenser. The unit had been returned to an operable status prior to the retest requirements being completed as prescribed in the associated maintenance procedure. This issue was determined to be a violation of the licensee's Operational Quality Assurance Program Manual, Section 8, "Maintenance Planning and Control." Although the risk associated with this finding was very low and did not affect any cornerstones, the inspectors noted that this finding was similar to previous NRC-identified findings and therefore was of greater than minor significance and warranted documentation. (Section 1R19).
Inspection Report# : 2000020(pdf)
Significance:        Sep 29, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IDENTIFY CORROSION AS POTENTIAL FAILURE MECHANISM.
The inspectors identified that the licensee failed to identify corrosion as a potential failure mechanism in the operability determination for a carbon steel key in the service water system. Thus, the licensee failed to quantify the corrosion rate and therefore did not adequately evaluate the expected service life of the carbon steel key. One non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was identified.
Inspection Report# : 2000019(pdf)
Significance:        Sep 29, 2000 Identified By: NRC Item Type: NCV NonCited Violation PRESSURE RATING OF AUXILIARY FEEDWATER STEAM TRAPS.
The inspectors identified that a root cause evaluation for a 1996 equipment issue in the turbine-driven auxiliary feedwater system was not completed until 1999. The evaluation stated that the internals of the steam traps were designed to operate at pressures up to a maximum 600 psig but that the traps were exposed to pressures up to 1025 psig. A corrective action item to initiate a design change request to replace the steam traps with a different model rated for the design pressure of the system was described in the evaluation. However, the inspectors identified that the design change request had never been initiated and the KAP ( Kewaunee corrective action document) had been closed. As a result, the corrective action item for this design problem was lost. In addition, operability of the system had never been formally evaluated despite the identification that the system design requirements were not met. The licensee subsequently determined that the steam traps remained operable and was planning to initiate the design change to correct the problem. One non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was identified.
Inspection Report# : 2000019(pdf)
Significance:        Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO DOCUMENT IN THE CORRECTIVE ACTION PROGRAM THE USE OF INCORRECT MATERIAL IN A SERVICE WATER PUMP KEY.
An example of a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was identified because of inadequate corrective action to correct an incorrect coupling adjust nut set screw and a low strength "soft" key material, which had contributed to a pump shaft failure. Licensee personnel had known of the "soft" key material since July 21, 1999. The "soft" key material was found in other service water pumps but had not been removed from all pumps. As of July 25, 2000, licensee personnel had not documented the existence of the "soft" key material in the corrective action program.
 
4Q/2000 Inspection Findings - Kewaunee                                                                                              Page 2 of 11 Inspection Report# : 2000012(pdf)
Significance:          Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation DESIGN CONTROL VIOLATION FOR AUXILIARY FEEDWATER STRAINER MESH SIZE.
The inspectors questioned the mesh size of the strainers, which were installed in the suction of the three auxiliary feedwater (AFW) pumps. As a result of the inspectors' questions, license personnel inspected the strainers on August 21, 2000, and found the strainers to have 1/16 inch openings. A note was later found on Figure 10.2-3 of the UFSAR that indicated that the AFW suction strainer size was 1/8 inch. The smaller openings would not support the use of service water as a safety related source for AFW and as a result all three trains of AFW were declared inoperable. This condition had apparently existed for approximately 25 years and was identified as a non-cited violation of Criterion III, "Design Control," of 10 CFR 50, Appendix B.
Inspection Report# : 2000012(pdf)
Significance: N/A Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation RETRIEVAL OF SERVICE WATER SYSTEM DESIGN INFORMATION.
In many cases, design basis information for the service water system was difficult if not impossible to locate. Licensee personnel wrote KAP WO 00-002566 to enter the problem in the corrective action program.
Inspection Report# : 2000012(pdf)
Significance:          Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PROPERLY CONTROL CORRECTED TEST DATA NECESSARY FOR DESIGN CALCULATIONS.
An example of a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," was identified in the handling of service water system flow test data, which was subsequently used in calculations. Gauge readings corrected for post test calibration checks, gauge reading corrections for elevation considerations, and flow values corrected for pump degradation were contained in spreadsheets in the possession of an individual staff member, but not currently packaged with raw test data, and not bearing evidence of a formal review and control process. The connection between the test data, which had been vaulted, and the values used in the calculation, could not be made without use of the uncontrolled spreadsheet.
Inspection Report# : 2000012(pdf)
Significance:          Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation IMPROPER DESIGN CALCULATION IDENTIFICATION, NON-CONSERVATIVE ASSUMPTIONS, CALCULATION ERRORS, AND DUPLICATE CALCULATIONS.
An example of a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," was identified because of inadequate control of design calculations. The control failures included improper identification of calculations, non-conservative assumptions, calculation errors, and duplicate or superceded calculations not properly identified or canceled. The failure to follow the established design control process increased the potential for errors in the design and operation of the service water system. Because the system was subsequently demonstrated to be capable of removing the design heat load, the actual significance was low and this finding screened out as having very low risk significance.
Inspection Report# : 2000012(pdf)
Significance:          Jul 07, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO COMPLETE COMPONENT RETEST REQUIREMENTS IN ACCORDANCE WITH PROCEDURE.
On June 26, 2000, during a review of post maintenance testing requirements following maintenance performed on the control room post accident system charcoal filter heat detector, the inspectors identified that maintenance technicians had not completed the component re-test requirements, as required by a preventative maintenance procedure prior to the system being returned to an operable status. On July 7, the inspectors identified a second example of failing to complete component re-test requirements following maintenance on the zone special ventilation system charcoal filter heat detector as required. The issue was considered to be of very low safety significance based on the determination that although the licensee had not completed all of the component retest requirements prior to returning the equipment to service, the subsequent testing determined that the equipment was in an operable status. The failure to complete the component retest requirements in accordance with site procedures was
 
4Q/2000 Inspection Findings - Kewaunee                                                                                                    Page 3 of 11 identified as a Non-Cited Violation.
Inspection Report# : 2000014(pdf)
Significance:        Jun 22, 2000 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO TEST ADDITIONAL RELIEF VALVES IN ACCORDANCE WITH TECHINCAL SPECIFICATIONS.
The licensee identified that the suction relief valve for an auxiliary feedwater pump may have failed its relief test criteria, but did not process the documented deficiency until several weeks later. The licensee then expanded the scope of the relief testing to the suction relief valves associated with the other auxiliary feedwater pumps to meet technical specification requirements. Since any one train of auxiliary feedwater was capable of supplying 100 percent of the decay heat removal requirements, this issue was screened as very low risk significance. However, the time delay in complying with technical specification requirements for testing other relief valves was identified as an NCV.
Inspection Report# : 2000008(pdf)
Significance:        Jun 22, 2000 Identified By: NRC Item Type: FIN Finding FAILURE TO MEET SINGLE FAILURE CRITERIA FOR RESIDUAL HEAT REMOVAL VALVE CIRCUITRY.
The licensee identified that the circuitry associated with the residual heat removal system discharge to safety injection system suction isolation valves did not meet single failure criteria. The inspectors noted that this design requirement was identified in the facility's updated safety analysis report. The licensee subsequently implemented a temporary change to the facility. The inspectors reviewed the issue and identified that the facility had been operating outside of its design basis, which was reportable to the NRC. The licensee subsequently made a one hour non-emergency report to the NRC. Since there was no actual loss of safety function to the system, this issue was screened as very low risk significance.
Inspection Report# : 2000008(pdf)
Significance:        Jun 22, 2000 Identified By: NRC Item Type: FIN Finding MAINTENANCE RULE FAILURES ASSOCIATED WITH REACTOR HEAD VENT VALVE.
The inspectors reviewed the licensee's implementation of the maintenance rule for failures associated with a reactor head vent valve. The licensee's corrective action documents identified a potential maintenance rule functional failure but the completed evaluation of the problem did not document the final determination. However, the inspectors identified that the repeated failures may have been prevented if maintenance activities such as valve disassembly and cleaning had been performed. In this case, maintenance rule reliability goals were not exceeded. The licensee had documented similar maintenance rule program deficiencies and developed a corrective action program to address the deficiencies. Although programmatic deficiencies exist, since no maintenance rule reliability criteria had been exceeded, this issue was considered of very low risk significance.
Inspection Report# : 2000008(pdf)
Significance:        Jun 22, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO INITIATE KEWAUNEE ASSESSMENT PROCESS DOCUMENT REGARDING REFUELING WATER STORAGE TANK LOW-LOW LEVEL ALARM INOPERABILITY.
The inspectors identified that the refueling water storage tank low-low level alarm which was actuating five percent higher than normal had not been documented in a Kewaunee Assessment Process form by the licensee, and therefore had not received an operability evaluation. This failure was identified as contrary to site administrative procedures. Following the licensee's documentation of the problem, the inspectors identified that the associated operability evaluation considered the acceptability of an operator workaround to address the issue, but did not address any safety implications or consequences of the alarm actuating early. A subsequent operability evaluation by the licensee was evaluated as adequate by the inspectors. Since the subsequent operability evaluation was adequate and it was determined that no safety mitigation equipment was adversely affected by the early actuation of the alarm, this issue was considered of very low risk significance. A non-cited violation (NCV) was identified for failing to document a non-conforming condition, contrary to site administrative procedure requirements.
Inspection Report# : 2000008(pdf)
Significance:        May 22, 2000 Identified By: NRC
 
4Q/2000 Inspection Findings - Kewaunee                                                                                                  Page 4 of 11 Item Type: FIN Finding FIRE EXTINGUISHERS NOT LOCATED IN ALL AREAS OF CONTAINMENT BUILDING AT BEGINNING OF REFUELING OUTAGE.
During a walkdown of the containment building, the inspectors identified that portable fire extinguishers were not located in the containment basement at the beginning of the plant refueling outage. Additionally, site fire protection procedures required that responsible fire protection personnel perform inspections of selected plant areas to ensure that the quantity of combustible material was minimized. However, the procedure did not list the containment as an area to be inspected and the procedures did not require the placement and location of portable fire suppression equipment inside containment during the refueling outage. Due to a low number of work activities ongoing at the time, this issue was screened as Green (very low risk significance).
Inspection Report# : 2000007(pdf)
Significance:          May 22, 2000 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO INSTALL RAYCHEM HEAT SHRINK MATERIAL IN ACCORDANCE WITH PROCEDURE REQUIREMENTS.
The licensee identified that two Raychem electrical cable splices utilized in environmentally qualified (EQ) safety-related equipment had not been installed in accordance with EQ requirements. These splices were associated with pressurizer level transmitters and were installed in 1984. The licensee subsequently performed extensive EQ testing of the splices to determine the qualification of the splices' as-found configurations. Test results indicated that the splices would have been able to perform their intended function in a harsh environment inside containment. This issue was considered to be of low safety significance based on the successful EQ testing of the as-found splices' configurations and was screened as Green (very low risk significance). One non-cited violation was identified.
Inspection Report# : 2000007(pdf)
Significance:          Feb 21, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform Thorough 10 CFR 50.59 Safety Evaluation The inspectors identified a Non-Cited Violation for failure to perform an adequate 10 CFR 50.59 safety evaluation associated with emergency operating procedure changes to address component cooling water pump dead-head operational concerns. The safety evaluation did not evaluate the potential for initiating a loss-of-coolant accident via the reactor coolant loop seals during conditions of a complete loss of component cooling water.
Inspection Report# : 2001017(pdf)
Significance: N/A Sep 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ESTABLISH CONTINGENCY PLANS FOR ORANGE RISK CONDITION The inspectors identified the failure to establish contingency plans during a planned high risk plant configuration. Contrary to administrative requirements, the licensee approved an orange risk condition during a refueling outage with no contingency plans to mitigate the consequences of a loss of spent fuel pool cooling with a full core offload in the pool. A Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was identified. The finding was of very low safety significance because although the licensee had not approved appropriate contingency actions for the orange risk condition, the licensee subsequently rescheduled the planned maintenance to eliminate the orange risk condition.
Inspection Report# : 2001013(pdf)
Significance: N/A Aug 24, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO CONDUCT BIENNIAL SURVIELLANCE OF SAFETY-RELATED PROCEDURES PER TS 6.8.c A Non-Cited Violation of Technical Specification 6.8.c was identified for the failure to perform a biennial surveillance of safety-related procedues.
This issue was more than minor because if left uncorrected, could under the same condition become a more significant safety concern. However, since no specific cornerstone had been impacted, this finding is designated as No Color.
Inspection Report# : 2001012(pdf)
Significance:          Aug 24, 2001 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE CORRECTIVE ACTIONS FOR PREVIOUS PROBLEM WITH THE CONTROL OF SAFETY-RELATED MATERIALS A Non-Cited Violation of Criterion XVI, "Corrective Action," of Appendix B of 10 CFR Part 50 was identified for ineffective corrective actions for a problem with the control of the storage of consumable materials, such as thread sealant, used in safety-related applications. These ineffective
 
4Q/2000 Inspection Findings - Kewaunee                                                                                                  Page 5 of 11 actions subsequently resulted in the inadequate control of the storage of grease used in safety-related breakers. This issue was more than minor because if left uncorrected, could under the same condition become a more significant safety concern. In that this issue could credibly affect the operability, availability, reliability, or function of a system or train in a mitigating system, it is a Green finding.
Inspection Report# : 2001012(pdf)
Significance:          Aug 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation NON-RATED FIRE BARRIER A Non-Cited Violation [of 10 CFR Part 50, Appendix R, Section III.G.2.a] was identified for failure to provide a 3-hour rated fire barrier to separate redundant trains of safe shutdown equipment. This finding was of very low safety significance because the licensee tested a replica of the fire barrier and demonstrated that the fire barrier provided protection for at least 60 minutes, which was sufficient for the hazards in the area.
Inspection Report# : 2001011(pdf)
Significance:          Jul 20, 2001 Identified By: NRC Item Type: FIN Finding LICENSED OPERATOR REQUALIFICATION EXAMINATION RESULTS The inspectors identified that two of eight crews examined during the licensee's calendar year 2001 licensed operator requalification operating test had failed. The finding was of very low safety significance because both crews that had failed received remedial training prior to being returned to shift, and the results of the licensee's operator licensing requalification operating test given in calendar year 2000 indicated that only one crew, out of a total of eight crews tested, had failed.
Inspection Report# : 2001011(pdf)
Significance: N/A Jun 30, 2001 Identified By: NRC Item Type: FIN Finding FAILURE TO BALANCE RELIABILITY AND AVAILABILITY AS REQUIRED BY 10 CFR 50.65(a)(3)
The inspectors identified a failure to evaluate whether adjustments were necessary such that there would be an appropriate balance between systems' availability and reliability in accordance with 10 CFR 50.65(a)(3) of the maintenance rule. The inspectors identified that the licensee did not have an administrative process to track maintenance rule functional failures and maintenance preventible maintenance functional failures. As a result, reliability and availability could not be balanced as required by the Maintenance Rule periodic evaluation. The safety significance of the specific finding was very low because it did not affect the operability of the systems, and the licensee entered the finding in the corrective action program. However, this finding was considered to be of regulatory concern in the area of maintenance rule implementation due to the extent of the problems with the Maintenance Rule Program.
Inspection Report# : 2001009(pdf)
Significance:          Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IDENTIFY DEFICIENT CONDITION OF VALVE AFW-1B The inspectors identified that the licensee failed to promptly identify and correct the B' train auxiliary feedwater pump discharge check valve which was stuck in an intermediate position. A Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was issued. The finding was of very low safety significance because, although the check valve was stuck in an intermediate position, the time that it was known to have been stuck was less than the technical specification allowed outage time for one train of auxiliary feedwater to be out of service (less than 72 hours). Additionally, the other two trains of auxiliary feedwater were each capable of 100 percent decay heat removal.
Inspection Report# : 2001009(pdf)
Significance:          Jun 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO TRACK UNAVAILABILITY OF SYSTEMS REQUIRED DURING SHUTDOWN OPERATION 10 CFR 50.65(a)(1), required, in part, that the licensee monitor the performance or condition of SSCs within the scope of the rule as defined by 10 CFR 50.65(b), against licensee-established goals, in a manner sufficient to provide reasonable assurance that such SSCs are capable of fulfilling their intended functions. 10 CFR 50.65(a)(2) stated, in part, that monitoring as specified in 10 CFR 50.65(a)(1) was not required where it had been demonstrated that the performance or condition of an SSC was being effectively controlled through the performance of appropriate preventive maintenance, such that the SSC remained capable of performing its intended function. Contrary to the above, the licensee failed to demonstrate
 
4Q/2000 Inspection Findings - Kewaunee                                                                                                      Page 6 of 11 that the performance or condition of systems required to be available during shutdown conditions and within the scope of the rule had been effectively controlled through the performance of appropriate preventive maintenance and did not monitor against licensee-established goals.
Specifically, the licensee failed to monitor the unavailability of systems required during shutdown operation.
Inspection Report# : 2001009(pdf)
Significance:        Jun 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO ESTABLISH MAINTENANCE RULE (a)(1) GOALS 10 CFR 50.65(a)(1), requires, in part, that the licensee monitor the performance or condition of SSCs within the scope of the rule as defined by 10 CFR 50.65(b), against licensee-established goals, in a manner sufficient to provide reasonable assurance that such structures, systems, and components, are capable of fulfilling their intended functions. Such goals shall be established commensurate with safety. When the performance or condition of a structure, system, or component does not meet established goals, appropriate corrective action shall be taken. Contrary to the above, from 1996, the licensee did not take appropriate corrective actions when the performance of those systems in (a)(1) did not meet licensee established goals. Specifically, the licensee determined timely and appropriate corrective actions had not been taken for five systems that had been in (a)(1) category for approximately 3 years to 5 years: component cooling (entered (a)(1) on April 23, 1997), control room air conditioning (July 24, 1996), station and instrument air (July 3, 1997), auxiliary building air ventilation (July 31, 1997), and control rod drive (August 6, 1998). This issue is in the licensee's corrective action system as KAP WO 01-3323. The inspectors evaluated the risk significance of this issue using the Significance Determination Process. The inspectors did not identify where this failure resulted in a total loss of a risk significant SSC. Therefore, this issue was screened as Green (very low risk significance) after a Phase 1 Significance Determination Process review. Although the risk significance of this issue was low, the inspectors concluded that this was more than a minor concern because the failure to recognize and correct ineffective maintenance practices resulted in risk significant systems in (a)(1) for years with no improvement in performance. The NRC tracking number for this issue is 50-305/01-09-02.
Inspection Report# : 2001009(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation FIRE DOOR FUSIBLE LINKS.
On February 20, 2001, the licensee determined that the installed fusible link arrangement on roll-up fire Doors 279 and 281, which separated both trains of service water pumps, would not actuate as designed to ensure that the doors would automatically close to provide a 3-hour fire barrier, contrary to 10 CFR Part 50, Appendix R, Section III.G.2.a which required, in part, separation of cables and equipment of redundant trains by a fire barrier having a 3-hour rating.
Inspection Report# : 2001006(pdf)
Significance:        Mar 12, 2001 Identified By: Licensee Item Type: NCV NonCited Violation INADEQUATE SMOKE DETECTOR COVERAGE IN FIRE ZONE TU-95B.
Licensee identified violation of licensee's operating license that the licensee failed to install a detector in each beam pocket in Fire Zone TU-95B.
Inspection Report# : 2001002(pdf)
Significance:        Feb 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO TEST FIRE DOOR IN ACCORDANCE WITH FIRE PLAN.
The inspectors identified a non-cited violation for failure to properly test a fire door in accordance with the facility's fire protection program plan. The finding was of very low safety significance because, although the fire door separated both trains of service water pumps and did not fully close as designed when subsequently tested, the fire loading in the area was insufficient to credibly impact more than two of the four service water pumps in the area.
Inspection Report# : 2001004(pdf)
Barrier Integrity
 
4Q/2000 Inspection Findings - Kewaunee                                                                                                  Page 7 of 11 Significance:        Jun 22, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation FAILURE TO UPDATE COMPUTER ALARM FOR CURRENT AXIAL FLUX DISTRIBUTION TARGET BAND.
The licensee identified, following plant startup, that a computer alarm had not been updated properly to alarm if axial flux distribution deviated outside of the flux distribution target band. This condition was contrary to technical specification requirements. The licensee reviewed the axial flux distribution history since the startup and determined that at no time was the flux distribution outside of the target band. Since the axial flux distribution was never outside of the target band, this issue was screened as very low risk significance. An NCV was identified for failing to comply with technical specification requirements for monitoring axial flux distributions.
Inspection Report# : 2000008(pdf)
Significance:        May 22, 2000 Identified By: Licensee Item Type: FIN Finding TECHNICAL SPECIFICATIONS INTERPRETATION FOR TESTING REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES.
The NRC determined that the licensee's practice of testing reactor coolant system pressure isolation Valve SI-22B prior to entering the cold shutdown condition was contrary to Technical Specification requirements 4.2.a.3.a. Technical Specification 4.2.a.3.a required that periodic leakage testing of Valve SI-22B be accomplished prior to reaching operating mode after the plant was placed in cold shutdown. This issue was considered to be of low safety significance because of a subsequent successful valve test and was screened as Green (very low risk significance).
Enforcement discretion was applied to this item in accordance with Section VII.B.6 of the Enforcement Policy.
Inspection Report# : 2000007(pdf)
Significance:        May 19, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PROPERLY IMPLEMENT FLAW ACCEPTANCE CRITERIA FOR SLEEVE WELD INSPECTIONS.
During 1998 inservice inspection examinations, the licensee failed to properly implement the flaw acceptance criteria for laser welded sleeve inspection within two steam generator tubes. The safety significance was very low based on the absence of adverse consequences, and May 2000 in-situ pressure testing where both welds exhibited zero leakage at normal operating pressure, main steam line break pressure, and three times normal operating differential pressures. As such, this issue was characterized as a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion IX, Control of Special Processes. To correct the error, the licensee plugged both tubes (Section 1RO8).
Inspection Report# : 2000011(pdf)
Emergency Preparedness Significance:        Sep 21, 2000 Identified By: NRC Item Type: FIN Finding INADEQUATE ROOT CAUSE EVALUATION FOR YELLOW ALERT AND NOTIFICATION SYSTEM PERFORMANCE INDICATOR.
The licensee's evaluation of the Yellow Alert and Notification (siren) System Performance Indicator (PI) was inadequate. The inspector concluded that the licensee's evaluation was not performed at the depth necessary to identify the root causes of the siren performance problems and, instead, only identified the symptoms of the root causes. Specifically, the inspector identified the following substantive weaknesses in the licensee's evaluation of the siren system performance, which appeared to result from systemic corrective action program deficiencies within this cornerstone:
* The licensee's evaluation was not of sufficient depth to clearly identify the root causes associated with the decline in siren system performance.
* Licensee management did not provide well-understood and clear guidance/expectations for performing root cause evaluations.
* The licensee's evaluation of the quality assurance program was narrowly focused and was not critical of its role in failing to identify and correct the siren performance problems.
* The licensee did not establish a priority for each of the long-term corrective actions in accordance with the associated significance or risk.
* The licensee did not have any formal provisions for measuring the effectiveness of its corrective actions.
* Within the licensee's evaluation, the licensee had not evaluated common causes or the extent of the condition. Due to the corrective action program performance deficiencies within this cornerstone, we have been unable to conclude that the performance issues that resulted in the yellow PI have been addressed. Therefore, we are issuing a yellow finding that corresponds to the original issues that resulted in a yellow PI. Additional inspection effort will be focused on the licensee's further evaluation of the siren reliability root causes and the continuing corrective action program implementation deficiencies identified during this inspection.
Inspection Report# : 2000017(pdf)
 
4Q/2000 Inspection Findings - Kewaunee                                                                                                    Page 8 of 11 Significance: N/A Apr 05, 2000 Identified By: NRC Item Type: FIN Finding LICENSEE FAILED TO IDENTIFY THE FULL SCOPE OF PROBLEMS WITH THE ALERT AND NOTIFICATION SYSTEM PERFORMANCE.
The inspectors concluded that the licensee's assessment was not sufficiently comprehensive to identify the full scope of problems associated with the Alert and Notification System (ANS) performance program. As a result, licensee corrective actions generally were focused on the equipment problem rather than all root causes and contributing factors. For example, the inspectors identified that: (1) management oversight of the ANS performance program was limited; (2) an audit failed to identify degrading ANS performance as a concern; (3) annual preventive maintenance was not consistently performed on the system; (4) the corrective action program was not used consistently to document ANS problems; and (5) maintenance procedures and records were deficient. Collectively, these problems indicate that the ANS performance program lacked sufficient structure and oversight.
Inspection Report# : 2000006(pdf)
Significance:          Mar 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation SUPPLEMENTAL INSPECTION OF WHITE ERO AUGMENTATION FINDING AND RESULTING GREEN FINDING.
This supplemental inspection was performed by the NRC to evaluate the licensees evaluation associated with the failure to conduct successful quarterly, off-hours, unannounced staff augmentation drills during the second, third, and fourth quarters of 1999 and the second quarter of 2000.
This performance issue was previously characterized as having low to moderate risk significance (White) in NRC Inspection Report No. 50-305/2000015(DRS). During this supplemental inspection, performed in accordance with Inspection Procedure95002, the inspector concluded that the licensee performed a comprehensive evaluation of the unsuccessful staff augmentation drills. The licensees evaluation identifiedtwo root causes which resulted in the unsuccessful drills and in the staffs inability tocorrect thedeficiencies: (1) Management has not effectively acted to provide increased depth and flexibility in the emergency response organization following a reduction in staffing several years ago; and (2)
Management has accepted an adverse trend of test failures without requiring investigation into the root causes. The inspector reviewed the licensees corrective actions, both completed and planned, and concluded that the programmatic corrective actions appeared to address the identified root causes. In particular, the licensee assigned certain positions to an on-call rotation to ensure personnel were capable of augmenting in a timely manner, and the licensee was progressing in training additional staff to increase the depth of personnel assigned to key emergency response positions. In addition, the licensee was continuing its efforts in improving its corrective action program. The inspector reviewed the licensees immediate response to the issue and identified that one of the licensees initial corrective actions resulted in a Non-Cited Violation of regulatory requirements. To obtain a timely response of a key emergency response position (severe accident management - core hydraulics), the emergency preparedness staff effectively changed the emergency plan without revising the necessary procedures and without formally assessing the impact of that change. The staff instructed and trained personnel to respond to a location other than the Technical Support Center, which was contrary to the licensees current implementing procedures. While this change enabled the licensee to augment its staff in a timely manner, the change was not performed in accordance with NRC requirements. In order to make such a change, the licensees emergency plan required that the change be formally assessed to ensure that it did not reduce the effectiveness of the plan or any other implementing procedure. Since the issue did not result in a failure to meet an emergency preparedness planning standard, the failure to adequately implement the emergency plan was determined to be a violation of very low safety significance (Green) (Section 02.3(a)). Due to the licensees acceptable performance in assessing the emergency response augmentation drill deficiencies, the White finding associated with this issue will only be considered in assessing plant performance for a total of four quarters in accordance with the guidance in IMC 0305, Operating Reactor Assessment Program. Implementation of the licensees corrective actions will be reviewed during a future inspection.
Inspection Report# : 2001007(pdf)
Significance:          Mar 30, 2001 Identified By: NRC Item Type: VIO Violation FAILURE TO CORRECT SELF-IDENTIFIED ERO AUGMENTATION DRILL DEFICIENCIES.
During a baseline inspection of the emergency preparedness program conducted on August 14 - September 21, 2000, the NRC identified a preliminary White issue and potential violation for the licensees failure to successfully correct deficiencies identified during staff augmentation drills and to demonstrate timely staff augmentation in 1999 and 2000. The issue was unresolved pending the outcome of the NRCs final significance determination. On January 30, 2001, the NRC conducted a regulatory conference with the licensee and subsequently issued the licensee a White finding and Notice of Violation (Enforcement Action No. 00-214) associated with the performance issue.
Inspection Report# : 2001007(pdf)
Significance:          Mar 23, 2001 Identified By: NRC Item Type: NCV NonCited Violation POST-ACCIDENT SAMPLING SYSTEM CONTAINMENT AIR SAMPLE PANEL TESTING.
A Non-Cited Violation of Technical Specification 6.14 was identified for the failure to implement a program that ensured the capability to obtain and analyze containment atmosphere samples under accident samples using the cantainment air sampling panel (CASP). Although the CASP was
 
4Q/2000 Inspection Findings - Kewaunee                                                                                              Page 9 of 11 installed, as was indicated in the emergency plan, the licensee had neither developed procedures nor had tested its capability to obtain a containment atmosphere sample using the CASP. The licensee could not recall if and when containment air samples were last obtained using the CASP.
Inspection Report# : 2001006(pdf)
Significance: N/A Feb 27, 2001 Identified By: NRC Item Type: FIN Finding SUPPLEMENTAL INSPECTION OF YELLOW ANS PERFORMANCE INDICATOR AND ASSOCIATED CORRECTIVE ACTION PROGRAM DEFICIENCIES.
This supplemental inspection was performed by the NRC to assess the licensee's evaluation associated with a Yellow performance indicator for the Alert and Notification System (ANS) and the associated Yellow finding related to the licensee's corrective action program. During this supplemental inspection, performed in accordance with Inspection Procedure 95002, the inspectors concluded that the licensee performed comprehensive evaluations of the performance problems associated with the ANS and its corrective action program. These evaluations identified primary root causes and contributing causes for both issues. Along with the electronics/hardware problems, the licensee identified the primary root causes for the ANS to be the failure to make changes to the system via a change control process and the failure of the activation procedure to provide for alternate/backup activation methods and to provide clear success criteria. In the case of the corrective action program, the licensee concluded that the primary root causes were plant management's inadequate risk evaluation regarding decisions affecting the corrective action program and the development of a culture in the licensee's organization, which minimized the importance of information from outside organizations. In particular, the licensee focused on low-cost power operation and failed to recognize the value of changes and improvements which had occurred throughout the industry, such as the value and expansion of the corrective action program. In the case of the Yellow ANS performance indicator, the licensee had completed several significant corrective actions to address the root causes and contributing causes identified in its evaluation. The inspectors found that the corrective actions appeared appropriate to address the underlying root causes and that ANS testing data indicated an improving trend in the NRC performance indicator. The licensee also performed comprehensive assessments of the emergency preparedness program, quality assurance program, plant operations, and other plant programs to determine the extent of condition (re. the root causes described above). Based on these evaluations, the licensee began to implement significant actions to correct the deficiencies in the corrective action program and other weaknesses identified. Generally, the inspectors observed progress in the licensee's initial implementation of these corrective actions. Due to the licensee's acceptable performance in assessing the Yellow ANS performance indicator and the associated Yellow finding, the Yellow finding will not be considered in assessing future plant performance.
Inspection Report# : 2001005(pdf)
Occupational Radiation Safety Significance:        May 19, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO POST A VERY HIGH RADIATION AREA.
The inspectors identified a noncited violation for the failure to post a very high radiation area in accordance with 10 CFR 20.1902(c). Although the area was not adequately posted, the licensee had provided physical controls and barriers that were consistent with its requirements for a very high radiation area. Based on the adequacy of these controls, the potential for an overexposure from the inadvertent entry of personnel into the area was low. Consequently, this finding was determined to be of very low safety significance (Section 20S1.1).
Inspection Report# : 2000009(pdf)
Significance:        Oct 02, 2001 Identified By: NRC Item Type: NCV NonCited Violation HIGH RADIATION AREA ACCESS CONTROLS Non-Cited Violation of Technical Specification 6.13 and an associated Green Finding for failure to 'barricade' three ladders that provided entry to high radiation areas (less than 1000 mrem/hour) located on the steam generator/pressurizer platforms.
Inspection Report# : 2001014(pdf)
Public Radiation Safety
 
4Q/2000 Inspection Findings - Kewaunee                                                                                              Page 10 of 11 Physical Protection Significance: N/A Jul 21, 2000 Identified By: NRC Item Type: FIN Finding CORRECTIVE ACTION ON SEARCH ISSUES DID NOT WORK.
The inspector determined that the licensee's effectiveness of implemented corrective actions for a previously identified inspection finding regarding an inadequate vehicle search was not totally effective in preventing recurrence. Previous corrective action was not adequately focused (Section 3PP2.2).
Inspection Report# : 2000013(pdf)
Significance:          Jul 21, 2000 Identified By: NRC Item Type: NCV NonCited Violation SEARCHES INADEQUATE (REPEAT).
The inspector identified a Non-Cited violation by observing that, a security officer failed to search an easily accessible compartment on one vehicle.
The failure resulted from human error because the officer did not observe the access panel to the compartment during the vehicle search process.
When searched, no prohibited items were found. Corrective actions were implemented. The inspector determined that the licensee's effectiveness of implemented corrective actions for a previously identified inspection finding regarding an inadequate vehicle search was not totally effective in preventing recurrence. Previous corrective action was not adequately focused.
Inspection Report# : 2000013(pdf)
Miscellaneous Significance: N/A Nov 09, 2000 Identified By: NRC Item Type: FIN Finding FAILURE TO TAKE ADEQUATE CORRECTIVE ACTIONS TO ENSURE COMPONENT RE-TEST REQUIREMENTS COMPLETED ACCORDING TO MAINTENANCE PROCEDURES.
No Color. The inspectors determined that a negative performance trend had developed in the licensee's ability to identify and promptly take appropriate corrective actions to prevent recurrence based on two previously identified examples (NCV 50-305/2000014-01) and one example identified during this inspection period (NCV 50-305/00-20-01). All three examples related to the licensee returning safety-related equipment to service prior to completing all required post-maintenance retesting. While the risk of the individual examples was very low, the licensee had failed to ensure that all retest requirements had been completed before returning safety-related equipment to service. These findings collectively indicated a problem with the licensee's ability to provide timely and adequate corrective actions to prevent recurrence. (Section 4OA2).
Inspection Report# : 2000020(pdf)
Significance: N/A Sep 29, 2000 Identified By: NRC Item Type: NCV NonCited Violation NO PROCEDURAL GUIDANCE FOR TRENDING CONDITIONS ADVERSE TO QUALITY.
The inspectors reviewed the quality assurance (QA) manual requirements against Kewaunee's implementing procedures and identified that two QA manual requirements were not being implemented. Specifically, QA Program Requirement, 3.1.10, stated that directives and procedures shall provide for analyzing trends of conditions adverse to quality. Once identified these trends were required to be considered significant conditions adverse to quality. The inspectors found that conditions adverse to quality were not defined in the KAP (Kewaunee corrective action program) procedures and that no procedure existed for trending. This finding does not directly affect a cornerstone. As a result, this issue was not evaluated with the Significance Determination Process and was not assigned a color. One example of a non-cited violation of 10 CFR Part 50, Appendix B, Criterion II, "Quality Assurance Program," was identified.
Inspection Report# : 2000019(pdf)
Significance: N/A Sep 29, 2000 Identified By: NRC Item Type: FIN Finding EFFECTIVENESS OF CORRECTIVE ACTION PROGRAM.
Based on the results of this inspection, the NRC concluded that the corrective action program at Kewaunee showed significant weaknesses and inconsistencies across all of the procedural elements inspected. These weaknesses existed across departments and affected multiple cornerstones in the strategic performance areas of Reactor Safety, Radiation Safety, and Safeguards. Of particular note was the lack of procedures for determining the significance of conditions adverse to quality and for trending of issues and the complete lack of trending within your corrective action program. We also identified a lack of urgency in correcting issues which resulted in repeat examples occurring and, coupled with a poor tracking system, a tendency for issues to be dropped. While none of the specific examples identified by the team were of high risk significance
 
4Q/2000 Inspection Findings - Kewaunee                                                                                                Page 11 of 11 when looked at in isolation, in the aggregate they were similar in nature to prior issues in the emergency preparedness area that rose to a higher significance level and contributed to a degraded cornerstone. While we concluded that the station had fostered an environment in which personnel freely identified conditions adverse to quality without fear of discrimination or retaliation, we also concluded that significant weaknesses with, and inconsistent implementation of, the station corrective action program resulted in multiple examples where station personnel did not enter deficiencies into the station's formal corrective action program.
Inspection Report# : 2000019(pdf)
Significance: N/A Sep 29, 2000 Identified By: NRC Item Type: NCV NonCited Violation NO PROCEDURAL GUIDANCE FOR DETERMINING IF CONDITIONS ADVERSE TO QUALITY ARE SIGNIFICANT.
The inspectors reviewed the quality assurance (QA) manual requirements against Kewaunee's implementing procedures and identified that two QA manual requirements were not being implemented. Specifically, Program Requirement 3.1.9 which stated that directives and procedures shall provide for the review of conditions adverse to quality to determine if the conditions are significant in nature. This requirement paralleled 10 CFR Part 50, Appendix B, Criterion XVI, which requires that the cause of significant conditions adverse to quality be determined and corrective actions taken to prevent recurrence. The inspectors reviewed the Nuclear Administrative Directive (NAD 11.08) and the procedure (GNP 11.08.01) governing the KAP (Kewaunee corrective action program) process and found no procedure requirements for identifying significant conditions adverse to quality. This finding does not directly affect a cornerstone. As a result, this issue was not evaluated with the Significance Determination Process and was not assigned a color. One example of a non-cited violation of 10 CFR Part 50, Appendix B, Criterion II, "Quality Assurance Programs," was identified.
Inspection Report# : 2000019(pdf)
Significance: N/A Jun 22, 2000 Identified By: NRC Item Type: FIN Finding CONTROL ROOM OPERATIONS HUMAN PERFORMANCE ISSUES.
The inspectors interviewed operators to evaluate their awareness of degraded control room indications and alarms, and their ability to adequately take manual actions based on degraded alarm functions. The inspectors identified, during interviews, that there was a lack of awareness by operators of a degraded refueling water storage tank low-low level alarm which would be potentially confusing to operators and therefore increase the risk associated with initiating long term sump recirculation.
Inspection Report# : 2000008(pdf)
Significance: N/A Feb 21, 2002 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Implement Required Fire Watch Following Completion of Hot Work Activities A licensee-identified violation was reviewed by the inspectors. Corrective actions taken or planned by the licensee appeared reasonable.
Inspection Report# : 2001017(pdf)
Significance: N/A Aug 24, 2001 Identified By: NRC Item Type: FIN Finding IDENTIFICATION AND RESOLUTION OF PROBLEMS The team concluded that the licensee was generally effective at identifying problems and putting them into the corrective action program. The program itself contained all the necessary attributes of an acceptable corrective action program and was generally successful in correcting identified issues. However, the team noted that, although licensee management had taken efforts to ensure that issues were resolved in accordance with program guidance and requirements, additional efforts appeared necessary to ensure timely resolution of issues. A positive program initiative was the establishment of positions in each of the major plant departments to serve as liaisons between the departments and the corrective action program and to assist with self-assessments. However, examples were identified by the inspectors of problems with the licensee's identification and resolution of problems, prioritization and evaluation of issues, and the effectiveness of corrective actions. Included in these examples were the routine granting of due date extensions for problem evaluation and corrective action implementation, failure to perform a Technical Specification-required biennial surveillance of safety-related procedures, and ineffective corrective actions that resulted in the lack of proper controls over the storage of grease used in safety-related breakers. Based on a review of records and discussions with plant staff, the inspectors concluded that workers at the site felt free to input safety issues into the corrective action program.
Inspection Report# : 2001012(pdf)
Last modified : March 28, 2002
 
1Q/2001 Inspection Findings - Kewaunee                                                                                                      Page 1 of 11 Kewaunee Initiating Events Mitigating Systems Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation FIRE DOOR FUSIBLE LINKS.
On February 20, 2001, the licensee determined that the installed fusible link arrangement on roll-up fire Doors 279 and 281, which separated both trains of service water pumps, would not actuate as designed to ensure that the doors would automatically close to provide a 3-hour fire barrier, contrary to 10 CFR Part 50, Appendix R, Section III.G.2.a which required, in part, separation of cables and equipment of redundant trains by a fire barrier having a 3-hour rating.
Inspection Report# : 2001006(pdf)
Significance:        Mar 12, 2001 Identified By: Licensee Item Type: NCV NonCited Violation INADEQUATE SMOKE DETECTOR COVERAGE IN FIRE ZONE TU-95B.
Licensee identified violation of licensee's operating license that the licensee failed to install a detector in each beam pocket in Fire Zone TU-95B.
Inspection Report# : 2001002(pdf)
Significance:        Feb 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO TEST FIRE DOOR IN ACCORDANCE WITH FIRE PLAN.
The inspectors identified a non-cited violation for failure to properly test a fire door in accordance with the facility's fire protection program plan. The finding was of very low safety significance because, although the fire door separated both trains of service water pumps and did not fully close as designed when subsequently tested, the fire loading in the area was insufficient to credibly impact more than two of the four service water pumps in the area.
Inspection Report# : 2001004(pdf)
Significance: N/A Nov 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO COMPLETE COMPONENT RE-TEST REQUIREMENTS PRIOR TO RETURNING EQUIPMENT TO AN OPERABLE STATUS.
No Color. The inspectors identified a Non-Cited Violation for failure to complete component retest requirements following maintenance performed on the B train control room air conditioner compressor condenser. The unit had been returned to an operable status prior to the retest requirements being completed as prescribed in the associated maintenance procedure. This issue was determined to be a violation of the licensee's Operational Quality Assurance Program Manual, Section 8, "Maintenance Planning and Control." Although the risk associated with this finding was very low and did not affect any cornerstones, the inspectors noted that this finding was similar to previous NRC-identified findings and therefore was of greater than minor significance and warranted documentation. (Section 1R19).
Inspection Report# : 2000020(pdf)
Significance:        Sep 29, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IDENTIFY CORROSION AS POTENTIAL FAILURE MECHANISM.
The inspectors identified that the licensee failed to identify corrosion as a potential failure mechanism in the operability determination for a carbon
 
1Q/2001 Inspection Findings - Kewaunee                                                                                              Page 2 of 11 steel key in the service water system. Thus, the licensee failed to quantify the corrosion rate and therefore did not adequately evaluate the expected service life of the carbon steel key. One non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was identified.
Inspection Report# : 2000019(pdf)
Significance:        Sep 29, 2000 Identified By: NRC Item Type: NCV NonCited Violation PRESSURE RATING OF AUXILIARY FEEDWATER STEAM TRAPS.
The inspectors identified that a root cause evaluation for a 1996 equipment issue in the turbine-driven auxiliary feedwater system was not completed until 1999. The evaluation stated that the internals of the steam traps were designed to operate at pressures up to a maximum 600 psig but that the traps were exposed to pressures up to 1025 psig. A corrective action item to initiate a design change request to replace the steam traps with a different model rated for the design pressure of the system was described in the evaluation. However, the inspectors identified that the design change request had never been initiated and the KAP ( Kewaunee corrective action document) had been closed. As a result, the corrective action item for this design problem was lost. In addition, operability of the system had never been formally evaluated despite the identification that the system design requirements were not met. The licensee subsequently determined that the steam traps remained operable and was planning to initiate the design change to correct the problem. One non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was identified.
Inspection Report# : 2000019(pdf)
Significance:        Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation DESIGN CONTROL VIOLATION FOR AUXILIARY FEEDWATER STRAINER MESH SIZE.
The inspectors questioned the mesh size of the strainers, which were installed in the suction of the three auxiliary feedwater (AFW) pumps. As a result of the inspectors' questions, license personnel inspected the strainers on August 21, 2000, and found the strainers to have 1/16 inch openings. A note was later found on Figure 10.2-3 of the UFSAR that indicated that the AFW suction strainer size was 1/8 inch. The smaller openings would not support the use of service water as a safety related source for AFW and as a result all three trains of AFW were declared inoperable. This condition had apparently existed for approximately 25 years and was identified as a non-cited violation of Criterion III, "Design Control," of 10 CFR 50, Appendix B.
Inspection Report# : 2000012(pdf)
Significance: N/A Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation RETRIEVAL OF SERVICE WATER SYSTEM DESIGN INFORMATION.
In many cases, design basis information for the service water system was difficult if not impossible to locate. Licensee personnel wrote KAP WO 00-002566 to enter the problem in the corrective action program.
Inspection Report# : 2000012(pdf)
Significance:        Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PROPERLY CONTROL CORRECTED TEST DATA NECESSARY FOR DESIGN CALCULATIONS.
An example of a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," was identified in the handling of service water system flow test data, which was subsequently used in calculations. Gauge readings corrected for post test calibration checks, gauge reading corrections for elevation considerations, and flow values corrected for pump degradation were contained in spreadsheets in the possession of an individual staff member, but not currently packaged with raw test data, and not bearing evidence of a formal review and control process. The connection between the test data, which had been vaulted, and the values used in the calculation, could not be made without use of the uncontrolled spreadsheet.
Inspection Report# : 2000012(pdf)
Significance:        Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation IMPROPER DESIGN CALCULATION IDENTIFICATION, NON-CONSERVATIVE ASSUMPTIONS, CALCULATION ERRORS, AND DUPLICATE CALCULATIONS.
 
1Q/2001 Inspection Findings - Kewaunee                                                                                                  Page 3 of 11 An example of a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," was identified because of inadequate control of design calculations. The control failures included improper identification of calculations, non-conservative assumptions, calculation errors, and duplicate or superceded calculations not properly identified or canceled. The failure to follow the established design control process increased the potential for errors in the design and operation of the service water system. Because the system was subsequently demonstrated to be capable of removing the design heat load, the actual significance was low and this finding screened out as having very low risk significance.
Inspection Report# : 2000012(pdf)
Significance:          Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO DOCUMENT IN THE CORRECTIVE ACTION PROGRAM THE USE OF INCORRECT MATERIAL IN A SERVICE WATER PUMP KEY.
An example of a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was identified because of inadequate corrective action to correct an incorrect coupling adjust nut set screw and a low strength "soft" key material, which had contributed to a pump shaft failure. Licensee personnel had known of the "soft" key material since July 21, 1999. The "soft" key material was found in other service water pumps but had not been removed from all pumps. As of July 25, 2000, licensee personnel had not documented the existence of the "soft" key material in the corrective action program.
Inspection Report# : 2000012(pdf)
Significance:          Jul 07, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO COMPLETE COMPONENT RETEST REQUIREMENTS IN ACCORDANCE WITH PROCEDURE.
On June 26, 2000, during a review of post maintenance testing requirements following maintenance performed on the control room post accident system charcoal filter heat detector, the inspectors identified that maintenance technicians had not completed the component re-test requirements, as required by a preventative maintenance procedure prior to the system being returned to an operable status. On July 7, the inspectors identified a second example of failing to complete component re-test requirements following maintenance on the zone special ventilation system charcoal filter heat detector as required. The issue was considered to be of very low safety significance based on the determination that although the licensee had not completed all of the component retest requirements prior to returning the equipment to service, the subsequent testing determined that the equipment was in an operable status. The failure to complete the component retest requirements in accordance with site procedures was identified as a Non-Cited Violation.
Inspection Report# : 2000014(pdf)
Significance:          Jun 22, 2000 Identified By: NRC Item Type: FIN Finding FAILURE TO MEET SINGLE FAILURE CRITERIA FOR RESIDUAL HEAT REMOVAL VALVE CIRCUITRY.
The licensee identified that the circuitry associated with the residual heat removal system discharge to safety injection system suction isolation valves did not meet single failure criteria. The inspectors noted that this design requirement was identified in the facility's updated safety analysis report. The licensee subsequently implemented a temporary change to the facility. The inspectors reviewed the issue and identified that the facility had been operating outside of its design basis, which was reportable to the NRC. The licensee subsequently made a one hour non-emergency report to the NRC. Since there was no actual loss of safety function to the system, this issue was screened as very low risk significance.
Inspection Report# : 2000008(pdf)
Significance:          Jun 22, 2000 Identified By: NRC Item Type: FIN Finding MAINTENANCE RULE FAILURES ASSOCIATED WITH REACTOR HEAD VENT VALVE.
The inspectors reviewed the licensee's implementation of the maintenance rule for failures associated with a reactor head vent valve. The licensee's corrective action documents identified a potential maintenance rule functional failure but the completed evaluation of the problem did not document the final determination. However, the inspectors identified that the repeated failures may have been prevented if maintenance activities such as valve disassembly and cleaning had been performed. In this case, maintenance rule reliability goals were not exceeded. The licensee had documented similar maintenance rule program deficiencies and developed a corrective action program to address the deficiencies. Although programmatic deficiencies exist, since no maintenance rule reliability criteria had been exceeded, this issue was considered of very low risk significance.
Inspection Report# : 2000008(pdf)
 
1Q/2001 Inspection Findings - Kewaunee                                                                                                    Page 4 of 11 Significance:        Jun 22, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO INITIATE KEWAUNEE ASSESSMENT PROCESS DOCUMENT REGARDING REFUELING WATER STORAGE TANK LOW-LOW LEVEL ALARM INOPERABILITY.
The inspectors identified that the refueling water storage tank low-low level alarm which was actuating five percent higher than normal had not been documented in a Kewaunee Assessment Process form by the licensee, and therefore had not received an operability evaluation. This failure was identified as contrary to site administrative procedures. Following the licensee's documentation of the problem, the inspectors identified that the associated operability evaluation considered the acceptability of an operator workaround to address the issue, but did not address any safety implications or consequences of the alarm actuating early. A subsequent operability evaluation by the licensee was evaluated as adequate by the inspectors. Since the subsequent operability evaluation was adequate and it was determined that no safety mitigation equipment was adversely affected by the early actuation of the alarm, this issue was considered of very low risk significance. A non-cited violation (NCV) was identified for failing to document a non-conforming condition, contrary to site administrative procedure requirements.
Inspection Report# : 2000008(pdf)
Significance:        Jun 22, 2000 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO TEST ADDITIONAL RELIEF VALVES IN ACCORDANCE WITH TECHINCAL SPECIFICATIONS.
The licensee identified that the suction relief valve for an auxiliary feedwater pump may have failed its relief test criteria, but did not process the documented deficiency until several weeks later. The licensee then expanded the scope of the relief testing to the suction relief valves associated with the other auxiliary feedwater pumps to meet technical specification requirements. Since any one train of auxiliary feedwater was capable of supplying 100 percent of the decay heat removal requirements, this issue was screened as very low risk significance. However, the time delay in complying with technical specification requirements for testing other relief valves was identified as an NCV.
Inspection Report# : 2000008(pdf)
Significance:        May 22, 2000 Identified By: NRC Item Type: FIN Finding FIRE EXTINGUISHERS NOT LOCATED IN ALL AREAS OF CONTAINMENT BUILDING AT BEGINNING OF REFUELING OUTAGE.
During a walkdown of the containment building, the inspectors identified that portable fire extinguishers were not located in the containment basement at the beginning of the plant refueling outage. Additionally, site fire protection procedures required that responsible fire protection personnel perform inspections of selected plant areas to ensure that the quantity of combustible material was minimized. However, the procedure did not list the containment as an area to be inspected and the procedures did not require the placement and location of portable fire suppression equipment inside containment during the refueling outage. Due to a low number of work activities ongoing at the time, this issue was screened as Green (very low risk significance).
Inspection Report# : 2000007(pdf)
Significance:        May 22, 2000 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO INSTALL RAYCHEM HEAT SHRINK MATERIAL IN ACCORDANCE WITH PROCEDURE REQUIREMENTS.
The licensee identified that two Raychem electrical cable splices utilized in environmentally qualified (EQ) safety-related equipment had not been installed in accordance with EQ requirements. These splices were associated with pressurizer level transmitters and were installed in 1984. The licensee subsequently performed extensive EQ testing of the splices to determine the qualification of the splices' as-found configurations. Test results indicated that the splices would have been able to perform their intended function in a harsh environment inside containment. This issue was considered to be of low safety significance based on the successful EQ testing of the as-found splices' configurations and was screened as Green (very low risk significance). One non-cited violation was identified.
Inspection Report# : 2000007(pdf)
Significance:        Feb 21, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform Thorough 10 CFR 50.59 Safety Evaluation The inspectors identified a Non-Cited Violation for failure to perform an adequate 10 CFR 50.59 safety evaluation associated with emergency
 
1Q/2001 Inspection Findings - Kewaunee                                                                                                  Page 5 of 11 operating procedure changes to address component cooling water pump dead-head operational concerns. The safety evaluation did not evaluate the potential for initiating a loss-of-coolant accident via the reactor coolant loop seals during conditions of a complete loss of component cooling water.
Inspection Report# : 2001017(pdf)
Significance: N/A Sep 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ESTABLISH CONTINGENCY PLANS FOR ORANGE RISK CONDITION The inspectors identified the failure to establish contingency plans during a planned high risk plant configuration. Contrary to administrative requirements, the licensee approved an orange risk condition during a refueling outage with no contingency plans to mitigate the consequences of a loss of spent fuel pool cooling with a full core offload in the pool. A Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was identified. The finding was of very low safety significance because although the licensee had not approved appropriate contingency actions for the orange risk condition, the licensee subsequently rescheduled the planned maintenance to eliminate the orange risk condition.
Inspection Report# : 2001013(pdf)
Significance: N/A Aug 24, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO CONDUCT BIENNIAL SURVIELLANCE OF SAFETY-RELATED PROCEDURES PER TS 6.8.c A Non-Cited Violation of Technical Specification 6.8.c was identified for the failure to perform a biennial surveillance of safety-related procedues.
This issue was more than minor because if left uncorrected, could under the same condition become a more significant safety concern. However, since no specific cornerstone had been impacted, this finding is designated as No Color.
Inspection Report# : 2001012(pdf)
Significance:          Aug 24, 2001 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE CORRECTIVE ACTIONS FOR PREVIOUS PROBLEM WITH THE CONTROL OF SAFETY-RELATED MATERIALS A Non-Cited Violation of Criterion XVI, "Corrective Action," of Appendix B of 10 CFR Part 50 was identified for ineffective corrective actions for a problem with the control of the storage of consumable materials, such as thread sealant, used in safety-related applications. These ineffective actions subsequently resulted in the inadequate control of the storage of grease used in safety-related breakers. This issue was more than minor because if left uncorrected, could under the same condition become a more significant safety concern. In that this issue could credibly affect the operability, availability, reliability, or function of a system or train in a mitigating system, it is a Green finding.
Inspection Report# : 2001012(pdf)
Significance:          Aug 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation NON-RATED FIRE BARRIER A Non-Cited Violation [of 10 CFR Part 50, Appendix R, Section III.G.2.a] was identified for failure to provide a 3-hour rated fire barrier to separate redundant trains of safe shutdown equipment. This finding was of very low safety significance because the licensee tested a replica of the fire barrier and demonstrated that the fire barrier provided protection for at least 60 minutes, which was sufficient for the hazards in the area.
Inspection Report# : 2001011(pdf)
Significance:          Jul 20, 2001 Identified By: NRC Item Type: FIN Finding LICENSED OPERATOR REQUALIFICATION EXAMINATION RESULTS The inspectors identified that two of eight crews examined during the licensee's calendar year 2001 licensed operator requalification operating test had failed. The finding was of very low safety significance because both crews that had failed received remedial training prior to being returned to shift, and the results of the licensee's operator licensing requalification operating test given in calendar year 2000 indicated that only one crew, out of a total of eight crews tested, had failed.
Inspection Report# : 2001011(pdf)
Significance: N/A Jun 30, 2001 Identified By: NRC Item Type: FIN Finding
 
1Q/2001 Inspection Findings - Kewaunee                                                                                                  Page 6 of 11 FAILURE TO BALANCE RELIABILITY AND AVAILABILITY AS REQUIRED BY 10 CFR 50.65(a)(3)
The inspectors identified a failure to evaluate whether adjustments were necessary such that there would be an appropriate balance between systems' availability and reliability in accordance with 10 CFR 50.65(a)(3) of the maintenance rule. The inspectors identified that the licensee did not have an administrative process to track maintenance rule functional failures and maintenance preventible maintenance functional failures. As a result, reliability and availability could not be balanced as required by the Maintenance Rule periodic evaluation. The safety significance of the specific finding was very low because it did not affect the operability of the systems, and the licensee entered the finding in the corrective action program. However, this finding was considered to be of regulatory concern in the area of maintenance rule implementation due to the extent of the problems with the Maintenance Rule Program.
Inspection Report# : 2001009(pdf)
Significance:          Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IDENTIFY DEFICIENT CONDITION OF VALVE AFW-1B The inspectors identified that the licensee failed to promptly identify and correct the B' train auxiliary feedwater pump discharge check valve which was stuck in an intermediate position. A Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was issued. The finding was of very low safety significance because, although the check valve was stuck in an intermediate position, the time that it was known to have been stuck was less than the technical specification allowed outage time for one train of auxiliary feedwater to be out of service (less than 72 hours). Additionally, the other two trains of auxiliary feedwater were each capable of 100 percent decay heat removal.
Inspection Report# : 2001009(pdf)
Significance:          Jun 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO TRACK UNAVAILABILITY OF SYSTEMS REQUIRED DURING SHUTDOWN OPERATION 10 CFR 50.65(a)(1), required, in part, that the licensee monitor the performance or condition of SSCs within the scope of the rule as defined by 10 CFR 50.65(b), against licensee-established goals, in a manner sufficient to provide reasonable assurance that such SSCs are capable of fulfilling their intended functions. 10 CFR 50.65(a)(2) stated, in part, that monitoring as specified in 10 CFR 50.65(a)(1) was not required where it had been demonstrated that the performance or condition of an SSC was being effectively controlled through the performance of appropriate preventive maintenance, such that the SSC remained capable of performing its intended function. Contrary to the above, the licensee failed to demonstrate that the performance or condition of systems required to be available during shutdown conditions and within the scope of the rule had been effectively controlled through the performance of appropriate preventive maintenance and did not monitor against licensee-established goals.
Specifically, the licensee failed to monitor the unavailability of systems required during shutdown operation.
Inspection Report# : 2001009(pdf)
Significance:          Jun 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO ESTABLISH MAINTENANCE RULE (a)(1) GOALS 10 CFR 50.65(a)(1), requires, in part, that the licensee monitor the performance or condition of SSCs within the scope of the rule as defined by 10 CFR 50.65(b), against licensee-established goals, in a manner sufficient to provide reasonable assurance that such structures, systems, and components, are capable of fulfilling their intended functions. Such goals shall be established commensurate with safety. When the performance or condition of a structure, system, or component does not meet established goals, appropriate corrective action shall be taken. Contrary to the above, from 1996, the licensee did not take appropriate corrective actions when the performance of those systems in (a)(1) did not meet licensee established goals. Specifically, the licensee determined timely and appropriate corrective actions had not been taken for five systems that had been in (a)(1) category for approximately 3 years to 5 years: component cooling (entered (a)(1) on April 23, 1997), control room air conditioning (July 24, 1996), station and instrument air (July 3, 1997), auxiliary building air ventilation (July 31, 1997), and control rod drive (August 6, 1998). This issue is in the licensee's corrective action system as KAP WO 01-3323. The inspectors evaluated the risk significance of this issue using the Significance Determination Process. The inspectors did not identify where this failure resulted in a total loss of a risk significant SSC. Therefore, this issue was screened as Green (very low risk significance) after a Phase 1 Significance Determination Process review. Although the risk significance of this issue was low, the inspectors concluded that this was more than a minor concern because the failure to recognize and correct ineffective maintenance practices resulted in risk significant systems in (a)(1) for years with no improvement in performance. The NRC tracking number for this issue is 50-305/01-09-02.
Inspection Report# : 2001009(pdf)
Barrier Integrity
 
1Q/2001 Inspection Findings - Kewaunee                                                                                                  Page 7 of 11 Significance:        Jun 22, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation FAILURE TO UPDATE COMPUTER ALARM FOR CURRENT AXIAL FLUX DISTRIBUTION TARGET BAND.
The licensee identified, following plant startup, that a computer alarm had not been updated properly to alarm if axial flux distribution deviated outside of the flux distribution target band. This condition was contrary to technical specification requirements. The licensee reviewed the axial flux distribution history since the startup and determined that at no time was the flux distribution outside of the target band. Since the axial flux distribution was never outside of the target band, this issue was screened as very low risk significance. An NCV was identified for failing to comply with technical specification requirements for monitoring axial flux distributions.
Inspection Report# : 2000008(pdf)
Significance:        May 22, 2000 Identified By: Licensee Item Type: FIN Finding TECHNICAL SPECIFICATIONS INTERPRETATION FOR TESTING REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES.
The NRC determined that the licensee's practice of testing reactor coolant system pressure isolation Valve SI-22B prior to entering the cold shutdown condition was contrary to Technical Specification requirements 4.2.a.3.a. Technical Specification 4.2.a.3.a required that periodic leakage testing of Valve SI-22B be accomplished prior to reaching operating mode after the plant was placed in cold shutdown. This issue was considered to be of low safety significance because of a subsequent successful valve test and was screened as Green (very low risk significance).
Enforcement discretion was applied to this item in accordance with Section VII.B.6 of the Enforcement Policy.
Inspection Report# : 2000007(pdf)
Significance:        May 19, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PROPERLY IMPLEMENT FLAW ACCEPTANCE CRITERIA FOR SLEEVE WELD INSPECTIONS.
During 1998 inservice inspection examinations, the licensee failed to properly implement the flaw acceptance criteria for laser welded sleeve inspection within two steam generator tubes. The safety significance was very low based on the absence of adverse consequences, and May 2000 in-situ pressure testing where both welds exhibited zero leakage at normal operating pressure, main steam line break pressure, and three times normal operating differential pressures. As such, this issue was characterized as a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion IX, Control of Special Processes. To correct the error, the licensee plugged both tubes (Section 1RO8).
Inspection Report# : 2000011(pdf)
Emergency Preparedness Significance:        Mar 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation SUPPLEMENTAL INSPECTION OF WHITE ERO AUGMENTATION FINDING AND RESULTING GREEN FINDING.
This supplemental inspection was performed by the NRC to evaluate the licensees evaluation associated with the failure to conduct successful quarterly, off-hours, unannounced staff augmentation drills during the second, third, and fourth quarters of 1999 and the second quarter of 2000.
This performance issue was previously characterized as having low to moderate risk significance (White) in NRC Inspection Report No. 50-305/2000015(DRS). During this supplemental inspection, performed in accordance with Inspection Procedure95002, the inspector concluded that the licensee performed a comprehensive evaluation of the unsuccessful staff augmentation drills. The licensees evaluation identifiedtwo root causes which resulted in the unsuccessful drills and in the staffs inability tocorrect thedeficiencies: (1) Management has not effectively acted to provide increased depth and flexibility in the emergency response organization following a reduction in staffing several years ago; and (2)
Management has accepted an adverse trend of test failures without requiring investigation into the root causes. The inspector reviewed the licensees corrective actions, both completed and planned, and concluded that the programmatic corrective actions appeared to address the identified root causes. In particular, the licensee assigned certain positions to an on-call rotation to ensure personnel were capable of augmenting in a timely manner, and the licensee was progressing in training additional staff to increase the depth of personnel assigned to key emergency response positions. In addition, the licensee was continuing its efforts in improving its corrective action program. The inspector reviewed the licensees immediate response to the issue and identified that one of the licensees initial corrective actions resulted in a Non-Cited Violation of regulatory requirements. To obtain a timely response of a key emergency response position (severe accident management - core hydraulics), the emergency preparedness staff effectively changed the emergency plan without revising the necessary procedures and without formally assessing the impact of that change. The staff instructed and trained personnel to respond to a location other than the Technical Support Center, which was
 
1Q/2001 Inspection Findings - Kewaunee                                                                                                    Page 8 of 11 contrary to the licensees current implementing procedures. While this change enabled the licensee to augment its staff in a timely manner, the change was not performed in accordance with NRC requirements. In order to make such a change, the licensees emergency plan required that the change be formally assessed to ensure that it did not reduce the effectiveness of the plan or any other implementing procedure. Since the issue did not result in a failure to meet an emergency preparedness planning standard, the failure to adequately implement the emergency plan was determined to be a violation of very low safety significance (Green) (Section 02.3(a)). Due to the licensees acceptable performance in assessing the emergency response augmentation drill deficiencies, the White finding associated with this issue will only be considered in assessing plant performance for a total of four quarters in accordance with the guidance in IMC 0305, Operating Reactor Assessment Program. Implementation of the licensees corrective actions will be reviewed during a future inspection.
Inspection Report# : 2001007(pdf)
Significance:          Mar 30, 2001 Identified By: NRC Item Type: VIO Violation FAILURE TO CORRECT SELF-IDENTIFIED ERO AUGMENTATION DRILL DEFICIENCIES.
During a baseline inspection of the emergency preparedness program conducted on August 14 - September 21, 2000, the NRC identified a preliminary White issue and potential violation for the licensees failure to successfully correct deficiencies identified during staff augmentation drills and to demonstrate timely staff augmentation in 1999 and 2000. The issue was unresolved pending the outcome of the NRCs final significance determination. On January 30, 2001, the NRC conducted a regulatory conference with the licensee and subsequently issued the licensee a White finding and Notice of Violation (Enforcement Action No. 00-214) associated with the performance issue.
Inspection Report# : 2001007(pdf)
Significance:          Mar 23, 2001 Identified By: NRC Item Type: NCV NonCited Violation POST-ACCIDENT SAMPLING SYSTEM CONTAINMENT AIR SAMPLE PANEL TESTING.
A Non-Cited Violation of Technical Specification 6.14 was identified for the failure to implement a program that ensured the capability to obtain and analyze containment atmosphere samples under accident samples using the cantainment air sampling panel (CASP). Although the CASP was installed, as was indicated in the emergency plan, the licensee had neither developed procedures nor had tested its capability to obtain a containment atmosphere sample using the CASP. The licensee could not recall if and when containment air samples were last obtained using the CASP.
Inspection Report# : 2001006(pdf)
Significance: N/A Feb 27, 2001 Identified By: NRC Item Type: FIN Finding SUPPLEMENTAL INSPECTION OF YELLOW ANS PERFORMANCE INDICATOR AND ASSOCIATED CORRECTIVE ACTION PROGRAM DEFICIENCIES.
This supplemental inspection was performed by the NRC to assess the licensee's evaluation associated with a Yellow performance indicator for the Alert and Notification System (ANS) and the associated Yellow finding related to the licensee's corrective action program. During this supplemental inspection, performed in accordance with Inspection Procedure 95002, the inspectors concluded that the licensee performed comprehensive evaluations of the performance problems associated with the ANS and its corrective action program. These evaluations identified primary root causes and contributing causes for both issues. Along with the electronics/hardware problems, the licensee identified the primary root causes for the ANS to be the failure to make changes to the system via a change control process and the failure of the activation procedure to provide for alternate/backup activation methods and to provide clear success criteria. In the case of the corrective action program, the licensee concluded that the primary root causes were plant management's inadequate risk evaluation regarding decisions affecting the corrective action program and the development of a culture in the licensee's organization, which minimized the importance of information from outside organizations. In particular, the licensee focused on low-cost power operation and failed to recognize the value of changes and improvements which had occurred throughout the industry, such as the value and expansion of the corrective action program. In the case of the Yellow ANS performance indicator, the licensee had completed several significant corrective actions to address the root causes and contributing causes identified in its evaluation. The inspectors found that the corrective actions appeared appropriate to address the underlying root causes and that ANS testing data indicated an improving trend in the NRC performance indicator. The licensee also performed comprehensive assessments of the emergency preparedness program, quality assurance program, plant operations, and other plant programs to determine the extent of condition (re. the root causes described above). Based on these evaluations, the licensee began to implement significant actions to correct the deficiencies in the corrective action program and other weaknesses identified. Generally, the inspectors observed progress in the licensee's initial implementation of these corrective actions. Due to the licensee's acceptable performance in assessing the Yellow ANS performance indicator and the associated Yellow finding, the Yellow finding will not be considered in assessing future plant performance.
Inspection Report# : 2001005(pdf)
Significance:          Sep 21, 2000 Identified By: NRC
 
1Q/2001 Inspection Findings - Kewaunee                                                                                                  Page 9 of 11 Item Type: FIN Finding INADEQUATE ROOT CAUSE EVALUATION FOR YELLOW ALERT AND NOTIFICATION SYSTEM PERFORMANCE INDICATOR.
The licensee's evaluation of the Yellow Alert and Notification (siren) System Performance Indicator (PI) was inadequate. The inspector concluded that the licensee's evaluation was not performed at the depth necessary to identify the root causes of the siren performance problems and, instead, only identified the symptoms of the root causes. Specifically, the inspector identified the following substantive weaknesses in the licensee's evaluation of the siren system performance, which appeared to result from systemic corrective action program deficiencies within this cornerstone:
* The licensee's evaluation was not of sufficient depth to clearly identify the root causes associated with the decline in siren system performance.
* Licensee management did not provide well-understood and clear guidance/expectations for performing root cause evaluations.
* The licensee's evaluation of the quality assurance program was narrowly focused and was not critical of its role in failing to identify and correct the siren performance problems.
* The licensee did not establish a priority for each of the long-term corrective actions in accordance with the associated significance or risk.
* The licensee did not have any formal provisions for measuring the effectiveness of its corrective actions.
* Within the licensee's evaluation, the licensee had not evaluated common causes or the extent of the condition. Due to the corrective action program performance deficiencies within this cornerstone, we have been unable to conclude that the performance issues that resulted in the yellow PI have been addressed. Therefore, we are issuing a yellow finding that corresponds to the original issues that resulted in a yellow PI. Additional inspection effort will be focused on the licensee's further evaluation of the siren reliability root causes and the continuing corrective action program implementation deficiencies identified during this inspection.
Inspection Report# : 2000017(pdf)
Significance: N/A Apr 05, 2000 Identified By: NRC Item Type: FIN Finding LICENSEE FAILED TO IDENTIFY THE FULL SCOPE OF PROBLEMS WITH THE ALERT AND NOTIFICATION SYSTEM PERFORMANCE.
The inspectors concluded that the licensee's assessment was not sufficiently comprehensive to identify the full scope of problems associated with the Alert and Notification System (ANS) performance program. As a result, licensee corrective actions generally were focused on the equipment problem rather than all root causes and contributing factors. For example, the inspectors identified that: (1) management oversight of the ANS performance program was limited; (2) an audit failed to identify degrading ANS performance as a concern; (3) annual preventive maintenance was not consistently performed on the system; (4) the corrective action program was not used consistently to document ANS problems; and (5) maintenance procedures and records were deficient. Collectively, these problems indicate that the ANS performance program lacked sufficient structure and oversight.
Inspection Report# : 2000006(pdf)
Occupational Radiation Safety Significance:        May 19, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO POST A VERY HIGH RADIATION AREA.
The inspectors identified a noncited violation for the failure to post a very high radiation area in accordance with 10 CFR 20.1902(c). Although the area was not adequately posted, the licensee had provided physical controls and barriers that were consistent with its requirements for a very high radiation area. Based on the adequacy of these controls, the potential for an overexposure from the inadvertent entry of personnel into the area was low. Consequently, this finding was determined to be of very low safety significance (Section 20S1.1).
Inspection Report# : 2000009(pdf)
Significance:        Oct 02, 2001 Identified By: NRC Item Type: NCV NonCited Violation HIGH RADIATION AREA ACCESS CONTROLS Non-Cited Violation of Technical Specification 6.13 and an associated Green Finding for failure to 'barricade' three ladders that provided entry to high radiation areas (less than 1000 mrem/hour) located on the steam generator/pressurizer platforms.
Inspection Report# : 2001014(pdf)
Public Radiation Safety
 
1Q/2001 Inspection Findings - Kewaunee                                                                                                Page 10 of 11 Physical Protection Significance: N/A Jul 21, 2000 Identified By: NRC Item Type: FIN Finding CORRECTIVE ACTION ON SEARCH ISSUES DID NOT WORK.
The inspector determined that the licensee's effectiveness of implemented corrective actions for a previously identified inspection finding regarding an inadequate vehicle search was not totally effective in preventing recurrence. Previous corrective action was not adequately focused (Section 3PP2.2).
Inspection Report# : 2000013(pdf)
Significance:        Jul 21, 2000 Identified By: NRC Item Type: NCV NonCited Violation SEARCHES INADEQUATE (REPEAT).
The inspector identified a Non-Cited violation by observing that, a security officer failed to search an easily accessible compartment on one vehicle.
The failure resulted from human error because the officer did not observe the access panel to the compartment during the vehicle search process.
When searched, no prohibited items were found. Corrective actions were implemented. The inspector determined that the licensee's effectiveness of implemented corrective actions for a previously identified inspection finding regarding an inadequate vehicle search was not totally effective in preventing recurrence. Previous corrective action was not adequately focused.
Inspection Report# : 2000013(pdf)
Miscellaneous Significance: N/A Nov 09, 2000 Identified By: NRC Item Type: FIN Finding FAILURE TO TAKE ADEQUATE CORRECTIVE ACTIONS TO ENSURE COMPONENT RE-TEST REQUIREMENTS COMPLETED ACCORDING TO MAINTENANCE PROCEDURES.
No Color. The inspectors determined that a negative performance trend had developed in the licensee's ability to identify and promptly take appropriate corrective actions to prevent recurrence based on two previously identified examples (NCV 50-305/2000014-01) and one example identified during this inspection period (NCV 50-305/00-20-01). All three examples related to the licensee returning safety-related equipment to service prior to completing all required post-maintenance retesting. While the risk of the individual examples was very low, the licensee had failed to ensure that all retest requirements had been completed before returning safety-related equipment to service. These findings collectively indicated a problem with the licensee's ability to provide timely and adequate corrective actions to prevent recurrence. (Section 4OA2).
Inspection Report# : 2000020(pdf)
Significance: N/A Sep 29, 2000 Identified By: NRC Item Type: FIN Finding EFFECTIVENESS OF CORRECTIVE ACTION PROGRAM.
Based on the results of this inspection, the NRC concluded that the corrective action program at Kewaunee showed significant weaknesses and inconsistencies across all of the procedural elements inspected. These weaknesses existed across departments and affected multiple cornerstones in the strategic performance areas of Reactor Safety, Radiation Safety, and Safeguards. Of particular note was the lack of procedures for determining the significance of conditions adverse to quality and for trending of issues and the complete lack of trending within your corrective action program. We also identified a lack of urgency in correcting issues which resulted in repeat examples occurring and, coupled with a poor tracking system, a tendency for issues to be dropped. While none of the specific examples identified by the team were of high risk significance when looked at in isolation, in the aggregate they were similar in nature to prior issues in the emergency preparedness area that rose to a higher significance level and contributed to a degraded cornerstone. While we concluded that the station had fostered an environment in which personnel freely identified conditions adverse to quality without fear of discrimination or retaliation, we also concluded that significant weaknesses with, and inconsistent implementation of, the station corrective action program resulted in multiple examples where station personnel did not enter deficiencies into the station's formal corrective action program.
Inspection Report# : 2000019(pdf)
Significance: N/A Sep 29, 2000 Identified By: NRC Item Type: NCV NonCited Violation NO PROCEDURAL GUIDANCE FOR DETERMINING IF CONDITIONS ADVERSE TO QUALITY ARE SIGNIFICANT.
The inspectors reviewed the quality assurance (QA) manual requirements against Kewaunee's implementing procedures and identified that two QA manual requirements were not being implemented. Specifically, Program Requirement 3.1.9 which stated that directives and procedures shall
 
1Q/2001 Inspection Findings - Kewaunee                                                                                                Page 11 of 11 provide for the review of conditions adverse to quality to determine if the conditions are significant in nature. This requirement paralleled 10 CFR Part 50, Appendix B, Criterion XVI, which requires that the cause of significant conditions adverse to quality be determined and corrective actions taken to prevent recurrence. The inspectors reviewed the Nuclear Administrative Directive (NAD 11.08) and the procedure (GNP 11.08.01) governing the KAP (Kewaunee corrective action program) process and found no procedure requirements for identifying significant conditions adverse to quality. This finding does not directly affect a cornerstone. As a result, this issue was not evaluated with the Significance Determination Process and was not assigned a color. One example of a non-cited violation of 10 CFR Part 50, Appendix B, Criterion II, "Quality Assurance Programs," was identified.
Inspection Report# : 2000019(pdf)
Significance: N/A Sep 29, 2000 Identified By: NRC Item Type: NCV NonCited Violation NO PROCEDURAL GUIDANCE FOR TRENDING CONDITIONS ADVERSE TO QUALITY.
The inspectors reviewed the quality assurance (QA) manual requirements against Kewaunee's implementing procedures and identified that two QA manual requirements were not being implemented. Specifically, QA Program Requirement, 3.1.10, stated that directives and procedures shall provide for analyzing trends of conditions adverse to quality. Once identified these trends were required to be considered significant conditions adverse to quality. The inspectors found that conditions adverse to quality were not defined in the KAP (Kewaunee corrective action program) procedures and that no procedure existed for trending. This finding does not directly affect a cornerstone. As a result, this issue was not evaluated with the Significance Determination Process and was not assigned a color. One example of a non-cited violation of 10 CFR Part 50, Appendix B, Criterion II, "Quality Assurance Program," was identified.
Inspection Report# : 2000019(pdf)
Significance: N/A Jun 22, 2000 Identified By: NRC Item Type: FIN Finding CONTROL ROOM OPERATIONS HUMAN PERFORMANCE ISSUES.
The inspectors interviewed operators to evaluate their awareness of degraded control room indications and alarms, and their ability to adequately take manual actions based on degraded alarm functions. The inspectors identified, during interviews, that there was a lack of awareness by operators of a degraded refueling water storage tank low-low level alarm which would be potentially confusing to operators and therefore increase the risk associated with initiating long term sump recirculation.
Inspection Report# : 2000008(pdf)
Significance: N/A Feb 21, 2002 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Implement Required Fire Watch Following Completion of Hot Work Activities A licensee-identified violation was reviewed by the inspectors. Corrective actions taken or planned by the licensee appeared reasonable.
Inspection Report# : 2001017(pdf)
Significance: N/A Aug 24, 2001 Identified By: NRC Item Type: FIN Finding IDENTIFICATION AND RESOLUTION OF PROBLEMS The team concluded that the licensee was generally effective at identifying problems and putting them into the corrective action program. The program itself contained all the necessary attributes of an acceptable corrective action program and was generally successful in correcting identified issues. However, the team noted that, although licensee management had taken efforts to ensure that issues were resolved in accordance with program guidance and requirements, additional efforts appeared necessary to ensure timely resolution of issues. A positive program initiative was the establishment of positions in each of the major plant departments to serve as liaisons between the departments and the corrective action program and to assist with self-assessments. However, examples were identified by the inspectors of problems with the licensee's identification and resolution of problems, prioritization and evaluation of issues, and the effectiveness of corrective actions. Included in these examples were the routine granting of due date extensions for problem evaluation and corrective action implementation, failure to perform a Technical Specification-required biennial surveillance of safety-related procedures, and ineffective corrective actions that resulted in the lack of proper controls over the storage of grease used in safety-related breakers. Based on a review of records and discussions with plant staff, the inspectors concluded that workers at the site felt free to input safety issues into the corrective action program.
Inspection Report# : 2001012(pdf)
Last modified : March 28, 2002
 
2Q/2001 Inspection Findings - Kewaunee                                                                                                  Page 1 of 11 Kewaunee Initiating Events Mitigating Systems Significance:          Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IDENTIFY DEFICIENT CONDITION OF VALVE AFW-1B The inspectors identified that the licensee failed to promptly identify and correct the B' train auxiliary feedwater pump discharge check valve which was stuck in an intermediate position. A Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was issued. The finding was of very low safety significance because, although the check valve was stuck in an intermediate position, the time that it was known to have been stuck was less than the technical specification allowed outage time for one train of auxiliary feedwater to be out of service (less than 72 hours). Additionally, the other two trains of auxiliary feedwater were each capable of 100 percent decay heat removal.
Inspection Report# : 2001009(pdf)
Significance: N/A Jun 30, 2001 Identified By: NRC Item Type: FIN Finding FAILURE TO BALANCE RELIABILITY AND AVAILABILITY AS REQUIRED BY 10 CFR 50.65(a)(3)
The inspectors identified a failure to evaluate whether adjustments were necessary such that there would be an appropriate balance between systems' availability and reliability in accordance with 10 CFR 50.65(a)(3) of the maintenance rule. The inspectors identified that the licensee did not have an administrative process to track maintenance rule functional failures and maintenance preventible maintenance functional failures. As a result, reliability and availability could not be balanced as required by the Maintenance Rule periodic evaluation. The safety significance of the specific finding was very low because it did not affect the operability of the systems, and the licensee entered the finding in the corrective action program. However, this finding was considered to be of regulatory concern in the area of maintenance rule implementation due to the extent of the problems with the Maintenance Rule Program.
Inspection Report# : 2001009(pdf)
Significance:          Jun 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO TRACK UNAVAILABILITY OF SYSTEMS REQUIRED DURING SHUTDOWN OPERATION 10 CFR 50.65(a)(1), required, in part, that the licensee monitor the performance or condition of SSCs within the scope of the rule as defined by 10 CFR 50.65(b), against licensee-established goals, in a manner sufficient to provide reasonable assurance that such SSCs are capable of fulfilling their intended functions. 10 CFR 50.65(a)(2) stated, in part, that monitoring as specified in 10 CFR 50.65(a)(1) was not required where it had been demonstrated that the performance or condition of an SSC was being effectively controlled through the performance of appropriate preventive maintenance, such that the SSC remained capable of performing its intended function. Contrary to the above, the licensee failed to demonstrate that the performance or condition of systems required to be available during shutdown conditions and within the scope of the rule had been effectively controlled through the performance of appropriate preventive maintenance and did not monitor against licensee-established goals.
Specifically, the licensee failed to monitor the unavailability of systems required during shutdown operation.
Inspection Report# : 2001009(pdf)
Significance:          Jun 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO ESTABLISH MAINTENANCE RULE (a)(1) GOALS 10 CFR 50.65(a)(1), requires, in part, that the licensee monitor the performance or condition of SSCs within the scope of the rule as defined by 10 CFR 50.65(b), against licensee-established goals, in a manner sufficient to provide reasonable assurance that such structures, systems, and components, are capable of fulfilling their intended functions. Such goals shall be established commensurate with safety. When the performance or condition of a structure, system, or component does not meet established goals, appropriate corrective action shall be taken. Contrary to the above, from 1996, the licensee did not take appropriate corrective actions when the performance of those systems in (a)(1) did not meet licensee
 
2Q/2001 Inspection Findings - Kewaunee                                                                                                      Page 2 of 11 established goals. Specifically, the licensee determined timely and appropriate corrective actions had not been taken for five systems that had been in (a)(1) category for approximately 3 years to 5 years: component cooling (entered (a)(1) on April 23, 1997), control room air conditioning (July 24, 1996), station and instrument air (July 3, 1997), auxiliary building air ventilation (July 31, 1997), and control rod drive (August 6, 1998). This issue is in the licensee's corrective action system as KAP WO 01-3323. The inspectors evaluated the risk significance of this issue using the Significance Determination Process. The inspectors did not identify where this failure resulted in a total loss of a risk significant SSC. Therefore, this issue was screened as Green (very low risk significance) after a Phase 1 Significance Determination Process review. Although the risk significance of this issue was low, the inspectors concluded that this was more than a minor concern because the failure to recognize and correct ineffective maintenance practices resulted in risk significant systems in (a)(1) for years with no improvement in performance. The NRC tracking number for this issue is 50-305/01-09-02.
Inspection Report# : 2001009(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation FIRE DOOR FUSIBLE LINKS.
On February 20, 2001, the licensee determined that the installed fusible link arrangement on roll-up fire Doors 279 and 281, which separated both trains of service water pumps, would not actuate as designed to ensure that the doors would automatically close to provide a 3-hour fire barrier, contrary to 10 CFR Part 50, Appendix R, Section III.G.2.a which required, in part, separation of cables and equipment of redundant trains by a fire barrier having a 3-hour rating.
Inspection Report# : 2001006(pdf)
Significance:        Mar 12, 2001 Identified By: Licensee Item Type: NCV NonCited Violation INADEQUATE SMOKE DETECTOR COVERAGE IN FIRE ZONE TU-95B.
Licensee identified violation of licensee's operating license that the licensee failed to install a detector in each beam pocket in Fire Zone TU-95B.
Inspection Report# : 2001002(pdf)
Significance:        Feb 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO TEST FIRE DOOR IN ACCORDANCE WITH FIRE PLAN.
The inspectors identified a non-cited violation for failure to properly test a fire door in accordance with the facility's fire protection program plan. The finding was of very low safety significance because, although the fire door separated both trains of service water pumps and did not fully close as designed when subsequently tested, the fire loading in the area was insufficient to credibly impact more than two of the four service water pumps in the area.
Inspection Report# : 2001004(pdf)
Significance: N/A Nov 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO COMPLETE COMPONENT RE-TEST REQUIREMENTS PRIOR TO RETURNING EQUIPMENT TO AN OPERABLE STATUS.
No Color. The inspectors identified a Non-Cited Violation for failure to complete component retest requirements following maintenance performed on the B train control room air conditioner compressor condenser. The unit had been returned to an operable status prior to the retest requirements being completed as prescribed in the associated maintenance procedure. This issue was determined to be a violation of the licensee's Operational Quality Assurance Program Manual, Section 8, "Maintenance Planning and Control." Although the risk associated with this finding was very low and did not affect any cornerstones, the inspectors noted that this finding was similar to previous NRC-identified findings and therefore was of greater than minor significance and warranted documentation. (Section 1R19).
Inspection Report# : 2000020(pdf)
Significance:        Sep 29, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IDENTIFY CORROSION AS POTENTIAL FAILURE MECHANISM.
The inspectors identified that the licensee failed to identify corrosion as a potential failure mechanism in the operability determination for a carbon steel key in the service water system. Thus, the licensee failed to quantify the corrosion rate and therefore did not adequately evaluate the expected service life of the carbon steel key. One non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was
 
2Q/2001 Inspection Findings - Kewaunee                                                                                              Page 3 of 11 identified.
Inspection Report# : 2000019(pdf)
Significance:        Sep 29, 2000 Identified By: NRC Item Type: NCV NonCited Violation PRESSURE RATING OF AUXILIARY FEEDWATER STEAM TRAPS.
The inspectors identified that a root cause evaluation for a 1996 equipment issue in the turbine-driven auxiliary feedwater system was not completed until 1999. The evaluation stated that the internals of the steam traps were designed to operate at pressures up to a maximum 600 psig but that the traps were exposed to pressures up to 1025 psig. A corrective action item to initiate a design change request to replace the steam traps with a different model rated for the design pressure of the system was described in the evaluation. However, the inspectors identified that the design change request had never been initiated and the KAP ( Kewaunee corrective action document) had been closed. As a result, the corrective action item for this design problem was lost. In addition, operability of the system had never been formally evaluated despite the identification that the system design requirements were not met. The licensee subsequently determined that the steam traps remained operable and was planning to initiate the design change to correct the problem. One non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was identified.
Inspection Report# : 2000019(pdf)
Significance:        Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation DESIGN CONTROL VIOLATION FOR AUXILIARY FEEDWATER STRAINER MESH SIZE.
The inspectors questioned the mesh size of the strainers, which were installed in the suction of the three auxiliary feedwater (AFW) pumps. As a result of the inspectors' questions, license personnel inspected the strainers on August 21, 2000, and found the strainers to have 1/16 inch openings. A note was later found on Figure 10.2-3 of the UFSAR that indicated that the AFW suction strainer size was 1/8 inch. The smaller openings would not support the use of service water as a safety related source for AFW and as a result all three trains of AFW were declared inoperable. This condition had apparently existed for approximately 25 years and was identified as a non-cited violation of Criterion III, "Design Control," of 10 CFR 50, Appendix B.
Inspection Report# : 2000012(pdf)
Significance: N/A Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation RETRIEVAL OF SERVICE WATER SYSTEM DESIGN INFORMATION.
In many cases, design basis information for the service water system was difficult if not impossible to locate. Licensee personnel wrote KAP WO 00-002566 to enter the problem in the corrective action program.
Inspection Report# : 2000012(pdf)
Significance:        Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PROPERLY CONTROL CORRECTED TEST DATA NECESSARY FOR DESIGN CALCULATIONS.
An example of a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," was identified in the handling of service water system flow test data, which was subsequently used in calculations. Gauge readings corrected for post test calibration checks, gauge reading corrections for elevation considerations, and flow values corrected for pump degradation were contained in spreadsheets in the possession of an individual staff member, but not currently packaged with raw test data, and not bearing evidence of a formal review and control process. The connection between the test data, which had been vaulted, and the values used in the calculation, could not be made without use of the uncontrolled spreadsheet.
Inspection Report# : 2000012(pdf)
Significance:        Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation IMPROPER DESIGN CALCULATION IDENTIFICATION, NON-CONSERVATIVE ASSUMPTIONS, CALCULATION ERRORS, AND DUPLICATE CALCULATIONS.
An example of a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," was identified because of inadequate control of design calculations. The control failures included improper identification of calculations, non-conservative assumptions, calculation errors, and
 
2Q/2001 Inspection Findings - Kewaunee                                                                                                  Page 4 of 11 duplicate or superceded calculations not properly identified or canceled. The failure to follow the established design control process increased the potential for errors in the design and operation of the service water system. Because the system was subsequently demonstrated to be capable of removing the design heat load, the actual significance was low and this finding screened out as having very low risk significance.
Inspection Report# : 2000012(pdf)
Significance:          Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO DOCUMENT IN THE CORRECTIVE ACTION PROGRAM THE USE OF INCORRECT MATERIAL IN A SERVICE WATER PUMP KEY.
An example of a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was identified because of inadequate corrective action to correct an incorrect coupling adjust nut set screw and a low strength "soft" key material, which had contributed to a pump shaft failure. Licensee personnel had known of the "soft" key material since July 21, 1999. The "soft" key material was found in other service water pumps but had not been removed from all pumps. As of July 25, 2000, licensee personnel had not documented the existence of the "soft" key material in the corrective action program.
Inspection Report# : 2000012(pdf)
Significance:          Jul 07, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO COMPLETE COMPONENT RETEST REQUIREMENTS IN ACCORDANCE WITH PROCEDURE.
On June 26, 2000, during a review of post maintenance testing requirements following maintenance performed on the control room post accident system charcoal filter heat detector, the inspectors identified that maintenance technicians had not completed the component re-test requirements, as required by a preventative maintenance procedure prior to the system being returned to an operable status. On July 7, the inspectors identified a second example of failing to complete component re-test requirements following maintenance on the zone special ventilation system charcoal filter heat detector as required. The issue was considered to be of very low safety significance based on the determination that although the licensee had not completed all of the component retest requirements prior to returning the equipment to service, the subsequent testing determined that the equipment was in an operable status. The failure to complete the component retest requirements in accordance with site procedures was identified as a Non-Cited Violation.
Inspection Report# : 2000014(pdf)
Significance:          Feb 21, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform Thorough 10 CFR 50.59 Safety Evaluation The inspectors identified a Non-Cited Violation for failure to perform an adequate 10 CFR 50.59 safety evaluation associated with emergency operating procedure changes to address component cooling water pump dead-head operational concerns. The safety evaluation did not evaluate the potential for initiating a loss-of-coolant accident via the reactor coolant loop seals during conditions of a complete loss of component cooling water.
Inspection Report# : 2001017(pdf)
Significance: N/A Sep 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ESTABLISH CONTINGENCY PLANS FOR ORANGE RISK CONDITION The inspectors identified the failure to establish contingency plans during a planned high risk plant configuration. Contrary to administrative requirements, the licensee approved an orange risk condition during a refueling outage with no contingency plans to mitigate the consequences of a loss of spent fuel pool cooling with a full core offload in the pool. A Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was identified. The finding was of very low safety significance because although the licensee had not approved appropriate contingency actions for the orange risk condition, the licensee subsequently rescheduled the planned maintenance to eliminate the orange risk condition.
Inspection Report# : 2001013(pdf)
Significance: N/A Aug 24, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO CONDUCT BIENNIAL SURVIELLANCE OF SAFETY-RELATED PROCEDURES PER TS 6.8.c A Non-Cited Violation of Technical Specification 6.8.c was identified for the failure to perform a biennial surveillance of safety-related procedues.
This issue was more than minor because if left uncorrected, could under the same condition become a more significant safety concern. However,
 
2Q/2001 Inspection Findings - Kewaunee                                                                                                  Page 5 of 11 since no specific cornerstone had been impacted, this finding is designated as No Color.
Inspection Report# : 2001012(pdf)
Significance:          Aug 24, 2001 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE CORRECTIVE ACTIONS FOR PREVIOUS PROBLEM WITH THE CONTROL OF SAFETY-RELATED MATERIALS A Non-Cited Violation of Criterion XVI, "Corrective Action," of Appendix B of 10 CFR Part 50 was identified for ineffective corrective actions for a problem with the control of the storage of consumable materials, such as thread sealant, used in safety-related applications. These ineffective actions subsequently resulted in the inadequate control of the storage of grease used in safety-related breakers. This issue was more than minor because if left uncorrected, could under the same condition become a more significant safety concern. In that this issue could credibly affect the operability, availability, reliability, or function of a system or train in a mitigating system, it is a Green finding.
Inspection Report# : 2001012(pdf)
Significance:          Aug 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation NON-RATED FIRE BARRIER A Non-Cited Violation [of 10 CFR Part 50, Appendix R, Section III.G.2.a] was identified for failure to provide a 3-hour rated fire barrier to separate redundant trains of safe shutdown equipment. This finding was of very low safety significance because the licensee tested a replica of the fire barrier and demonstrated that the fire barrier provided protection for at least 60 minutes, which was sufficient for the hazards in the area.
Inspection Report# : 2001011(pdf)
Significance:          Jul 20, 2001 Identified By: NRC Item Type: FIN Finding LICENSED OPERATOR REQUALIFICATION EXAMINATION RESULTS The inspectors identified that two of eight crews examined during the licensee's calendar year 2001 licensed operator requalification operating test had failed. The finding was of very low safety significance because both crews that had failed received remedial training prior to being returned to shift, and the results of the licensee's operator licensing requalification operating test given in calendar year 2000 indicated that only one crew, out of a total of eight crews tested, had failed.
Inspection Report# : 2001011(pdf)
Significance:          Jun 22, 2000 Identified By: NRC Item Type: FIN Finding FAILURE TO MEET SINGLE FAILURE CRITERIA FOR RESIDUAL HEAT REMOVAL VALVE CIRCUITRY.
The licensee identified that the circuitry associated with the residual heat removal system discharge to safety injection system suction isolation valves did not meet single failure criteria. The inspectors noted that this design requirement was identified in the facility's updated safety analysis report. The licensee subsequently implemented a temporary change to the facility. The inspectors reviewed the issue and identified that the facility had been operating outside of its design basis, which was reportable to the NRC. The licensee subsequently made a one hour non-emergency report to the NRC. Since there was no actual loss of safety function to the system, this issue was screened as very low risk significance.
Inspection Report# : 2000008(pdf)
Significance:          Jun 22, 2000 Identified By: NRC Item Type: FIN Finding MAINTENANCE RULE FAILURES ASSOCIATED WITH REACTOR HEAD VENT VALVE.
The inspectors reviewed the licensee's implementation of the maintenance rule for failures associated with a reactor head vent valve. The licensee's corrective action documents identified a potential maintenance rule functional failure but the completed evaluation of the problem did not document the final determination. However, the inspectors identified that the repeated failures may have been prevented if maintenance activities such as valve disassembly and cleaning had been performed. In this case, maintenance rule reliability goals were not exceeded. The licensee had documented similar maintenance rule program deficiencies and developed a corrective action program to address the deficiencies. Although programmatic deficiencies exist, since no maintenance rule reliability criteria had been exceeded, this issue was considered of very low risk significance.
 
2Q/2001 Inspection Findings - Kewaunee                                                                                                    Page 6 of 11 Inspection Report# : 2000008(pdf)
Significance:        Jun 22, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO INITIATE KEWAUNEE ASSESSMENT PROCESS DOCUMENT REGARDING REFUELING WATER STORAGE TANK LOW-LOW LEVEL ALARM INOPERABILITY.
The inspectors identified that the refueling water storage tank low-low level alarm which was actuating five percent higher than normal had not been documented in a Kewaunee Assessment Process form by the licensee, and therefore had not received an operability evaluation. This failure was identified as contrary to site administrative procedures. Following the licensee's documentation of the problem, the inspectors identified that the associated operability evaluation considered the acceptability of an operator workaround to address the issue, but did not address any safety implications or consequences of the alarm actuating early. A subsequent operability evaluation by the licensee was evaluated as adequate by the inspectors. Since the subsequent operability evaluation was adequate and it was determined that no safety mitigation equipment was adversely affected by the early actuation of the alarm, this issue was considered of very low risk significance. A non-cited violation (NCV) was identified for failing to document a non-conforming condition, contrary to site administrative procedure requirements.
Inspection Report# : 2000008(pdf)
Significance:        Jun 22, 2000 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO TEST ADDITIONAL RELIEF VALVES IN ACCORDANCE WITH TECHINCAL SPECIFICATIONS.
The licensee identified that the suction relief valve for an auxiliary feedwater pump may have failed its relief test criteria, but did not process the documented deficiency until several weeks later. The licensee then expanded the scope of the relief testing to the suction relief valves associated with the other auxiliary feedwater pumps to meet technical specification requirements. Since any one train of auxiliary feedwater was capable of supplying 100 percent of the decay heat removal requirements, this issue was screened as very low risk significance. However, the time delay in complying with technical specification requirements for testing other relief valves was identified as an NCV.
Inspection Report# : 2000008(pdf)
Significance:        May 22, 2000 Identified By: NRC Item Type: FIN Finding FIRE EXTINGUISHERS NOT LOCATED IN ALL AREAS OF CONTAINMENT BUILDING AT BEGINNING OF REFUELING OUTAGE.
During a walkdown of the containment building, the inspectors identified that portable fire extinguishers were not located in the containment basement at the beginning of the plant refueling outage. Additionally, site fire protection procedures required that responsible fire protection personnel perform inspections of selected plant areas to ensure that the quantity of combustible material was minimized. However, the procedure did not list the containment as an area to be inspected and the procedures did not require the placement and location of portable fire suppression equipment inside containment during the refueling outage. Due to a low number of work activities ongoing at the time, this issue was screened as Green (very low risk significance).
Inspection Report# : 2000007(pdf)
Significance:        May 22, 2000 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO INSTALL RAYCHEM HEAT SHRINK MATERIAL IN ACCORDANCE WITH PROCEDURE REQUIREMENTS.
The licensee identified that two Raychem electrical cable splices utilized in environmentally qualified (EQ) safety-related equipment had not been installed in accordance with EQ requirements. These splices were associated with pressurizer level transmitters and were installed in 1984. The licensee subsequently performed extensive EQ testing of the splices to determine the qualification of the splices' as-found configurations. Test results indicated that the splices would have been able to perform their intended function in a harsh environment inside containment. This issue was considered to be of low safety significance based on the successful EQ testing of the as-found splices' configurations and was screened as Green (very low risk significance). One non-cited violation was identified.
Inspection Report# : 2000007(pdf)
Barrier Integrity
 
2Q/2001 Inspection Findings - Kewaunee                                                                                                  Page 7 of 11 Significance:        Jun 22, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation FAILURE TO UPDATE COMPUTER ALARM FOR CURRENT AXIAL FLUX DISTRIBUTION TARGET BAND.
The licensee identified, following plant startup, that a computer alarm had not been updated properly to alarm if axial flux distribution deviated outside of the flux distribution target band. This condition was contrary to technical specification requirements. The licensee reviewed the axial flux distribution history since the startup and determined that at no time was the flux distribution outside of the target band. Since the axial flux distribution was never outside of the target band, this issue was screened as very low risk significance. An NCV was identified for failing to comply with technical specification requirements for monitoring axial flux distributions.
Inspection Report# : 2000008(pdf)
Significance:        May 22, 2000 Identified By: Licensee Item Type: FIN Finding TECHNICAL SPECIFICATIONS INTERPRETATION FOR TESTING REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES.
The NRC determined that the licensee's practice of testing reactor coolant system pressure isolation Valve SI-22B prior to entering the cold shutdown condition was contrary to Technical Specification requirements 4.2.a.3.a. Technical Specification 4.2.a.3.a required that periodic leakage testing of Valve SI-22B be accomplished prior to reaching operating mode after the plant was placed in cold shutdown. This issue was considered to be of low safety significance because of a subsequent successful valve test and was screened as Green (very low risk significance).
Enforcement discretion was applied to this item in accordance with Section VII.B.6 of the Enforcement Policy.
Inspection Report# : 2000007(pdf)
Significance:        May 19, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PROPERLY IMPLEMENT FLAW ACCEPTANCE CRITERIA FOR SLEEVE WELD INSPECTIONS.
During 1998 inservice inspection examinations, the licensee failed to properly implement the flaw acceptance criteria for laser welded sleeve inspection within two steam generator tubes. The safety significance was very low based on the absence of adverse consequences, and May 2000 in-situ pressure testing where both welds exhibited zero leakage at normal operating pressure, main steam line break pressure, and three times normal operating differential pressures. As such, this issue was characterized as a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion IX, Control of Special Processes. To correct the error, the licensee plugged both tubes (Section 1RO8).
Inspection Report# : 2000011(pdf)
Emergency Preparedness Significance:        Mar 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation SUPPLEMENTAL INSPECTION OF WHITE ERO AUGMENTATION FINDING AND RESULTING GREEN FINDING.
This supplemental inspection was performed by the NRC to evaluate the licensees evaluation associated with the failure to conduct successful quarterly, off-hours, unannounced staff augmentation drills during the second, third, and fourth quarters of 1999 and the second quarter of 2000.
This performance issue was previously characterized as having low to moderate risk significance (White) in NRC Inspection Report No. 50-305/2000015(DRS). During this supplemental inspection, performed in accordance with Inspection Procedure95002, the inspector concluded that the licensee performed a comprehensive evaluation of the unsuccessful staff augmentation drills. The licensees evaluation identifiedtwo root causes which resulted in the unsuccessful drills and in the staffs inability tocorrect thedeficiencies: (1) Management has not effectively acted to provide increased depth and flexibility in the emergency response organization following a reduction in staffing several years ago; and (2)
Management has accepted an adverse trend of test failures without requiring investigation into the root causes. The inspector reviewed the licensees corrective actions, both completed and planned, and concluded that the programmatic corrective actions appeared to address the identified root causes. In particular, the licensee assigned certain positions to an on-call rotation to ensure personnel were capable of augmenting in a timely manner, and the licensee was progressing in training additional staff to increase the depth of personnel assigned to key emergency response positions. In addition, the licensee was continuing its efforts in improving its corrective action program. The inspector reviewed the licensees immediate response to the issue and identified that one of the licensees initial corrective actions resulted in a Non-Cited Violation of regulatory requirements. To obtain a timely response of a key emergency response position (severe accident management - core hydraulics), the emergency preparedness staff effectively changed the emergency plan without revising the necessary procedures and without formally assessing the impact of that change. The staff instructed and trained personnel to respond to a location other than the Technical Support Center, which was
 
2Q/2001 Inspection Findings - Kewaunee                                                                                                    Page 8 of 11 contrary to the licensees current implementing procedures. While this change enabled the licensee to augment its staff in a timely manner, the change was not performed in accordance with NRC requirements. In order to make such a change, the licensees emergency plan required that the change be formally assessed to ensure that it did not reduce the effectiveness of the plan or any other implementing procedure. Since the issue did not result in a failure to meet an emergency preparedness planning standard, the failure to adequately implement the emergency plan was determined to be a violation of very low safety significance (Green) (Section 02.3(a)). Due to the licensees acceptable performance in assessing the emergency response augmentation drill deficiencies, the White finding associated with this issue will only be considered in assessing plant performance for a total of four quarters in accordance with the guidance in IMC 0305, Operating Reactor Assessment Program. Implementation of the licensees corrective actions will be reviewed during a future inspection.
Inspection Report# : 2001007(pdf)
Significance:          Mar 30, 2001 Identified By: NRC Item Type: VIO Violation FAILURE TO CORRECT SELF-IDENTIFIED ERO AUGMENTATION DRILL DEFICIENCIES.
During a baseline inspection of the emergency preparedness program conducted on August 14 - September 21, 2000, the NRC identified a preliminary White issue and potential violation for the licensees failure to successfully correct deficiencies identified during staff augmentation drills and to demonstrate timely staff augmentation in 1999 and 2000. The issue was unresolved pending the outcome of the NRCs final significance determination. On January 30, 2001, the NRC conducted a regulatory conference with the licensee and subsequently issued the licensee a White finding and Notice of Violation (Enforcement Action No. 00-214) associated with the performance issue.
Inspection Report# : 2001007(pdf)
Significance:          Mar 23, 2001 Identified By: NRC Item Type: NCV NonCited Violation POST-ACCIDENT SAMPLING SYSTEM CONTAINMENT AIR SAMPLE PANEL TESTING.
A Non-Cited Violation of Technical Specification 6.14 was identified for the failure to implement a program that ensured the capability to obtain and analyze containment atmosphere samples under accident samples using the cantainment air sampling panel (CASP). Although the CASP was installed, as was indicated in the emergency plan, the licensee had neither developed procedures nor had tested its capability to obtain a containment atmosphere sample using the CASP. The licensee could not recall if and when containment air samples were last obtained using the CASP.
Inspection Report# : 2001006(pdf)
Significance: N/A Feb 27, 2001 Identified By: NRC Item Type: FIN Finding SUPPLEMENTAL INSPECTION OF YELLOW ANS PERFORMANCE INDICATOR AND ASSOCIATED CORRECTIVE ACTION PROGRAM DEFICIENCIES.
This supplemental inspection was performed by the NRC to assess the licensee's evaluation associated with a Yellow performance indicator for the Alert and Notification System (ANS) and the associated Yellow finding related to the licensee's corrective action program. During this supplemental inspection, performed in accordance with Inspection Procedure 95002, the inspectors concluded that the licensee performed comprehensive evaluations of the performance problems associated with the ANS and its corrective action program. These evaluations identified primary root causes and contributing causes for both issues. Along with the electronics/hardware problems, the licensee identified the primary root causes for the ANS to be the failure to make changes to the system via a change control process and the failure of the activation procedure to provide for alternate/backup activation methods and to provide clear success criteria. In the case of the corrective action program, the licensee concluded that the primary root causes were plant management's inadequate risk evaluation regarding decisions affecting the corrective action program and the development of a culture in the licensee's organization, which minimized the importance of information from outside organizations. In particular, the licensee focused on low-cost power operation and failed to recognize the value of changes and improvements which had occurred throughout the industry, such as the value and expansion of the corrective action program. In the case of the Yellow ANS performance indicator, the licensee had completed several significant corrective actions to address the root causes and contributing causes identified in its evaluation. The inspectors found that the corrective actions appeared appropriate to address the underlying root causes and that ANS testing data indicated an improving trend in the NRC performance indicator. The licensee also performed comprehensive assessments of the emergency preparedness program, quality assurance program, plant operations, and other plant programs to determine the extent of condition (re. the root causes described above). Based on these evaluations, the licensee began to implement significant actions to correct the deficiencies in the corrective action program and other weaknesses identified. Generally, the inspectors observed progress in the licensee's initial implementation of these corrective actions. Due to the licensee's acceptable performance in assessing the Yellow ANS performance indicator and the associated Yellow finding, the Yellow finding will not be considered in assessing future plant performance.
Inspection Report# : 2001005(pdf)
Significance:          Sep 21, 2000 Identified By: NRC
 
2Q/2001 Inspection Findings - Kewaunee                                                                                                  Page 9 of 11 Item Type: FIN Finding INADEQUATE ROOT CAUSE EVALUATION FOR YELLOW ALERT AND NOTIFICATION SYSTEM PERFORMANCE INDICATOR.
The licensee's evaluation of the Yellow Alert and Notification (siren) System Performance Indicator (PI) was inadequate. The inspector concluded that the licensee's evaluation was not performed at the depth necessary to identify the root causes of the siren performance problems and, instead, only identified the symptoms of the root causes. Specifically, the inspector identified the following substantive weaknesses in the licensee's evaluation of the siren system performance, which appeared to result from systemic corrective action program deficiencies within this cornerstone:
* The licensee's evaluation was not of sufficient depth to clearly identify the root causes associated with the decline in siren system performance.
* Licensee management did not provide well-understood and clear guidance/expectations for performing root cause evaluations.
* The licensee's evaluation of the quality assurance program was narrowly focused and was not critical of its role in failing to identify and correct the siren performance problems.
* The licensee did not establish a priority for each of the long-term corrective actions in accordance with the associated significance or risk.
* The licensee did not have any formal provisions for measuring the effectiveness of its corrective actions.
* Within the licensee's evaluation, the licensee had not evaluated common causes or the extent of the condition. Due to the corrective action program performance deficiencies within this cornerstone, we have been unable to conclude that the performance issues that resulted in the yellow PI have been addressed. Therefore, we are issuing a yellow finding that corresponds to the original issues that resulted in a yellow PI. Additional inspection effort will be focused on the licensee's further evaluation of the siren reliability root causes and the continuing corrective action program implementation deficiencies identified during this inspection.
Inspection Report# : 2000017(pdf)
Significance: N/A Apr 05, 2000 Identified By: NRC Item Type: FIN Finding LICENSEE FAILED TO IDENTIFY THE FULL SCOPE OF PROBLEMS WITH THE ALERT AND NOTIFICATION SYSTEM PERFORMANCE.
The inspectors concluded that the licensee's assessment was not sufficiently comprehensive to identify the full scope of problems associated with the Alert and Notification System (ANS) performance program. As a result, licensee corrective actions generally were focused on the equipment problem rather than all root causes and contributing factors. For example, the inspectors identified that: (1) management oversight of the ANS performance program was limited; (2) an audit failed to identify degrading ANS performance as a concern; (3) annual preventive maintenance was not consistently performed on the system; (4) the corrective action program was not used consistently to document ANS problems; and (5) maintenance procedures and records were deficient. Collectively, these problems indicate that the ANS performance program lacked sufficient structure and oversight.
Inspection Report# : 2000006(pdf)
Occupational Radiation Safety Significance:        Oct 02, 2001 Identified By: NRC Item Type: NCV NonCited Violation HIGH RADIATION AREA ACCESS CONTROLS Non-Cited Violation of Technical Specification 6.13 and an associated Green Finding for failure to 'barricade' three ladders that provided entry to high radiation areas (less than 1000 mrem/hour) located on the steam generator/pressurizer platforms.
Inspection Report# : 2001014(pdf)
Significance:        May 19, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO POST A VERY HIGH RADIATION AREA.
The inspectors identified a noncited violation for the failure to post a very high radiation area in accordance with 10 CFR 20.1902(c). Although the area was not adequately posted, the licensee had provided physical controls and barriers that were consistent with its requirements for a very high radiation area. Based on the adequacy of these controls, the potential for an overexposure from the inadvertent entry of personnel into the area was low. Consequently, this finding was determined to be of very low safety significance (Section 20S1.1).
Inspection Report# : 2000009(pdf)
Public Radiation Safety
 
2Q/2001 Inspection Findings - Kewaunee                                                                                                Page 10 of 11 Physical Protection Significance: N/A Jul 21, 2000 Identified By: NRC Item Type: FIN Finding CORRECTIVE ACTION ON SEARCH ISSUES DID NOT WORK.
The inspector determined that the licensee's effectiveness of implemented corrective actions for a previously identified inspection finding regarding an inadequate vehicle search was not totally effective in preventing recurrence. Previous corrective action was not adequately focused (Section 3PP2.2).
Inspection Report# : 2000013(pdf)
Significance:        Jul 21, 2000 Identified By: NRC Item Type: NCV NonCited Violation SEARCHES INADEQUATE (REPEAT).
The inspector identified a Non-Cited violation by observing that, a security officer failed to search an easily accessible compartment on one vehicle.
The failure resulted from human error because the officer did not observe the access panel to the compartment during the vehicle search process.
When searched, no prohibited items were found. Corrective actions were implemented. The inspector determined that the licensee's effectiveness of implemented corrective actions for a previously identified inspection finding regarding an inadequate vehicle search was not totally effective in preventing recurrence. Previous corrective action was not adequately focused.
Inspection Report# : 2000013(pdf)
Miscellaneous Significance: N/A Nov 09, 2000 Identified By: NRC Item Type: FIN Finding FAILURE TO TAKE ADEQUATE CORRECTIVE ACTIONS TO ENSURE COMPONENT RE-TEST REQUIREMENTS COMPLETED ACCORDING TO MAINTENANCE PROCEDURES.
No Color. The inspectors determined that a negative performance trend had developed in the licensee's ability to identify and promptly take appropriate corrective actions to prevent recurrence based on two previously identified examples (NCV 50-305/2000014-01) and one example identified during this inspection period (NCV 50-305/00-20-01). All three examples related to the licensee returning safety-related equipment to service prior to completing all required post-maintenance retesting. While the risk of the individual examples was very low, the licensee had failed to ensure that all retest requirements had been completed before returning safety-related equipment to service. These findings collectively indicated a problem with the licensee's ability to provide timely and adequate corrective actions to prevent recurrence. (Section 4OA2).
Inspection Report# : 2000020(pdf)
Significance: N/A Sep 29, 2000 Identified By: NRC Item Type: FIN Finding EFFECTIVENESS OF CORRECTIVE ACTION PROGRAM.
Based on the results of this inspection, the NRC concluded that the corrective action program at Kewaunee showed significant weaknesses and inconsistencies across all of the procedural elements inspected. These weaknesses existed across departments and affected multiple cornerstones in the strategic performance areas of Reactor Safety, Radiation Safety, and Safeguards. Of particular note was the lack of procedures for determining the significance of conditions adverse to quality and for trending of issues and the complete lack of trending within your corrective action program. We also identified a lack of urgency in correcting issues which resulted in repeat examples occurring and, coupled with a poor tracking system, a tendency for issues to be dropped. While none of the specific examples identified by the team were of high risk significance when looked at in isolation, in the aggregate they were similar in nature to prior issues in the emergency preparedness area that rose to a higher significance level and contributed to a degraded cornerstone. While we concluded that the station had fostered an environment in which personnel freely identified conditions adverse to quality without fear of discrimination or retaliation, we also concluded that significant weaknesses with, and inconsistent implementation of, the station corrective action program resulted in multiple examples where station personnel did not enter deficiencies into the station's formal corrective action program.
Inspection Report# : 2000019(pdf)
Significance: N/A Sep 29, 2000 Identified By: NRC Item Type: NCV NonCited Violation NO PROCEDURAL GUIDANCE FOR DETERMINING IF CONDITIONS ADVERSE TO QUALITY ARE SIGNIFICANT.
The inspectors reviewed the quality assurance (QA) manual requirements against Kewaunee's implementing procedures and identified that two QA manual requirements were not being implemented. Specifically, Program Requirement 3.1.9 which stated that directives and procedures shall
 
2Q/2001 Inspection Findings - Kewaunee                                                                                                Page 11 of 11 provide for the review of conditions adverse to quality to determine if the conditions are significant in nature. This requirement paralleled 10 CFR Part 50, Appendix B, Criterion XVI, which requires that the cause of significant conditions adverse to quality be determined and corrective actions taken to prevent recurrence. The inspectors reviewed the Nuclear Administrative Directive (NAD 11.08) and the procedure (GNP 11.08.01) governing the KAP (Kewaunee corrective action program) process and found no procedure requirements for identifying significant conditions adverse to quality. This finding does not directly affect a cornerstone. As a result, this issue was not evaluated with the Significance Determination Process and was not assigned a color. One example of a non-cited violation of 10 CFR Part 50, Appendix B, Criterion II, "Quality Assurance Programs," was identified.
Inspection Report# : 2000019(pdf)
Significance: N/A Sep 29, 2000 Identified By: NRC Item Type: NCV NonCited Violation NO PROCEDURAL GUIDANCE FOR TRENDING CONDITIONS ADVERSE TO QUALITY.
The inspectors reviewed the quality assurance (QA) manual requirements against Kewaunee's implementing procedures and identified that two QA manual requirements were not being implemented. Specifically, QA Program Requirement, 3.1.10, stated that directives and procedures shall provide for analyzing trends of conditions adverse to quality. Once identified these trends were required to be considered significant conditions adverse to quality. The inspectors found that conditions adverse to quality were not defined in the KAP (Kewaunee corrective action program) procedures and that no procedure existed for trending. This finding does not directly affect a cornerstone. As a result, this issue was not evaluated with the Significance Determination Process and was not assigned a color. One example of a non-cited violation of 10 CFR Part 50, Appendix B, Criterion II, "Quality Assurance Program," was identified.
Inspection Report# : 2000019(pdf)
Significance: N/A Feb 21, 2002 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Implement Required Fire Watch Following Completion of Hot Work Activities A licensee-identified violation was reviewed by the inspectors. Corrective actions taken or planned by the licensee appeared reasonable.
Inspection Report# : 2001017(pdf)
Significance: N/A Aug 24, 2001 Identified By: NRC Item Type: FIN Finding IDENTIFICATION AND RESOLUTION OF PROBLEMS The team concluded that the licensee was generally effective at identifying problems and putting them into the corrective action program. The program itself contained all the necessary attributes of an acceptable corrective action program and was generally successful in correcting identified issues. However, the team noted that, although licensee management had taken efforts to ensure that issues were resolved in accordance with program guidance and requirements, additional efforts appeared necessary to ensure timely resolution of issues. A positive program initiative was the establishment of positions in each of the major plant departments to serve as liaisons between the departments and the corrective action program and to assist with self-assessments. However, examples were identified by the inspectors of problems with the licensee's identification and resolution of problems, prioritization and evaluation of issues, and the effectiveness of corrective actions. Included in these examples were the routine granting of due date extensions for problem evaluation and corrective action implementation, failure to perform a Technical Specification-required biennial surveillance of safety-related procedures, and ineffective corrective actions that resulted in the lack of proper controls over the storage of grease used in safety-related breakers. Based on a review of records and discussions with plant staff, the inspectors concluded that workers at the site felt free to input safety issues into the corrective action program.
Inspection Report# : 2001012(pdf)
Significance: N/A Jun 22, 2000 Identified By: NRC Item Type: FIN Finding CONTROL ROOM OPERATIONS HUMAN PERFORMANCE ISSUES.
The inspectors interviewed operators to evaluate their awareness of degraded control room indications and alarms, and their ability to adequately take manual actions based on degraded alarm functions. The inspectors identified, during interviews, that there was a lack of awareness by operators of a degraded refueling water storage tank low-low level alarm which would be potentially confusing to operators and therefore increase the risk associated with initiating long term sump recirculation.
Inspection Report# : 2000008(pdf)
Last modified : March 27, 2002
 
3Q/2001 Inspection Findings - Kewaunee                                                                                                  Page 1 of 11 Kewaunee Initiating Events Mitigating Systems Significance: N/A Sep 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ESTABLISH CONTINGENCY PLANS FOR ORANGE RISK CONDITION The inspectors identified the failure to establish contingency plans during a planned high risk plant configuration. Contrary to administrative requirements, the licensee approved an orange risk condition during a refueling outage with no contingency plans to mitigate the consequences of a loss of spent fuel pool cooling with a full core offload in the pool. A Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was identified. The finding was of very low safety significance because although the licensee had not approved appropriate contingency actions for the orange risk condition, the licensee subsequently rescheduled the planned maintenance to eliminate the orange risk condition.
Inspection Report# : 2001013(pdf)
Significance: N/A Aug 24, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO CONDUCT BIENNIAL SURVIELLANCE OF SAFETY-RELATED PROCEDURES PER TS 6.8.c A Non-Cited Violation of Technical Specification 6.8.c was identified for the failure to perform a biennial surveillance of safety-related procedues.
This issue was more than minor because if left uncorrected, could under the same condition become a more significant safety concern. However, since no specific cornerstone had been impacted, this finding is designated as No Color.
Inspection Report# : 2001012(pdf)
Significance:          Aug 24, 2001 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE CORRECTIVE ACTIONS FOR PREVIOUS PROBLEM WITH THE CONTROL OF SAFETY-RELATED MATERIALS A Non-Cited Violation of Criterion XVI, "Corrective Action," of Appendix B of 10 CFR Part 50 was identified for ineffective corrective actions for a problem with the control of the storage of consumable materials, such as thread sealant, used in safety-related applications. These ineffective actions subsequently resulted in the inadequate control of the storage of grease used in safety-related breakers. This issue was more than minor because if left uncorrected, could under the same condition become a more significant safety concern. In that this issue could credibly affect the operability, availability, reliability, or function of a system or train in a mitigating system, it is a Green finding.
Inspection Report# : 2001012(pdf)
Significance:          Aug 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation NON-RATED FIRE BARRIER A Non-Cited Violation [of 10 CFR Part 50, Appendix R, Section III.G.2.a] was identified for failure to provide a 3-hour rated fire barrier to separate redundant trains of safe shutdown equipment. This finding was of very low safety significance because the licensee tested a replica of the fire barrier and demonstrated that the fire barrier provided protection for at least 60 minutes, which was sufficient for the hazards in the area.
Inspection Report# : 2001011(pdf)
Significance:          Jul 20, 2001 Identified By: NRC Item Type: FIN Finding LICENSED OPERATOR REQUALIFICATION EXAMINATION RESULTS The inspectors identified that two of eight crews examined during the licensee's calendar year 2001 licensed operator requalification operating test
 
3Q/2001 Inspection Findings - Kewaunee                                                                                                  Page 2 of 11 had failed. The finding was of very low safety significance because both crews that had failed received remedial training prior to being returned to shift, and the results of the licensee's operator licensing requalification operating test given in calendar year 2000 indicated that only one crew, out of a total of eight crews tested, had failed.
Inspection Report# : 2001011(pdf)
Significance: N/A Jun 30, 2001 Identified By: NRC Item Type: FIN Finding FAILURE TO BALANCE RELIABILITY AND AVAILABILITY AS REQUIRED BY 10 CFR 50.65(a)(3)
The inspectors identified a failure to evaluate whether adjustments were necessary such that there would be an appropriate balance between systems' availability and reliability in accordance with 10 CFR 50.65(a)(3) of the maintenance rule. The inspectors identified that the licensee did not have an administrative process to track maintenance rule functional failures and maintenance preventible maintenance functional failures. As a result, reliability and availability could not be balanced as required by the Maintenance Rule periodic evaluation. The safety significance of the specific finding was very low because it did not affect the operability of the systems, and the licensee entered the finding in the corrective action program. However, this finding was considered to be of regulatory concern in the area of maintenance rule implementation due to the extent of the problems with the Maintenance Rule Program.
Inspection Report# : 2001009(pdf)
Significance:          Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IDENTIFY DEFICIENT CONDITION OF VALVE AFW-1B The inspectors identified that the licensee failed to promptly identify and correct the B' train auxiliary feedwater pump discharge check valve which was stuck in an intermediate position. A Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was issued. The finding was of very low safety significance because, although the check valve was stuck in an intermediate position, the time that it was known to have been stuck was less than the technical specification allowed outage time for one train of auxiliary feedwater to be out of service (less than 72 hours). Additionally, the other two trains of auxiliary feedwater were each capable of 100 percent decay heat removal.
Inspection Report# : 2001009(pdf)
Significance:          Jun 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO TRACK UNAVAILABILITY OF SYSTEMS REQUIRED DURING SHUTDOWN OPERATION 10 CFR 50.65(a)(1), required, in part, that the licensee monitor the performance or condition of SSCs within the scope of the rule as defined by 10 CFR 50.65(b), against licensee-established goals, in a manner sufficient to provide reasonable assurance that such SSCs are capable of fulfilling their intended functions. 10 CFR 50.65(a)(2) stated, in part, that monitoring as specified in 10 CFR 50.65(a)(1) was not required where it had been demonstrated that the performance or condition of an SSC was being effectively controlled through the performance of appropriate preventive maintenance, such that the SSC remained capable of performing its intended function. Contrary to the above, the licensee failed to demonstrate that the performance or condition of systems required to be available during shutdown conditions and within the scope of the rule had been effectively controlled through the performance of appropriate preventive maintenance and did not monitor against licensee-established goals.
Specifically, the licensee failed to monitor the unavailability of systems required during shutdown operation.
Inspection Report# : 2001009(pdf)
Significance:          Jun 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO ESTABLISH MAINTENANCE RULE (a)(1) GOALS 10 CFR 50.65(a)(1), requires, in part, that the licensee monitor the performance or condition of SSCs within the scope of the rule as defined by 10 CFR 50.65(b), against licensee-established goals, in a manner sufficient to provide reasonable assurance that such structures, systems, and components, are capable of fulfilling their intended functions. Such goals shall be established commensurate with safety. When the performance or condition of a structure, system, or component does not meet established goals, appropriate corrective action shall be taken. Contrary to the above, from 1996, the licensee did not take appropriate corrective actions when the performance of those systems in (a)(1) did not meet licensee established goals. Specifically, the licensee determined timely and appropriate corrective actions had not been taken for five systems that had been in (a)(1) category for approximately 3 years to 5 years: component cooling (entered (a)(1) on April 23, 1997), control room air conditioning (July 24, 1996), station and instrument air (July 3, 1997), auxiliary building air ventilation (July 31, 1997), and control rod drive (August 6, 1998). This issue is in the licensee's corrective action system as KAP WO 01-3323. The inspectors evaluated the risk significance of this issue using the Significance Determination Process. The inspectors did not identify where this failure resulted in a total loss of a risk significant SSC. Therefore, this issue was screened as Green (very low risk significance) after a Phase 1 Significance Determination Process review. Although the risk significance of this issue was low, the inspectors concluded that this was more than a minor concern because the failure to recognize and correct ineffective maintenance practices resulted in risk significant systems in (a)(1) for years with no improvement in performance. The NRC tracking number for
 
3Q/2001 Inspection Findings - Kewaunee                                                                                                      Page 3 of 11 this issue is 50-305/01-09-02.
Inspection Report# : 2001009(pdf)
Significance:          Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation FIRE DOOR FUSIBLE LINKS.
On February 20, 2001, the licensee determined that the installed fusible link arrangement on roll-up fire Doors 279 and 281, which separated both trains of service water pumps, would not actuate as designed to ensure that the doors would automatically close to provide a 3-hour fire barrier, contrary to 10 CFR Part 50, Appendix R, Section III.G.2.a which required, in part, separation of cables and equipment of redundant trains by a fire barrier having a 3-hour rating.
Inspection Report# : 2001006(pdf)
Significance:          Mar 12, 2001 Identified By: Licensee Item Type: NCV NonCited Violation INADEQUATE SMOKE DETECTOR COVERAGE IN FIRE ZONE TU-95B.
Licensee identified violation of licensee's operating license that the licensee failed to install a detector in each beam pocket in Fire Zone TU-95B.
Inspection Report# : 2001002(pdf)
Significance:          Feb 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO TEST FIRE DOOR IN ACCORDANCE WITH FIRE PLAN.
The inspectors identified a non-cited violation for failure to properly test a fire door in accordance with the facility's fire protection program plan. The finding was of very low safety significance because, although the fire door separated both trains of service water pumps and did not fully close as designed when subsequently tested, the fire loading in the area was insufficient to credibly impact more than two of the four service water pumps in the area.
Inspection Report# : 2001004(pdf)
Significance: N/A Nov 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO COMPLETE COMPONENT RE-TEST REQUIREMENTS PRIOR TO RETURNING EQUIPMENT TO AN OPERABLE STATUS.
No Color. The inspectors identified a Non-Cited Violation for failure to complete component retest requirements following maintenance performed on the B train control room air conditioner compressor condenser. The unit had been returned to an operable status prior to the retest requirements being completed as prescribed in the associated maintenance procedure. This issue was determined to be a violation of the licensee's Operational Quality Assurance Program Manual, Section 8, "Maintenance Planning and Control." Although the risk associated with this finding was very low and did not affect any cornerstones, the inspectors noted that this finding was similar to previous NRC-identified findings and therefore was of greater than minor significance and warranted documentation. (Section 1R19).
Inspection Report# : 2000020(pdf)
Significance:          Feb 21, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform Thorough 10 CFR 50.59 Safety Evaluation The inspectors identified a Non-Cited Violation for failure to perform an adequate 10 CFR 50.59 safety evaluation associated with emergency operating procedure changes to address component cooling water pump dead-head operational concerns. The safety evaluation did not evaluate the potential for initiating a loss-of-coolant accident via the reactor coolant loop seals during conditions of a complete loss of component cooling water.
Inspection Report# : 2001017(pdf)
Significance:          Sep 29, 2000 Identified By: NRC Item Type: NCV NonCited Violation
 
3Q/2001 Inspection Findings - Kewaunee                                                                                                  Page 4 of 11 FAILURE TO IDENTIFY CORROSION AS POTENTIAL FAILURE MECHANISM.
The inspectors identified that the licensee failed to identify corrosion as a potential failure mechanism in the operability determination for a carbon steel key in the service water system. Thus, the licensee failed to quantify the corrosion rate and therefore did not adequately evaluate the expected service life of the carbon steel key. One non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was identified.
Inspection Report# : 2000019(pdf)
Significance:        Sep 29, 2000 Identified By: NRC Item Type: NCV NonCited Violation PRESSURE RATING OF AUXILIARY FEEDWATER STEAM TRAPS.
The inspectors identified that a root cause evaluation for a 1996 equipment issue in the turbine-driven auxiliary feedwater system was not completed until 1999. The evaluation stated that the internals of the steam traps were designed to operate at pressures up to a maximum 600 psig but that the traps were exposed to pressures up to 1025 psig. A corrective action item to initiate a design change request to replace the steam traps with a different model rated for the design pressure of the system was described in the evaluation. However, the inspectors identified that the design change request had never been initiated and the KAP ( Kewaunee corrective action document) had been closed. As a result, the corrective action item for this design problem was lost. In addition, operability of the system had never been formally evaluated despite the identification that the system design requirements were not met. The licensee subsequently determined that the steam traps remained operable and was planning to initiate the design change to correct the problem. One non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was identified.
Inspection Report# : 2000019(pdf)
Significance:        Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation DESIGN CONTROL VIOLATION FOR AUXILIARY FEEDWATER STRAINER MESH SIZE.
The inspectors questioned the mesh size of the strainers, which were installed in the suction of the three auxiliary feedwater (AFW) pumps. As a result of the inspectors' questions, license personnel inspected the strainers on August 21, 2000, and found the strainers to have 1/16 inch openings. A note was later found on Figure 10.2-3 of the UFSAR that indicated that the AFW suction strainer size was 1/8 inch. The smaller openings would not support the use of service water as a safety related source for AFW and as a result all three trains of AFW were declared inoperable. This condition had apparently existed for approximately 25 years and was identified as a non-cited violation of Criterion III, "Design Control," of 10 CFR 50, Appendix B.
Inspection Report# : 2000012(pdf)
Significance: N/A Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation RETRIEVAL OF SERVICE WATER SYSTEM DESIGN INFORMATION.
In many cases, design basis information for the service water system was difficult if not impossible to locate. Licensee personnel wrote KAP WO 00-002566 to enter the problem in the corrective action program.
Inspection Report# : 2000012(pdf)
Significance:        Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PROPERLY CONTROL CORRECTED TEST DATA NECESSARY FOR DESIGN CALCULATIONS.
An example of a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," was identified in the handling of service water system flow test data, which was subsequently used in calculations. Gauge readings corrected for post test calibration checks, gauge reading corrections for elevation considerations, and flow values corrected for pump degradation were contained in spreadsheets in the possession of an individual staff member, but not currently packaged with raw test data, and not bearing evidence of a formal review and control process. The connection between the test data, which had been vaulted, and the values used in the calculation, could not be made without use of the uncontrolled spreadsheet.
Inspection Report# : 2000012(pdf)
Significance:        Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation
 
3Q/2001 Inspection Findings - Kewaunee                                                                                                  Page 5 of 11 IMPROPER DESIGN CALCULATION IDENTIFICATION, NON-CONSERVATIVE ASSUMPTIONS, CALCULATION ERRORS, AND DUPLICATE CALCULATIONS.
An example of a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," was identified because of inadequate control of design calculations. The control failures included improper identification of calculations, non-conservative assumptions, calculation errors, and duplicate or superceded calculations not properly identified or canceled. The failure to follow the established design control process increased the potential for errors in the design and operation of the service water system. Because the system was subsequently demonstrated to be capable of removing the design heat load, the actual significance was low and this finding screened out as having very low risk significance.
Inspection Report# : 2000012(pdf)
Significance:          Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO DOCUMENT IN THE CORRECTIVE ACTION PROGRAM THE USE OF INCORRECT MATERIAL IN A SERVICE WATER PUMP KEY.
An example of a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was identified because of inadequate corrective action to correct an incorrect coupling adjust nut set screw and a low strength "soft" key material, which had contributed to a pump shaft failure. Licensee personnel had known of the "soft" key material since July 21, 1999. The "soft" key material was found in other service water pumps but had not been removed from all pumps. As of July 25, 2000, licensee personnel had not documented the existence of the "soft" key material in the corrective action program.
Inspection Report# : 2000012(pdf)
Significance:          Jul 07, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO COMPLETE COMPONENT RETEST REQUIREMENTS IN ACCORDANCE WITH PROCEDURE.
On June 26, 2000, during a review of post maintenance testing requirements following maintenance performed on the control room post accident system charcoal filter heat detector, the inspectors identified that maintenance technicians had not completed the component re-test requirements, as required by a preventative maintenance procedure prior to the system being returned to an operable status. On July 7, the inspectors identified a second example of failing to complete component re-test requirements following maintenance on the zone special ventilation system charcoal filter heat detector as required. The issue was considered to be of very low safety significance based on the determination that although the licensee had not completed all of the component retest requirements prior to returning the equipment to service, the subsequent testing determined that the equipment was in an operable status. The failure to complete the component retest requirements in accordance with site procedures was identified as a Non-Cited Violation.
Inspection Report# : 2000014(pdf)
Significance:          Jun 22, 2000 Identified By: NRC Item Type: FIN Finding FAILURE TO MEET SINGLE FAILURE CRITERIA FOR RESIDUAL HEAT REMOVAL VALVE CIRCUITRY.
The licensee identified that the circuitry associated with the residual heat removal system discharge to safety injection system suction isolation valves did not meet single failure criteria. The inspectors noted that this design requirement was identified in the facility's updated safety analysis report. The licensee subsequently implemented a temporary change to the facility. The inspectors reviewed the issue and identified that the facility had been operating outside of its design basis, which was reportable to the NRC. The licensee subsequently made a one hour non-emergency report to the NRC. Since there was no actual loss of safety function to the system, this issue was screened as very low risk significance.
Inspection Report# : 2000008(pdf)
Significance:          Jun 22, 2000 Identified By: NRC Item Type: FIN Finding MAINTENANCE RULE FAILURES ASSOCIATED WITH REACTOR HEAD VENT VALVE.
The inspectors reviewed the licensee's implementation of the maintenance rule for failures associated with a reactor head vent valve. The licensee's corrective action documents identified a potential maintenance rule functional failure but the completed evaluation of the problem did not document the final determination. However, the inspectors identified that the repeated failures may have been prevented if maintenance activities such as valve disassembly and cleaning had been performed. In this case, maintenance rule reliability goals were not exceeded. The licensee had documented similar maintenance rule program deficiencies and developed a corrective action program to address the deficiencies. Although programmatic deficiencies exist, since no maintenance rule reliability criteria had been exceeded, this issue was considered of very low risk significance.
 
3Q/2001 Inspection Findings - Kewaunee                                                                                                    Page 6 of 11 Inspection Report# : 2000008(pdf)
Significance:        Jun 22, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO INITIATE KEWAUNEE ASSESSMENT PROCESS DOCUMENT REGARDING REFUELING WATER STORAGE TANK LOW-LOW LEVEL ALARM INOPERABILITY.
The inspectors identified that the refueling water storage tank low-low level alarm which was actuating five percent higher than normal had not been documented in a Kewaunee Assessment Process form by the licensee, and therefore had not received an operability evaluation. This failure was identified as contrary to site administrative procedures. Following the licensee's documentation of the problem, the inspectors identified that the associated operability evaluation considered the acceptability of an operator workaround to address the issue, but did not address any safety implications or consequences of the alarm actuating early. A subsequent operability evaluation by the licensee was evaluated as adequate by the inspectors. Since the subsequent operability evaluation was adequate and it was determined that no safety mitigation equipment was adversely affected by the early actuation of the alarm, this issue was considered of very low risk significance. A non-cited violation (NCV) was identified for failing to document a non-conforming condition, contrary to site administrative procedure requirements.
Inspection Report# : 2000008(pdf)
Significance:        Jun 22, 2000 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO TEST ADDITIONAL RELIEF VALVES IN ACCORDANCE WITH TECHINCAL SPECIFICATIONS.
The licensee identified that the suction relief valve for an auxiliary feedwater pump may have failed its relief test criteria, but did not process the documented deficiency until several weeks later. The licensee then expanded the scope of the relief testing to the suction relief valves associated with the other auxiliary feedwater pumps to meet technical specification requirements. Since any one train of auxiliary feedwater was capable of supplying 100 percent of the decay heat removal requirements, this issue was screened as very low risk significance. However, the time delay in complying with technical specification requirements for testing other relief valves was identified as an NCV.
Inspection Report# : 2000008(pdf)
Significance:        May 22, 2000 Identified By: NRC Item Type: FIN Finding FIRE EXTINGUISHERS NOT LOCATED IN ALL AREAS OF CONTAINMENT BUILDING AT BEGINNING OF REFUELING OUTAGE.
During a walkdown of the containment building, the inspectors identified that portable fire extinguishers were not located in the containment basement at the beginning of the plant refueling outage. Additionally, site fire protection procedures required that responsible fire protection personnel perform inspections of selected plant areas to ensure that the quantity of combustible material was minimized. However, the procedure did not list the containment as an area to be inspected and the procedures did not require the placement and location of portable fire suppression equipment inside containment during the refueling outage. Due to a low number of work activities ongoing at the time, this issue was screened as Green (very low risk significance).
Inspection Report# : 2000007(pdf)
Significance:        May 22, 2000 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO INSTALL RAYCHEM HEAT SHRINK MATERIAL IN ACCORDANCE WITH PROCEDURE REQUIREMENTS.
The licensee identified that two Raychem electrical cable splices utilized in environmentally qualified (EQ) safety-related equipment had not been installed in accordance with EQ requirements. These splices were associated with pressurizer level transmitters and were installed in 1984. The licensee subsequently performed extensive EQ testing of the splices to determine the qualification of the splices' as-found configurations. Test results indicated that the splices would have been able to perform their intended function in a harsh environment inside containment. This issue was considered to be of low safety significance based on the successful EQ testing of the as-found splices' configurations and was screened as Green (very low risk significance). One non-cited violation was identified.
Inspection Report# : 2000007(pdf)
Barrier Integrity
 
3Q/2001 Inspection Findings - Kewaunee                                                                                                  Page 7 of 11 Significance:        Jun 22, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation FAILURE TO UPDATE COMPUTER ALARM FOR CURRENT AXIAL FLUX DISTRIBUTION TARGET BAND.
The licensee identified, following plant startup, that a computer alarm had not been updated properly to alarm if axial flux distribution deviated outside of the flux distribution target band. This condition was contrary to technical specification requirements. The licensee reviewed the axial flux distribution history since the startup and determined that at no time was the flux distribution outside of the target band. Since the axial flux distribution was never outside of the target band, this issue was screened as very low risk significance. An NCV was identified for failing to comply with technical specification requirements for monitoring axial flux distributions.
Inspection Report# : 2000008(pdf)
Significance:        May 22, 2000 Identified By: Licensee Item Type: FIN Finding TECHNICAL SPECIFICATIONS INTERPRETATION FOR TESTING REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES.
The NRC determined that the licensee's practice of testing reactor coolant system pressure isolation Valve SI-22B prior to entering the cold shutdown condition was contrary to Technical Specification requirements 4.2.a.3.a. Technical Specification 4.2.a.3.a required that periodic leakage testing of Valve SI-22B be accomplished prior to reaching operating mode after the plant was placed in cold shutdown. This issue was considered to be of low safety significance because of a subsequent successful valve test and was screened as Green (very low risk significance).
Enforcement discretion was applied to this item in accordance with Section VII.B.6 of the Enforcement Policy.
Inspection Report# : 2000007(pdf)
Significance:        May 19, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PROPERLY IMPLEMENT FLAW ACCEPTANCE CRITERIA FOR SLEEVE WELD INSPECTIONS.
During 1998 inservice inspection examinations, the licensee failed to properly implement the flaw acceptance criteria for laser welded sleeve inspection within two steam generator tubes. The safety significance was very low based on the absence of adverse consequences, and May 2000 in-situ pressure testing where both welds exhibited zero leakage at normal operating pressure, main steam line break pressure, and three times normal operating differential pressures. As such, this issue was characterized as a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion IX, Control of Special Processes. To correct the error, the licensee plugged both tubes (Section 1RO8).
Inspection Report# : 2000011(pdf)
Emergency Preparedness Significance:        Mar 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation SUPPLEMENTAL INSPECTION OF WHITE ERO AUGMENTATION FINDING AND RESULTING GREEN FINDING.
This supplemental inspection was performed by the NRC to evaluate the licensees evaluation associated with the failure to conduct successful quarterly, off-hours, unannounced staff augmentation drills during the second, third, and fourth quarters of 1999 and the second quarter of 2000.
This performance issue was previously characterized as having low to moderate risk significance (White) in NRC Inspection Report No. 50-305/2000015(DRS). During this supplemental inspection, performed in accordance with Inspection Procedure95002, the inspector concluded that the licensee performed a comprehensive evaluation of the unsuccessful staff augmentation drills. The licensees evaluation identifiedtwo root causes which resulted in the unsuccessful drills and in the staffs inability tocorrect thedeficiencies: (1) Management has not effectively acted to provide increased depth and flexibility in the emergency response organization following a reduction in staffing several years ago; and (2)
Management has accepted an adverse trend of test failures without requiring investigation into the root causes. The inspector reviewed the licensees corrective actions, both completed and planned, and concluded that the programmatic corrective actions appeared to address the identified root causes. In particular, the licensee assigned certain positions to an on-call rotation to ensure personnel were capable of augmenting in a timely manner, and the licensee was progressing in training additional staff to increase the depth of personnel assigned to key emergency response positions. In addition, the licensee was continuing its efforts in improving its corrective action program. The inspector reviewed the licensees immediate response to the issue and identified that one of the licensees initial corrective actions resulted in a Non-Cited Violation of regulatory requirements. To obtain a timely response of a key emergency response position (severe accident management - core hydraulics), the emergency preparedness staff effectively changed the emergency plan without revising the necessary procedures and without formally assessing the impact of that change. The staff instructed and trained personnel to respond to a location other than the Technical Support Center, which was
 
3Q/2001 Inspection Findings - Kewaunee                                                                                                    Page 8 of 11 contrary to the licensees current implementing procedures. While this change enabled the licensee to augment its staff in a timely manner, the change was not performed in accordance with NRC requirements. In order to make such a change, the licensees emergency plan required that the change be formally assessed to ensure that it did not reduce the effectiveness of the plan or any other implementing procedure. Since the issue did not result in a failure to meet an emergency preparedness planning standard, the failure to adequately implement the emergency plan was determined to be a violation of very low safety significance (Green) (Section 02.3(a)). Due to the licensees acceptable performance in assessing the emergency response augmentation drill deficiencies, the White finding associated with this issue will only be considered in assessing plant performance for a total of four quarters in accordance with the guidance in IMC 0305, Operating Reactor Assessment Program. Implementation of the licensees corrective actions will be reviewed during a future inspection.
Inspection Report# : 2001007(pdf)
Significance:          Mar 30, 2001 Identified By: NRC Item Type: VIO Violation FAILURE TO CORRECT SELF-IDENTIFIED ERO AUGMENTATION DRILL DEFICIENCIES.
During a baseline inspection of the emergency preparedness program conducted on August 14 - September 21, 2000, the NRC identified a preliminary White issue and potential violation for the licensees failure to successfully correct deficiencies identified during staff augmentation drills and to demonstrate timely staff augmentation in 1999 and 2000. The issue was unresolved pending the outcome of the NRCs final significance determination. On January 30, 2001, the NRC conducted a regulatory conference with the licensee and subsequently issued the licensee a White finding and Notice of Violation (Enforcement Action No. 00-214) associated with the performance issue.
Inspection Report# : 2001007(pdf)
Significance:          Mar 23, 2001 Identified By: NRC Item Type: NCV NonCited Violation POST-ACCIDENT SAMPLING SYSTEM CONTAINMENT AIR SAMPLE PANEL TESTING.
A Non-Cited Violation of Technical Specification 6.14 was identified for the failure to implement a program that ensured the capability to obtain and analyze containment atmosphere samples under accident samples using the cantainment air sampling panel (CASP). Although the CASP was installed, as was indicated in the emergency plan, the licensee had neither developed procedures nor had tested its capability to obtain a containment atmosphere sample using the CASP. The licensee could not recall if and when containment air samples were last obtained using the CASP.
Inspection Report# : 2001006(pdf)
Significance: N/A Feb 27, 2001 Identified By: NRC Item Type: FIN Finding SUPPLEMENTAL INSPECTION OF YELLOW ANS PERFORMANCE INDICATOR AND ASSOCIATED CORRECTIVE ACTION PROGRAM DEFICIENCIES.
This supplemental inspection was performed by the NRC to assess the licensee's evaluation associated with a Yellow performance indicator for the Alert and Notification System (ANS) and the associated Yellow finding related to the licensee's corrective action program. During this supplemental inspection, performed in accordance with Inspection Procedure 95002, the inspectors concluded that the licensee performed comprehensive evaluations of the performance problems associated with the ANS and its corrective action program. These evaluations identified primary root causes and contributing causes for both issues. Along with the electronics/hardware problems, the licensee identified the primary root causes for the ANS to be the failure to make changes to the system via a change control process and the failure of the activation procedure to provide for alternate/backup activation methods and to provide clear success criteria. In the case of the corrective action program, the licensee concluded that the primary root causes were plant management's inadequate risk evaluation regarding decisions affecting the corrective action program and the development of a culture in the licensee's organization, which minimized the importance of information from outside organizations. In particular, the licensee focused on low-cost power operation and failed to recognize the value of changes and improvements which had occurred throughout the industry, such as the value and expansion of the corrective action program. In the case of the Yellow ANS performance indicator, the licensee had completed several significant corrective actions to address the root causes and contributing causes identified in its evaluation. The inspectors found that the corrective actions appeared appropriate to address the underlying root causes and that ANS testing data indicated an improving trend in the NRC performance indicator. The licensee also performed comprehensive assessments of the emergency preparedness program, quality assurance program, plant operations, and other plant programs to determine the extent of condition (re. the root causes described above). Based on these evaluations, the licensee began to implement significant actions to correct the deficiencies in the corrective action program and other weaknesses identified. Generally, the inspectors observed progress in the licensee's initial implementation of these corrective actions. Due to the licensee's acceptable performance in assessing the Yellow ANS performance indicator and the associated Yellow finding, the Yellow finding will not be considered in assessing future plant performance.
Inspection Report# : 2001005(pdf)
Significance:          Sep 21, 2000 Identified By: NRC
 
3Q/2001 Inspection Findings - Kewaunee                                                                                                  Page 9 of 11 Item Type: FIN Finding INADEQUATE ROOT CAUSE EVALUATION FOR YELLOW ALERT AND NOTIFICATION SYSTEM PERFORMANCE INDICATOR.
The licensee's evaluation of the Yellow Alert and Notification (siren) System Performance Indicator (PI) was inadequate. The inspector concluded that the licensee's evaluation was not performed at the depth necessary to identify the root causes of the siren performance problems and, instead, only identified the symptoms of the root causes. Specifically, the inspector identified the following substantive weaknesses in the licensee's evaluation of the siren system performance, which appeared to result from systemic corrective action program deficiencies within this cornerstone:
* The licensee's evaluation was not of sufficient depth to clearly identify the root causes associated with the decline in siren system performance.
* Licensee management did not provide well-understood and clear guidance/expectations for performing root cause evaluations.
* The licensee's evaluation of the quality assurance program was narrowly focused and was not critical of its role in failing to identify and correct the siren performance problems.
* The licensee did not establish a priority for each of the long-term corrective actions in accordance with the associated significance or risk.
* The licensee did not have any formal provisions for measuring the effectiveness of its corrective actions.
* Within the licensee's evaluation, the licensee had not evaluated common causes or the extent of the condition. Due to the corrective action program performance deficiencies within this cornerstone, we have been unable to conclude that the performance issues that resulted in the yellow PI have been addressed. Therefore, we are issuing a yellow finding that corresponds to the original issues that resulted in a yellow PI. Additional inspection effort will be focused on the licensee's further evaluation of the siren reliability root causes and the continuing corrective action program implementation deficiencies identified during this inspection.
Inspection Report# : 2000017(pdf)
Significance: N/A Apr 05, 2000 Identified By: NRC Item Type: FIN Finding LICENSEE FAILED TO IDENTIFY THE FULL SCOPE OF PROBLEMS WITH THE ALERT AND NOTIFICATION SYSTEM PERFORMANCE.
The inspectors concluded that the licensee's assessment was not sufficiently comprehensive to identify the full scope of problems associated with the Alert and Notification System (ANS) performance program. As a result, licensee corrective actions generally were focused on the equipment problem rather than all root causes and contributing factors. For example, the inspectors identified that: (1) management oversight of the ANS performance program was limited; (2) an audit failed to identify degrading ANS performance as a concern; (3) annual preventive maintenance was not consistently performed on the system; (4) the corrective action program was not used consistently to document ANS problems; and (5) maintenance procedures and records were deficient. Collectively, these problems indicate that the ANS performance program lacked sufficient structure and oversight.
Inspection Report# : 2000006(pdf)
Occupational Radiation Safety Significance:        Oct 02, 2001 Identified By: NRC Item Type: NCV NonCited Violation HIGH RADIATION AREA ACCESS CONTROLS Non-Cited Violation of Technical Specification 6.13 and an associated Green Finding for failure to 'barricade' three ladders that provided entry to high radiation areas (less than 1000 mrem/hour) located on the steam generator/pressurizer platforms.
Inspection Report# : 2001014(pdf)
Significance:        May 19, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO POST A VERY HIGH RADIATION AREA.
The inspectors identified a noncited violation for the failure to post a very high radiation area in accordance with 10 CFR 20.1902(c). Although the area was not adequately posted, the licensee had provided physical controls and barriers that were consistent with its requirements for a very high radiation area. Based on the adequacy of these controls, the potential for an overexposure from the inadvertent entry of personnel into the area was low. Consequently, this finding was determined to be of very low safety significance (Section 20S1.1).
Inspection Report# : 2000009(pdf)
Public Radiation Safety
 
3Q/2001 Inspection Findings - Kewaunee                                                                                                Page 10 of 11 Physical Protection Significance: N/A Jul 21, 2000 Identified By: NRC Item Type: FIN Finding CORRECTIVE ACTION ON SEARCH ISSUES DID NOT WORK.
The inspector determined that the licensee's effectiveness of implemented corrective actions for a previously identified inspection finding regarding an inadequate vehicle search was not totally effective in preventing recurrence. Previous corrective action was not adequately focused (Section 3PP2.2).
Inspection Report# : 2000013(pdf)
Significance:        Jul 21, 2000 Identified By: NRC Item Type: NCV NonCited Violation SEARCHES INADEQUATE (REPEAT).
The inspector identified a Non-Cited violation by observing that, a security officer failed to search an easily accessible compartment on one vehicle.
The failure resulted from human error because the officer did not observe the access panel to the compartment during the vehicle search process.
When searched, no prohibited items were found. Corrective actions were implemented. The inspector determined that the licensee's effectiveness of implemented corrective actions for a previously identified inspection finding regarding an inadequate vehicle search was not totally effective in preventing recurrence. Previous corrective action was not adequately focused.
Inspection Report# : 2000013(pdf)
Miscellaneous Significance: N/A Aug 24, 2001 Identified By: NRC Item Type: FIN Finding IDENTIFICATION AND RESOLUTION OF PROBLEMS The team concluded that the licensee was generally effective at identifying problems and putting them into the corrective action program. The program itself contained all the necessary attributes of an acceptable corrective action program and was generally successful in correcting identified issues. However, the team noted that, although licensee management had taken efforts to ensure that issues were resolved in accordance with program guidance and requirements, additional efforts appeared necessary to ensure timely resolution of issues. A positive program initiative was the establishment of positions in each of the major plant departments to serve as liaisons between the departments and the corrective action program and to assist with self-assessments. However, examples were identified by the inspectors of problems with the licensee's identification and resolution of problems, prioritization and evaluation of issues, and the effectiveness of corrective actions. Included in these examples were the routine granting of due date extensions for problem evaluation and corrective action implementation, failure to perform a Technical Specification-required biennial surveillance of safety-related procedures, and ineffective corrective actions that resulted in the lack of proper controls over the storage of grease used in safety-related breakers. Based on a review of records and discussions with plant staff, the inspectors concluded that workers at the site felt free to input safety issues into the corrective action program.
Inspection Report# : 2001012(pdf)
Significance: N/A Nov 09, 2000 Identified By: NRC Item Type: FIN Finding FAILURE TO TAKE ADEQUATE CORRECTIVE ACTIONS TO ENSURE COMPONENT RE-TEST REQUIREMENTS COMPLETED ACCORDING TO MAINTENANCE PROCEDURES.
No Color. The inspectors determined that a negative performance trend had developed in the licensee's ability to identify and promptly take appropriate corrective actions to prevent recurrence based on two previously identified examples (NCV 50-305/2000014-01) and one example identified during this inspection period (NCV 50-305/00-20-01). All three examples related to the licensee returning safety-related equipment to service prior to completing all required post-maintenance retesting. While the risk of the individual examples was very low, the licensee had failed to ensure that all retest requirements had been completed before returning safety-related equipment to service. These findings collectively indicated a problem with the licensee's ability to provide timely and adequate corrective actions to prevent recurrence. (Section 4OA2).
Inspection Report# : 2000020(pdf)
Significance: N/A Feb 21, 2002 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Implement Required Fire Watch Following Completion of Hot Work Activities A licensee-identified violation was reviewed by the inspectors. Corrective actions taken or planned by the licensee appeared reasonable.
Inspection Report# : 2001017(pdf)
 
3Q/2001 Inspection Findings - Kewaunee                                                                                                Page 11 of 11 Significance: N/A Sep 29, 2000 Identified By: NRC Item Type: FIN Finding EFFECTIVENESS OF CORRECTIVE ACTION PROGRAM.
Based on the results of this inspection, the NRC concluded that the corrective action program at Kewaunee showed significant weaknesses and inconsistencies across all of the procedural elements inspected. These weaknesses existed across departments and affected multiple cornerstones in the strategic performance areas of Reactor Safety, Radiation Safety, and Safeguards. Of particular note was the lack of procedures for determining the significance of conditions adverse to quality and for trending of issues and the complete lack of trending within your corrective action program. We also identified a lack of urgency in correcting issues which resulted in repeat examples occurring and, coupled with a poor tracking system, a tendency for issues to be dropped. While none of the specific examples identified by the team were of high risk significance when looked at in isolation, in the aggregate they were similar in nature to prior issues in the emergency preparedness area that rose to a higher significance level and contributed to a degraded cornerstone. While we concluded that the station had fostered an environment in which personnel freely identified conditions adverse to quality without fear of discrimination or retaliation, we also concluded that significant weaknesses with, and inconsistent implementation of, the station corrective action program resulted in multiple examples where station personnel did not enter deficiencies into the station's formal corrective action program.
Inspection Report# : 2000019(pdf)
Significance: N/A Sep 29, 2000 Identified By: NRC Item Type: NCV NonCited Violation NO PROCEDURAL GUIDANCE FOR DETERMINING IF CONDITIONS ADVERSE TO QUALITY ARE SIGNIFICANT.
The inspectors reviewed the quality assurance (QA) manual requirements against Kewaunee's implementing procedures and identified that two QA manual requirements were not being implemented. Specifically, Program Requirement 3.1.9 which stated that directives and procedures shall provide for the review of conditions adverse to quality to determine if the conditions are significant in nature. This requirement paralleled 10 CFR Part 50, Appendix B, Criterion XVI, which requires that the cause of significant conditions adverse to quality be determined and corrective actions taken to prevent recurrence. The inspectors reviewed the Nuclear Administrative Directive (NAD 11.08) and the procedure (GNP 11.08.01) governing the KAP (Kewaunee corrective action program) process and found no procedure requirements for identifying significant conditions adverse to quality. This finding does not directly affect a cornerstone. As a result, this issue was not evaluated with the Significance Determination Process and was not assigned a color. One example of a non-cited violation of 10 CFR Part 50, Appendix B, Criterion II, "Quality Assurance Programs," was identified.
Inspection Report# : 2000019(pdf)
Significance: N/A Sep 29, 2000 Identified By: NRC Item Type: NCV NonCited Violation NO PROCEDURAL GUIDANCE FOR TRENDING CONDITIONS ADVERSE TO QUALITY.
The inspectors reviewed the quality assurance (QA) manual requirements against Kewaunee's implementing procedures and identified that two QA manual requirements were not being implemented. Specifically, QA Program Requirement, 3.1.10, stated that directives and procedures shall provide for analyzing trends of conditions adverse to quality. Once identified these trends were required to be considered significant conditions adverse to quality. The inspectors found that conditions adverse to quality were not defined in the KAP (Kewaunee corrective action program) procedures and that no procedure existed for trending. This finding does not directly affect a cornerstone. As a result, this issue was not evaluated with the Significance Determination Process and was not assigned a color. One example of a non-cited violation of 10 CFR Part 50, Appendix B, Criterion II, "Quality Assurance Program," was identified.
Inspection Report# : 2000019(pdf)
Significance: N/A Jun 22, 2000 Identified By: NRC Item Type: FIN Finding CONTROL ROOM OPERATIONS HUMAN PERFORMANCE ISSUES.
The inspectors interviewed operators to evaluate their awareness of degraded control room indications and alarms, and their ability to adequately take manual actions based on degraded alarm functions. The inspectors identified, during interviews, that there was a lack of awareness by operators of a degraded refueling water storage tank low-low level alarm which would be potentially confusing to operators and therefore increase the risk associated with initiating long term sump recirculation.
Inspection Report# : 2000008(pdf)
Last modified : March 26, 2002
 
4Q/2001 Inspection Findings - Kewaunee                                                                                                  Page 1 of 10 Kewaunee Initiating Events Mitigating Systems Significance: N/A Sep 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ESTABLISH CONTINGENCY PLANS FOR ORANGE RISK CONDITION The inspectors identified the failure to establish contingency plans during a planned high risk plant configuration. Contrary to administrative requirements, the licensee approved an orange risk condition during a refueling outage with no contingency plans to mitigate the consequences of a loss of spent fuel pool cooling with a full core offload in the pool. A Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was identified. The finding was of very low safety significance because although the licensee had not approved appropriate contingency actions for the orange risk condition, the licensee subsequently rescheduled the planned maintenance to eliminate the orange risk condition.
Inspection Report# : 2001013(pdf)
Significance: N/A Aug 24, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO CONDUCT BIENNIAL SURVIELLANCE OF SAFETY-RELATED PROCEDURES PER TS 6.8.c A Non-Cited Violation of Technical Specification 6.8.c was identified for the failure to perform a biennial surveillance of safety-related procedues.
This issue was more than minor because if left uncorrected, could under the same condition become a more significant safety concern. However, since no specific cornerstone had been impacted, this finding is designated as No Color.
Inspection Report# : 2001012(pdf)
Significance:          Aug 24, 2001 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE CORRECTIVE ACTIONS FOR PREVIOUS PROBLEM WITH THE CONTROL OF SAFETY-RELATED MATERIALS A Non-Cited Violation of Criterion XVI, "Corrective Action," of Appendix B of 10 CFR Part 50 was identified for ineffective corrective actions for a problem with the control of the storage of consumable materials, such as thread sealant, used in safety-related applications. These ineffective actions subsequently resulted in the inadequate control of the storage of grease used in safety-related breakers. This issue was more than minor because if left uncorrected, could under the same condition become a more significant safety concern. In that this issue could credibly affect the operability, availability, reliability, or function of a system or train in a mitigating system, it is a Green finding.
Inspection Report# : 2001012(pdf)
Significance:          Aug 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation NON-RATED FIRE BARRIER A Non-Cited Violation [of 10 CFR Part 50, Appendix R, Section III.G.2.a] was identified for failure to provide a 3-hour rated fire barrier to separate redundant trains of safe shutdown equipment. This finding was of very low safety significance because the licensee tested a replica of the fire barrier and demonstrated that the fire barrier provided protection for at least 60 minutes, which was sufficient for the hazards in the area.
Inspection Report# : 2001011(pdf)
Significance:          Jul 20, 2001 Identified By: NRC Item Type: FIN Finding LICENSED OPERATOR REQUALIFICATION EXAMINATION RESULTS The inspectors identified that two of eight crews examined during the licensee's calendar year 2001 licensed operator requalification operating test had failed. The finding was of very low safety significance because both crews that had failed received remedial training prior to being returned to shift, and the results of the licensee's operator licensing requalification operating test given in calendar year 2000 indicated that only one crew, out of a total of eight crews tested, had failed.
Inspection Report# : 2001011(pdf)
Significance: N/A Jun 30, 2001 Identified By: NRC
 
4Q/2001 Inspection Findings - Kewaunee                                                                                                  Page 2 of 10 Item Type: FIN Finding FAILURE TO BALANCE RELIABILITY AND AVAILABILITY AS REQUIRED BY 10 CFR 50.65(a)(3)
The inspectors identified a failure to evaluate whether adjustments were necessary such that there would be an appropriate balance between systems' availability and reliability in accordance with 10 CFR 50.65(a)(3) of the maintenance rule. The inspectors identified that the licensee did not have an administrative process to track maintenance rule functional failures and maintenance preventible maintenance functional failures. As a result, reliability and availability could not be balanced as required by the Maintenance Rule periodic evaluation. The safety significance of the specific finding was very low because it did not affect the operability of the systems, and the licensee entered the finding in the corrective action program. However, this finding was considered to be of regulatory concern in the area of maintenance rule implementation due to the extent of the problems with the Maintenance Rule Program.
Inspection Report# : 2001009(pdf)
Significance:          Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IDENTIFY DEFICIENT CONDITION OF VALVE AFW-1B The inspectors identified that the licensee failed to promptly identify and correct the B' train auxiliary feedwater pump discharge check valve which was stuck in an intermediate position. A Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was issued. The finding was of very low safety significance because, although the check valve was stuck in an intermediate position, the time that it was known to have been stuck was less than the technical specification allowed outage time for one train of auxiliary feedwater to be out of service (less than 72 hours). Additionally, the other two trains of auxiliary feedwater were each capable of 100 percent decay heat removal.
Inspection Report# : 2001009(pdf)
Significance:          Jun 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO TRACK UNAVAILABILITY OF SYSTEMS REQUIRED DURING SHUTDOWN OPERATION 10 CFR 50.65(a)(1), required, in part, that the licensee monitor the performance or condition of SSCs within the scope of the rule as defined by 10 CFR 50.65(b), against licensee-established goals, in a manner sufficient to provide reasonable assurance that such SSCs are capable of fulfilling their intended functions. 10 CFR 50.65(a)(2) stated, in part, that monitoring as specified in 10 CFR 50.65(a)(1) was not required where it had been demonstrated that the performance or condition of an SSC was being effectively controlled through the performance of appropriate preventive maintenance, such that the SSC remained capable of performing its intended function. Contrary to the above, the licensee failed to demonstrate that the performance or condition of systems required to be available during shutdown conditions and within the scope of the rule had been effectively controlled through the performance of appropriate preventive maintenance and did not monitor against licensee-established goals.
Specifically, the licensee failed to monitor the unavailability of systems required during shutdown operation.
Inspection Report# : 2001009(pdf)
Significance:          Jun 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO ESTABLISH MAINTENANCE RULE (a)(1) GOALS 10 CFR 50.65(a)(1), requires, in part, that the licensee monitor the performance or condition of SSCs within the scope of the rule as defined by 10 CFR 50.65(b), against licensee-established goals, in a manner sufficient to provide reasonable assurance that such structures, systems, and components, are capable of fulfilling their intended functions. Such goals shall be established commensurate with safety. When the performance or condition of a structure, system, or component does not meet established goals, appropriate corrective action shall be taken. Contrary to the above, from 1996, the licensee did not take appropriate corrective actions when the performance of those systems in (a)(1) did not meet licensee established goals. Specifically, the licensee determined timely and appropriate corrective actions had not been taken for five systems that had been in (a)(1) category for approximately 3 years to 5 years: component cooling (entered (a)(1) on April 23, 1997), control room air conditioning (July 24, 1996), station and instrument air (July 3, 1997), auxiliary building air ventilation (July 31, 1997), and control rod drive (August 6, 1998). This issue is in the licensee's corrective action system as KAP WO 01-3323. The inspectors evaluated the risk significance of this issue using the Significance Determination Process. The inspectors did not identify where this failure resulted in a total loss of a risk significant SSC. Therefore, this issue was screened as Green (very low risk significance) after a Phase 1 Significance Determination Process review. Although the risk significance of this issue was low, the inspectors concluded that this was more than a minor concern because the failure to recognize and correct ineffective maintenance practices resulted in risk significant systems in (a)(1) for years with no improvement in performance. The NRC tracking number for this issue is 50-305/01-09-02.
Inspection Report# : 2001009(pdf)
Significance:          Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation FIRE DOOR FUSIBLE LINKS.
On February 20, 2001, the licensee determined that the installed fusible link arrangement on roll-up fire Doors 279 and 281, which separated both trains of service water pumps, would not actuate as designed to ensure that the doors would automatically close to provide a 3-hour fire barrier, contrary to 10 CFR Part 50, Appendix R, Section III.G.2.a which required, in part, separation of cables and equipment of redundant trains by a fire barrier having a 3-hour rating.
Inspection Report# : 2001006(pdf)
 
4Q/2001 Inspection Findings - Kewaunee                                                                                                      Page 3 of 10 Significance:        Mar 12, 2001 Identified By: Licensee Item Type: NCV NonCited Violation INADEQUATE SMOKE DETECTOR COVERAGE IN FIRE ZONE TU-95B.
Licensee identified violation of licensee's operating license that the licensee failed to install a detector in each beam pocket in Fire Zone TU-95B.
Inspection Report# : 2001002(pdf)
Significance:        Feb 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO TEST FIRE DOOR IN ACCORDANCE WITH FIRE PLAN.
The inspectors identified a non-cited violation for failure to properly test a fire door in accordance with the facility's fire protection program plan. The finding was of very low safety significance because, although the fire door separated both trains of service water pumps and did not fully close as designed when subsequently tested, the fire loading in the area was insufficient to credibly impact more than two of the four service water pumps in the area.
Inspection Report# : 2001004(pdf)
Significance: N/A Nov 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO COMPLETE COMPONENT RE-TEST REQUIREMENTS PRIOR TO RETURNING EQUIPMENT TO AN OPERABLE STATUS.
No Color. The inspectors identified a Non-Cited Violation for failure to complete component retest requirements following maintenance performed on the B train control room air conditioner compressor condenser. The unit had been returned to an operable status prior to the retest requirements being completed as prescribed in the associated maintenance procedure. This issue was determined to be a violation of the licensee's Operational Quality Assurance Program Manual, Section 8, "Maintenance Planning and Control." Although the risk associated with this finding was very low and did not affect any cornerstones, the inspectors noted that this finding was similar to previous NRC-identified findings and therefore was of greater than minor significance and warranted documentation. (Section 1R19).
Inspection Report# : 2000020(pdf)
Significance:        Sep 29, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IDENTIFY CORROSION AS POTENTIAL FAILURE MECHANISM.
The inspectors identified that the licensee failed to identify corrosion as a potential failure mechanism in the operability determination for a carbon steel key in the service water system. Thus, the licensee failed to quantify the corrosion rate and therefore did not adequately evaluate the expected service life of the carbon steel key. One non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was identified.
Inspection Report# : 2000019(pdf)
Significance:        Sep 29, 2000 Identified By: NRC Item Type: NCV NonCited Violation PRESSURE RATING OF AUXILIARY FEEDWATER STEAM TRAPS.
The inspectors identified that a root cause evaluation for a 1996 equipment issue in the turbine-driven auxiliary feedwater system was not completed until 1999. The evaluation stated that the internals of the steam traps were designed to operate at pressures up to a maximum 600 psig but that the traps were exposed to pressures up to 1025 psig. A corrective action item to initiate a design change request to replace the steam traps with a different model rated for the design pressure of the system was described in the evaluation. However, the inspectors identified that the design change request had never been initiated and the KAP ( Kewaunee corrective action document) had been closed. As a result, the corrective action item for this design problem was lost. In addition, operability of the system had never been formally evaluated despite the identification that the system design requirements were not met. The licensee subsequently determined that the steam traps remained operable and was planning to initiate the design change to correct the problem. One non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was identified.
Inspection Report# : 2000019(pdf)
Significance:        Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation DESIGN CONTROL VIOLATION FOR AUXILIARY FEEDWATER STRAINER MESH SIZE.
The inspectors questioned the mesh size of the strainers, which were installed in the suction of the three auxiliary feedwater (AFW) pumps. As a result of the inspectors' questions, license personnel inspected the strainers on August 21, 2000, and found the strainers to have 1/16 inch openings. A note was later found on Figure 10.2-3 of the UFSAR that indicated that the AFW suction strainer size was 1/8 inch. The smaller openings would not support the use of service water as a safety related source for AFW and as a result all three trains of AFW were declared
 
4Q/2001 Inspection Findings - Kewaunee                                                                                                Page 4 of 10 inoperable. This condition had apparently existed for approximately 25 years and was identified as a non-cited violation of Criterion III, "Design Control," of 10 CFR 50, Appendix B.
Inspection Report# : 2000012(pdf)
Significance: N/A Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation RETRIEVAL OF SERVICE WATER SYSTEM DESIGN INFORMATION.
In many cases, design basis information for the service water system was difficult if not impossible to locate. Licensee personnel wrote KAP WO 00-002566 to enter the problem in the corrective action program.
Inspection Report# : 2000012(pdf)
Significance:          Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PROPERLY CONTROL CORRECTED TEST DATA NECESSARY FOR DESIGN CALCULATIONS.
An example of a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," was identified in the handling of service water system flow test data, which was subsequently used in calculations. Gauge readings corrected for post test calibration checks, gauge reading corrections for elevation considerations, and flow values corrected for pump degradation were contained in spreadsheets in the possession of an individual staff member, but not currently packaged with raw test data, and not bearing evidence of a formal review and control process. The connection between the test data, which had been vaulted, and the values used in the calculation, could not be made without use of the uncontrolled spreadsheet.
Inspection Report# : 2000012(pdf)
Significance:          Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation IMPROPER DESIGN CALCULATION IDENTIFICATION, NON-CONSERVATIVE ASSUMPTIONS, CALCULATION ERRORS, AND DUPLICATE CALCULATIONS.
An example of a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," was identified because of inadequate control of design calculations. The control failures included improper identification of calculations, non-conservative assumptions, calculation errors, and duplicate or superceded calculations not properly identified or canceled. The failure to follow the established design control process increased the potential for errors in the design and operation of the service water system. Because the system was subsequently demonstrated to be capable of removing the design heat load, the actual significance was low and this finding screened out as having very low risk significance.
Inspection Report# : 2000012(pdf)
Significance:          Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO DOCUMENT IN THE CORRECTIVE ACTION PROGRAM THE USE OF INCORRECT MATERIAL IN A SERVICE WATER PUMP KEY.
An example of a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was identified because of inadequate corrective action to correct an incorrect coupling adjust nut set screw and a low strength "soft" key material, which had contributed to a pump shaft failure. Licensee personnel had known of the "soft" key material since July 21, 1999. The "soft" key material was found in other service water pumps but had not been removed from all pumps. As of July 25, 2000, licensee personnel had not documented the existence of the "soft" key material in the corrective action program.
Inspection Report# : 2000012(pdf)
Significance:          Jul 07, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO COMPLETE COMPONENT RETEST REQUIREMENTS IN ACCORDANCE WITH PROCEDURE.
On June 26, 2000, during a review of post maintenance testing requirements following maintenance performed on the control room post accident system charcoal filter heat detector, the inspectors identified that maintenance technicians had not completed the component re-test requirements, as required by a preventative maintenance procedure prior to the system being returned to an operable status. On July 7, the inspectors identified a second example of failing to complete component re-test requirements following maintenance on the zone special ventilation system charcoal filter heat detector as required. The issue was considered to be of very low safety significance based on the determination that although the licensee had not completed all of the component retest requirements prior to returning the equipment to service, the subsequent testing determined that the equipment was in an operable status. The failure to complete the component retest requirements in accordance with site procedures was identified as a Non-Cited Violation.
Inspection Report# : 2000014(pdf)
 
4Q/2001 Inspection Findings - Kewaunee                                                                                                    Page 5 of 10 Significance:        Jun 22, 2000 Identified By: NRC Item Type: FIN Finding FAILURE TO MEET SINGLE FAILURE CRITERIA FOR RESIDUAL HEAT REMOVAL VALVE CIRCUITRY.
The licensee identified that the circuitry associated with the residual heat removal system discharge to safety injection system suction isolation valves did not meet single failure criteria. The inspectors noted that this design requirement was identified in the facility's updated safety analysis report. The licensee subsequently implemented a temporary change to the facility. The inspectors reviewed the issue and identified that the facility had been operating outside of its design basis, which was reportable to the NRC. The licensee subsequently made a one hour non-emergency report to the NRC. Since there was no actual loss of safety function to the system, this issue was screened as very low risk significance.
Inspection Report# : 2000008(pdf)
Significance:        Jun 22, 2000 Identified By: NRC Item Type: FIN Finding MAINTENANCE RULE FAILURES ASSOCIATED WITH REACTOR HEAD VENT VALVE.
The inspectors reviewed the licensee's implementation of the maintenance rule for failures associated with a reactor head vent valve. The licensee's corrective action documents identified a potential maintenance rule functional failure but the completed evaluation of the problem did not document the final determination. However, the inspectors identified that the repeated failures may have been prevented if maintenance activities such as valve disassembly and cleaning had been performed. In this case, maintenance rule reliability goals were not exceeded. The licensee had documented similar maintenance rule program deficiencies and developed a corrective action program to address the deficiencies. Although programmatic deficiencies exist, since no maintenance rule reliability criteria had been exceeded, this issue was considered of very low risk significance.
Inspection Report# : 2000008(pdf)
Significance:        Jun 22, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO INITIATE KEWAUNEE ASSESSMENT PROCESS DOCUMENT REGARDING REFUELING WATER STORAGE TANK LOW-LOW LEVEL ALARM INOPERABILITY.
The inspectors identified that the refueling water storage tank low-low level alarm which was actuating five percent higher than normal had not been documented in a Kewaunee Assessment Process form by the licensee, and therefore had not received an operability evaluation. This failure was identified as contrary to site administrative procedures. Following the licensee's documentation of the problem, the inspectors identified that the associated operability evaluation considered the acceptability of an operator workaround to address the issue, but did not address any safety implications or consequences of the alarm actuating early. A subsequent operability evaluation by the licensee was evaluated as adequate by the inspectors. Since the subsequent operability evaluation was adequate and it was determined that no safety mitigation equipment was adversely affected by the early actuation of the alarm, this issue was considered of very low risk significance. A non-cited violation (NCV) was identified for failing to document a non-conforming condition, contrary to site administrative procedure requirements.
Inspection Report# : 2000008(pdf)
Significance:        Jun 22, 2000 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO TEST ADDITIONAL RELIEF VALVES IN ACCORDANCE WITH TECHINCAL SPECIFICATIONS.
The licensee identified that the suction relief valve for an auxiliary feedwater pump may have failed its relief test criteria, but did not process the documented deficiency until several weeks later. The licensee then expanded the scope of the relief testing to the suction relief valves associated with the other auxiliary feedwater pumps to meet technical specification requirements. Since any one train of auxiliary feedwater was capable of supplying 100 percent of the decay heat removal requirements, this issue was screened as very low risk significance. However, the time delay in complying with technical specification requirements for testing other relief valves was identified as an NCV.
Inspection Report# : 2000008(pdf)
Significance:        May 22, 2000 Identified By: NRC Item Type: FIN Finding FIRE EXTINGUISHERS NOT LOCATED IN ALL AREAS OF CONTAINMENT BUILDING AT BEGINNING OF REFUELING OUTAGE.
During a walkdown of the containment building, the inspectors identified that portable fire extinguishers were not located in the containment basement at the beginning of the plant refueling outage. Additionally, site fire protection procedures required that responsible fire protection personnel perform inspections of selected plant areas to ensure that the quantity of combustible material was minimized. However, the procedure did not list the containment as an area to be inspected and the procedures did not require the placement and location of portable fire suppression equipment inside containment during the refueling outage. Due to a low number of work activities ongoing at the time, this issue was screened as Green (very low risk significance).
Inspection Report# : 2000007(pdf)
 
4Q/2001 Inspection Findings - Kewaunee                                                                                                  Page 6 of 10 Significance:        May 22, 2000 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO INSTALL RAYCHEM HEAT SHRINK MATERIAL IN ACCORDANCE WITH PROCEDURE REQUIREMENTS.
The licensee identified that two Raychem electrical cable splices utilized in environmentally qualified (EQ) safety-related equipment had not been installed in accordance with EQ requirements. These splices were associated with pressurizer level transmitters and were installed in 1984. The licensee subsequently performed extensive EQ testing of the splices to determine the qualification of the splices' as-found configurations. Test results indicated that the splices would have been able to perform their intended function in a harsh environment inside containment. This issue was considered to be of low safety significance based on the successful EQ testing of the as-found splices' configurations and was screened as Green (very low risk significance). One non-cited violation was identified.
Inspection Report# : 2000007(pdf)
Barrier Integrity Significance:        Jun 22, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation FAILURE TO UPDATE COMPUTER ALARM FOR CURRENT AXIAL FLUX DISTRIBUTION TARGET BAND.
The licensee identified, following plant startup, that a computer alarm had not been updated properly to alarm if axial flux distribution deviated outside of the flux distribution target band. This condition was contrary to technical specification requirements. The licensee reviewed the axial flux distribution history since the startup and determined that at no time was the flux distribution outside of the target band. Since the axial flux distribution was never outside of the target band, this issue was screened as very low risk significance. An NCV was identified for failing to comply with technical specification requirements for monitoring axial flux distributions.
Inspection Report# : 2000008(pdf)
Significance:        May 22, 2000 Identified By: Licensee Item Type: FIN Finding TECHNICAL SPECIFICATIONS INTERPRETATION FOR TESTING REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES.
The NRC determined that the licensee's practice of testing reactor coolant system pressure isolation Valve SI-22B prior to entering the cold shutdown condition was contrary to Technical Specification requirements 4.2.a.3.a. Technical Specification 4.2.a.3.a required that periodic leakage testing of Valve SI-22B be accomplished prior to reaching operating mode after the plant was placed in cold shutdown. This issue was considered to be of low safety significance because of a subsequent successful valve test and was screened as Green (very low risk significance).
Enforcement discretion was applied to this item in accordance with Section VII.B.6 of the Enforcement Policy.
Inspection Report# : 2000007(pdf)
Significance:        May 19, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PROPERLY IMPLEMENT FLAW ACCEPTANCE CRITERIA FOR SLEEVE WELD INSPECTIONS.
During 1998 inservice inspection examinations, the licensee failed to properly implement the flaw acceptance criteria for laser welded sleeve inspection within two steam generator tubes. The safety significance was very low based on the absence of adverse consequences, and May 2000 in-situ pressure testing where both welds exhibited zero leakage at normal operating pressure, main steam line break pressure, and three times normal operating differential pressures. As such, this issue was characterized as a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion IX, Control of Special Processes. To correct the error, the licensee plugged both tubes (Section 1RO8).
Inspection Report# : 2000011(pdf)
Emergency Preparedness Significance:        Mar 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation SUPPLEMENTAL INSPECTION OF WHITE ERO AUGMENTATION FINDING AND RESULTING GREEN FINDING.
This supplemental inspection was performed by the NRC to evaluate the licensees evaluation associated with the failure to conduct successful quarterly, off-hours, unannounced staff augmentation drills during the second, third, and fourth quarters of 1999 and the second quarter of 2000.
This performance issue was previously characterized as having low to moderate risk significance (White) in NRC Inspection Report No. 50-
 
4Q/2001 Inspection Findings - Kewaunee                                                                                                    Page 7 of 10 305/2000015(DRS). During this supplemental inspection, performed in accordance with Inspection Procedure95002, the inspector concluded that the licensee performed a comprehensive evaluation of the unsuccessful staff augmentation drills. The licensees evaluation identifiedtwo root causes which resulted in the unsuccessful drills and in the staffs inability tocorrect thedeficiencies: (1) Management has not effectively acted to provide increased depth and flexibility in the emergency response organization following a reduction in staffing several years ago; and (2)
Management has accepted an adverse trend of test failures without requiring investigation into the root causes. The inspector reviewed the licensees corrective actions, both completed and planned, and concluded that the programmatic corrective actions appeared to address the identified root causes. In particular, the licensee assigned certain positions to an on-call rotation to ensure personnel were capable of augmenting in a timely manner, and the licensee was progressing in training additional staff to increase the depth of personnel assigned to key emergency response positions. In addition, the licensee was continuing its efforts in improving its corrective action program. The inspector reviewed the licensees immediate response to the issue and identified that one of the licensees initial corrective actions resulted in a Non-Cited Violation of regulatory requirements. To obtain a timely response of a key emergency response position (severe accident management - core hydraulics), the emergency preparedness staff effectively changed the emergency plan without revising the necessary procedures and without formally assessing the impact of that change. The staff instructed and trained personnel to respond to a location other than the Technical Support Center, which was contrary to the licensees current implementing procedures. While this change enabled the licensee to augment its staff in a timely manner, the change was not performed in accordance with NRC requirements. In order to make such a change, the licensees emergency plan required that the change be formally assessed to ensure that it did not reduce the effectiveness of the plan or any other implementing procedure. Since the issue did not result in a failure to meet an emergency preparedness planning standard, the failure to adequately implement the emergency plan was determined to be a violation of very low safety significance (Green) (Section 02.3(a)). Due to the licensees acceptable performance in assessing the emergency response augmentation drill deficiencies, the White finding associated with this issue will only be considered in assessing plant performance for a total of four quarters in accordance with the guidance in IMC 0305, Operating Reactor Assessment Program. Implementation of the licensees corrective actions will be reviewed during a future inspection.
Inspection Report# : 2001007(pdf)
Significance:          Mar 30, 2001 Identified By: NRC Item Type: VIO Violation FAILURE TO CORRECT SELF-IDENTIFIED ERO AUGMENTATION DRILL DEFICIENCIES.
During a baseline inspection of the emergency preparedness program conducted on August 14 - September 21, 2000, the NRC identified a preliminary White issue and potential violation for the licensees failure to successfully correct deficiencies identified during staff augmentation drills and to demonstrate timely staff augmentation in 1999 and 2000. The issue was unresolved pending the outcome of the NRCs final significance determination. On January 30, 2001, the NRC conducted a regulatory conference with the licensee and subsequently issued the licensee a White finding and Notice of Violation (Enforcement Action No. 00-214) associated with the performance issue.
Inspection Report# : 2001007(pdf)
Significance:          Mar 23, 2001 Identified By: NRC Item Type: NCV NonCited Violation POST-ACCIDENT SAMPLING SYSTEM CONTAINMENT AIR SAMPLE PANEL TESTING.
A Non-Cited Violation of Technical Specification 6.14 was identified for the failure to implement a program that ensured the capability to obtain and analyze containment atmosphere samples under accident samples using the cantainment air sampling panel (CASP). Although the CASP was installed, as was indicated in the emergency plan, the licensee had neither developed procedures nor had tested its capability to obtain a containment atmosphere sample using the CASP. The licensee could not recall if and when containment air samples were last obtained using the CASP.
Inspection Report# : 2001006(pdf)
Significance: N/A Feb 27, 2001 Identified By: NRC Item Type: FIN Finding SUPPLEMENTAL INSPECTION OF YELLOW ANS PERFORMANCE INDICATOR AND ASSOCIATED CORRECTIVE ACTION PROGRAM DEFICIENCIES.
This supplemental inspection was performed by the NRC to assess the licensee's evaluation associated with a Yellow performance indicator for the Alert and Notification System (ANS) and the associated Yellow finding related to the licensee's corrective action program. During this supplemental inspection, performed in accordance with Inspection Procedure 95002, the inspectors concluded that the licensee performed comprehensive evaluations of the performance problems associated with the ANS and its corrective action program. These evaluations identified primary root causes and contributing causes for both issues. Along with the electronics/hardware problems, the licensee identified the primary root causes for the ANS to be the failure to make changes to the system via a change control process and the failure of the activation procedure to provide for alternate/backup activation methods and to provide clear success criteria. In the case of the corrective action program, the licensee concluded that the primary root causes were plant management's inadequate risk evaluation regarding decisions affecting the corrective action program and the development of a culture in the licensee's organization, which minimized the importance of information from outside organizations. In particular, the licensee focused on low-cost power operation and failed to recognize the value of changes and improvements which had occurred throughout the industry, such as the value and expansion of the corrective action program. In the case of the Yellow ANS performance indicator, the licensee had completed several significant corrective actions to address the root causes and contributing causes identified in its evaluation. The inspectors found that the corrective actions appeared appropriate to address the underlying root causes and that ANS testing data indicated an improving trend in the NRC performance indicator. The licensee also performed comprehensive assessments of the emergency preparedness program, quality assurance program, plant operations, and other plant programs to determine the extent of condition (re. the root causes described above). Based on these evaluations, the licensee began to implement significant actions to correct the deficiencies in the corrective action program and other weaknesses identified. Generally, the inspectors observed progress in the licensee's initial implementation of these corrective actions. Due to the licensee's acceptable performance in assessing the Yellow ANS performance indicator and the associated Yellow finding, the Yellow finding will not be considered in assessing future plant performance.
Inspection Report# : 2001005(pdf)
 
4Q/2001 Inspection Findings - Kewaunee                                                                                                  Page 8 of 10 Significance:        Sep 21, 2000 Identified By: NRC Item Type: FIN Finding INADEQUATE ROOT CAUSE EVALUATION FOR YELLOW ALERT AND NOTIFICATION SYSTEM PERFORMANCE INDICATOR.
The licensee's evaluation of the Yellow Alert and Notification (siren) System Performance Indicator (PI) was inadequate. The inspector concluded that the licensee's evaluation was not performed at the depth necessary to identify the root causes of the siren performance problems and, instead, only identified the symptoms of the root causes. Specifically, the inspector identified the following substantive weaknesses in the licensee's evaluation of the siren system performance, which appeared to result from systemic corrective action program deficiencies within this cornerstone:
* The licensee's evaluation was not of sufficient depth to clearly identify the root causes associated with the decline in siren system performance.
* Licensee management did not provide well-understood and clear guidance/expectations for performing root cause evaluations.
* The licensee's evaluation of the quality assurance program was narrowly focused and was not critical of its role in failing to identify and correct the siren performance problems.
* The licensee did not establish a priority for each of the long-term corrective actions in accordance with the associated significance or risk.
* The licensee did not have any formal provisions for measuring the effectiveness of its corrective actions.
* Within the licensee's evaluation, the licensee had not evaluated common causes or the extent of the condition. Due to the corrective action program performance deficiencies within this cornerstone, we have been unable to conclude that the performance issues that resulted in the yellow PI have been addressed. Therefore, we are issuing a yellow finding that corresponds to the original issues that resulted in a yellow PI. Additional inspection effort will be focused on the licensee's further evaluation of the siren reliability root causes and the continuing corrective action program implementation deficiencies identified during this inspection.
Inspection Report# : 2000017(pdf)
Significance: N/A Apr 05, 2000 Identified By: NRC Item Type: FIN Finding LICENSEE FAILED TO IDENTIFY THE FULL SCOPE OF PROBLEMS WITH THE ALERT AND NOTIFICATION SYSTEM PERFORMANCE.
The inspectors concluded that the licensee's assessment was not sufficiently comprehensive to identify the full scope of problems associated with the Alert and Notification System (ANS) performance program. As a result, licensee corrective actions generally were focused on the equipment problem rather than all root causes and contributing factors. For example, the inspectors identified that: (1) management oversight of the ANS performance program was limited; (2) an audit failed to identify degrading ANS performance as a concern; (3) annual preventive maintenance was not consistently performed on the system; (4) the corrective action program was not used consistently to document ANS problems; and (5) maintenance procedures and records were deficient. Collectively, these problems indicate that the ANS performance program lacked sufficient structure and oversight.
Inspection Report# : 2000006(pdf)
Occupational Radiation Safety Significance:        Oct 02, 2001 Identified By: NRC Item Type: NCV NonCited Violation HIGH RADIATION AREA ACCESS CONTROLS Non-Cited Violation of Technical Specification 6.13 and an associated Green Finding for failure to 'barricade' three ladders that provided entry to high radiation areas (less than 1000 mrem/hour) located on the steam generator/pressurizer platforms.
Inspection Report# : 2001014(pdf)
Significance:        May 19, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO POST A VERY HIGH RADIATION AREA.
The inspectors identified a noncited violation for the failure to post a very high radiation area in accordance with 10 CFR 20.1902(c). Although the area was not adequately posted, the licensee had provided physical controls and barriers that were consistent with its requirements for a very high radiation area. Based on the adequacy of these controls, the potential for an overexposure from the inadvertent entry of personnel into the area was low. Consequently, this finding was determined to be of very low safety significance (Section 20S1.1).
Inspection Report# : 2000009(pdf)
Public Radiation Safety Physical Protection
 
4Q/2001 Inspection Findings - Kewaunee                                                                                                  Page 9 of 10 Significance: N/A Jul 21, 2000 Identified By: NRC Item Type: FIN Finding CORRECTIVE ACTION ON SEARCH ISSUES DID NOT WORK.
The inspector determined that the licensee's effectiveness of implemented corrective actions for a previously identified inspection finding regarding an inadequate vehicle search was not totally effective in preventing recurrence. Previous corrective action was not adequately focused (Section 3PP2.2).
Inspection Report# : 2000013(pdf)
Significance:        Jul 21, 2000 Identified By: NRC Item Type: NCV NonCited Violation SEARCHES INADEQUATE (REPEAT).
The inspector identified a Non-Cited violation by observing that, a security officer failed to search an easily accessible compartment on one vehicle.
The failure resulted from human error because the officer did not observe the access panel to the compartment during the vehicle search process.
When searched, no prohibited items were found. Corrective actions were implemented. The inspector determined that the licensee's effectiveness of implemented corrective actions for a previously identified inspection finding regarding an inadequate vehicle search was not totally effective in preventing recurrence. Previous corrective action was not adequately focused.
Inspection Report# : 2000013(pdf)
Miscellaneous Significance: N/A Aug 24, 2001 Identified By: NRC Item Type: FIN Finding IDENTIFICATION AND RESOLUTION OF PROBLEMS The team concluded that the licensee was generally effective at identifying problems and putting them into the corrective action program. The program itself contained all the necessary attributes of an acceptable corrective action program and was generally successful in correcting identified issues. However, the team noted that, although licensee management had taken efforts to ensure that issues were resolved in accordance with program guidance and requirements, additional efforts appeared necessary to ensure timely resolution of issues. A positive program initiative was the establishment of positions in each of the major plant departments to serve as liaisons between the departments and the corrective action program and to assist with self-assessments. However, examples were identified by the inspectors of problems with the licensee's identification and resolution of problems, prioritization and evaluation of issues, and the effectiveness of corrective actions. Included in these examples were the routine granting of due date extensions for problem evaluation and corrective action implementation, failure to perform a Technical Specification-required biennial surveillance of safety-related procedures, and ineffective corrective actions that resulted in the lack of proper controls over the storage of grease used in safety-related breakers. Based on a review of records and discussions with plant staff, the inspectors concluded that workers at the site felt free to input safety issues into the corrective action program.
Inspection Report# : 2001012(pdf)
Significance: N/A Nov 09, 2000 Identified By: NRC Item Type: FIN Finding FAILURE TO TAKE ADEQUATE CORRECTIVE ACTIONS TO ENSURE COMPONENT RE-TEST REQUIREMENTS COMPLETED ACCORDING TO MAINTENANCE PROCEDURES.
No Color. The inspectors determined that a negative performance trend had developed in the licensee's ability to identify and promptly take appropriate corrective actions to prevent recurrence based on two previously identified examples (NCV 50-305/2000014-01) and one example identified during this inspection period (NCV 50-305/00-20-01). All three examples related to the licensee returning safety-related equipment to service prior to completing all required post-maintenance retesting. While the risk of the individual examples was very low, the licensee had failed to ensure that all retest requirements had been completed before returning safety-related equipment to service. These findings collectively indicated a problem with the licensee's ability to provide timely and adequate corrective actions to prevent recurrence. (Section 4OA2).
Inspection Report# : 2000020(pdf)
Significance: N/A Sep 29, 2000 Identified By: NRC Item Type: FIN Finding EFFECTIVENESS OF CORRECTIVE ACTION PROGRAM.
Based on the results of this inspection, the NRC concluded that the corrective action program at Kewaunee showed significant weaknesses and inconsistencies across all of the procedural elements inspected. These weaknesses existed across departments and affected multiple cornerstones in the strategic performance areas of Reactor Safety, Radiation Safety, and Safeguards. Of particular note was the lack of procedures for determining the significance of conditions adverse to quality and for trending of issues and the complete lack of trending within your corrective action program. We also identified a lack of urgency in correcting issues which resulted in repeat examples occurring and, coupled with a poor tracking system, a tendency for issues to be dropped. While none of the specific examples identified by the team were of high risk significance when looked at in isolation, in the aggregate they were similar in nature to prior issues in the emergency preparedness area that rose to a higher significance level and contributed to a degraded cornerstone. While we concluded that the station had fostered an environment in which personnel freely identified conditions adverse to quality without fear of discrimination or retaliation, we also concluded that significant weaknesses with, and inconsistent implementation of, the station corrective action program resulted in multiple examples where station personnel did not enter deficiencies into the station's formal corrective action program.
Inspection Report# : 2000019(pdf)
 
4Q/2001 Inspection Findings - Kewaunee                                                                                                Page 10 of 10 Significance: N/A Sep 29, 2000 Identified By: NRC Item Type: NCV NonCited Violation NO PROCEDURAL GUIDANCE FOR DETERMINING IF CONDITIONS ADVERSE TO QUALITY ARE SIGNIFICANT.
The inspectors reviewed the quality assurance (QA) manual requirements against Kewaunee's implementing procedures and identified that two QA manual requirements were not being implemented. Specifically, Program Requirement 3.1.9 which stated that directives and procedures shall provide for the review of conditions adverse to quality to determine if the conditions are significant in nature. This requirement paralleled 10 CFR Part 50, Appendix B, Criterion XVI, which requires that the cause of significant conditions adverse to quality be determined and corrective actions taken to prevent recurrence. The inspectors reviewed the Nuclear Administrative Directive (NAD 11.08) and the procedure (GNP 11.08.01) governing the KAP (Kewaunee corrective action program) process and found no procedure requirements for identifying significant conditions adverse to quality. This finding does not directly affect a cornerstone. As a result, this issue was not evaluated with the Significance Determination Process and was not assigned a color. One example of a non-cited violation of 10 CFR Part 50, Appendix B, Criterion II, "Quality Assurance Programs," was identified.
Inspection Report# : 2000019(pdf)
Significance: N/A Sep 29, 2000 Identified By: NRC Item Type: NCV NonCited Violation NO PROCEDURAL GUIDANCE FOR TRENDING CONDITIONS ADVERSE TO QUALITY.
The inspectors reviewed the quality assurance (QA) manual requirements against Kewaunee's implementing procedures and identified that two QA manual requirements were not being implemented. Specifically, QA Program Requirement, 3.1.10, stated that directives and procedures shall provide for analyzing trends of conditions adverse to quality. Once identified these trends were required to be considered significant conditions adverse to quality. The inspectors found that conditions adverse to quality were not defined in the KAP (Kewaunee corrective action program) procedures and that no procedure existed for trending. This finding does not directly affect a cornerstone. As a result, this issue was not evaluated with the Significance Determination Process and was not assigned a color. One example of a non-cited violation of 10 CFR Part 50, Appendix B, Criterion II, "Quality Assurance Program," was identified.
Inspection Report# : 2000019(pdf)
Significance: N/A Jun 22, 2000 Identified By: NRC Item Type: FIN Finding CONTROL ROOM OPERATIONS HUMAN PERFORMANCE ISSUES.
The inspectors interviewed operators to evaluate their awareness of degraded control room indications and alarms, and their ability to adequately take manual actions based on degraded alarm functions. The inspectors identified, during interviews, that there was a lack of awareness by operators of a degraded refueling water storage tank low-low level alarm which would be potentially confusing to operators and therefore increase the risk associated with initiating long term sump recirculation.
Inspection Report# : 2000008(pdf)
Last modified : March 01, 2002
 
1Q/2002 Inspection Findings - Kewaunee                                                                                            Page 1 of 12 Kewaunee Initiating Events Mitigating Systems Significance:        Feb 21, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform Thorough 10 CFR 50.59 Safety Evaluation The inspectors identified a Non-Cited Violation for failure to perform an adequate 10 CFR 50.59 safety evaluation associated with emergency operating procedure changes to address component cooling water pump dead-head operational concerns. The safety evaluation did not evaluate the potential for initiating a loss-of-coolant accident via the reactor coolant loop seals during conditions of a complete loss of component cooling water.
Inspection Report# : 2001017(pdf)
Significance: N/A Sep 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ESTABLISH CONTINGENCY PLANS FOR ORANGE RISK CONDITION The inspectors identified the failure to establish contingency plans during a planned high risk plant configuration. Contrary to administrative requirements, the licensee approved an orange risk condition during a refueling outage with no contingency plans to mitigate the consequences of a loss of spent fuel pool cooling with a full core offload in the pool. A Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was identified. The finding was of very low safety significance because although the licensee had not approved appropriate contingency actions for the orange risk condition, the licensee subsequently rescheduled the planned maintenance to eliminate the orange risk condition.
Inspection Report# : 2001013(pdf)
Significance: N/A Aug 24, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO CONDUCT BIENNIAL SURVIELLANCE OF SAFETY-RELATED PROCEDURES PER TS 6.8.c A Non-Cited Violation of Technical Specification 6.8.c was identified for the failure to perform a biennial surveillance of safety-related procedues. This issue was more than minor because if left uncorrected, could under the same condition become a more significant safety concern. However, since no specific cornerstone had been impacted, this finding is designated as No Color.
Inspection Report# : 2001012(pdf)
Significance:        Aug 24, 2001 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE CORRECTIVE ACTIONS FOR PREVIOUS PROBLEM WITH THE CONTROL OF SAFETY-RELATED MATERIALS A Non-Cited Violation of Criterion XVI, "Corrective Action," of Appendix B of 10 CFR Part 50 was identified for ineffective corrective actions for a problem with the control of the storage of consumable materials, such as thread sealant, used in safety-related applications. These ineffective actions subsequently resulted in the inadequate control of the storage of grease used in safety-related breakers. This issue was more than minor because if left uncorrected, could under the same condition become a more significant safety concern. In that this issue could credibly affect the operability, availability, reliability, or function of a system or train in a mitigating system, it is a Green finding.
Inspection Report# : 2001012(pdf)
Significance:        Aug 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation NON-RATED FIRE BARRIER
 
1Q/2002 Inspection Findings - Kewaunee                                                                                      Page 2 of 12 A Non-Cited Violation [of 10 CFR Part 50, Appendix R, Section III.G.2.a] was identified for failure to provide a 3-hour rated fire barrier to separate redundant trains of safe shutdown equipment. This finding was of very low safety significance because the licensee tested a replica of the fire barrier and demonstrated that the fire barrier provided protection for at least 60 minutes, which was sufficient for the hazards in the area.
Inspection Report# : 2001011(pdf)
Significance:        Jul 20, 2001 Identified By: NRC Item Type: FIN Finding LICENSED OPERATOR REQUALIFICATION EXAMINATION RESULTS The inspectors identified that two of eight crews examined during the licensee's calendar year 2001 licensed operator requalification operating test had failed. The finding was of very low safety significance because both crews that had failed received remedial training prior to being returned to shift, and the results of the licensee's operator licensing requalification operating test given in calendar year 2000 indicated that only one crew, out of a total of eight crews tested, had failed.
Inspection Report# : 2001011(pdf)
Significance:        Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IDENTIFY DEFICIENT CONDITION OF VALVE AFW-1B The inspectors identified that the licensee failed to promptly identify and correct the B' train auxiliary feedwater pump discharge check valve which was stuck in an intermediate position. A Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was issued. The finding was of very low safety significance because, although the check valve was stuck in an intermediate position, the time that it was known to have been stuck was less than the technical specification allowed outage time for one train of auxiliary feedwater to be out of service (less than 72 hours). Additionally, the other two trains of auxiliary feedwater were each capable of 100 percent decay heat removal.
Inspection Report# : 2001009(pdf)
Significance: N/A Jun 30, 2001 Identified By: NRC Item Type: FIN Finding FAILURE TO BALANCE RELIABILITY AND AVAILABILITY AS REQUIRED BY 10 CFR 50.65(a)(3)
The inspectors identified a failure to evaluate whether adjustments were necessary such that there would be an appropriate balance between systems' availability and reliability in accordance with 10 CFR 50.65(a)(3) of the maintenance rule. The inspectors identified that the licensee did not have an administrative process to track maintenance rule functional failures and maintenance preventible maintenance functional failures. As a result, reliability and availability could not be balanced as required by the Maintenance Rule periodic evaluation. The safety significance of the specific finding was very low because it did not affect the operability of the systems, and the licensee entered the finding in the corrective action program. However, this finding was considered to be of regulatory concern in the area of maintenance rule implementation due to the extent of the problems with the Maintenance Rule Program.
Inspection Report# : 2001009(pdf)
Significance:        Jun 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO TRACK UNAVAILABILITY OF SYSTEMS REQUIRED DURING SHUTDOWN OPERATION 10 CFR 50.65(a)(1), required, in part, that the licensee monitor the performance or condition of SSCs within the scope of the rule as defined by 10 CFR 50.65(b), against licensee-established goals, in a manner sufficient to provide reasonable assurance that such SSCs are capable of fulfilling their intended functions. 10 CFR 50.65(a)(2) stated, in part, that monitoring as specified in 10 CFR 50.65(a)(1) was not required where it had been demonstrated that the performance or condition of an SSC was being effectively controlled through the performance of appropriate preventive maintenance, such that the SSC remained capable of performing its intended function. Contrary to the above, the licensee failed to demonstrate that the performance or condition of systems required to be available during shutdown conditions and within the scope of the rule had been effectively controlled through the performance of appropriate preventive maintenance and did not monitor against licensee-established goals. Specifically, the licensee failed to monitor the unavailability of systems required during shutdown operation.
Inspection Report# : 2001009(pdf)
Significance:        Jun 30, 2001 Identified By: Licensee
 
1Q/2002 Inspection Findings - Kewaunee                                                                                        Page 3 of 12 Item Type: NCV NonCited Violation FAILURE TO ESTABLISH MAINTENANCE RULE (a)(1) GOALS 10 CFR 50.65(a)(1), requires, in part, that the licensee monitor the performance or condition of SSCs within the scope of the rule as defined by 10 CFR 50.65(b), against licensee-established goals, in a manner sufficient to provide reasonable assurance that such structures, systems, and components, are capable of fulfilling their intended functions. Such goals shall be established commensurate with safety. When the performance or condition of a structure, system, or component does not meet established goals, appropriate corrective action shall be taken. Contrary to the above, from 1996, the licensee did not take appropriate corrective actions when the performance of those systems in (a)(1) did not meet licensee established goals. Specifically, the licensee determined timely and appropriate corrective actions had not been taken for five systems that had been in (a)(1) category for approximately 3 years to 5 years: component cooling (entered (a)(1) on April 23, 1997), control room air conditioning (July 24, 1996),
station and instrument air (July 3, 1997), auxiliary building air ventilation (July 31, 1997), and control rod drive (August 6, 1998). This issue is in the licensee's corrective action system as KAP WO 01-3323. The inspectors evaluated the risk significance of this issue using the Significance Determination Process. The inspectors did not identify where this failure resulted in a total loss of a risk significant SSC. Therefore, this issue was screened as Green (very low risk significance) after a Phase 1 Significance Determination Process review. Although the risk significance of this issue was low, the inspectors concluded that this was more than a minor concern because the failure to recognize and correct ineffective maintenance practices resulted in risk significant systems in (a)(1) for years with no improvement in performance. The NRC tracking number for this issue is 50-305/01-09-02.
Inspection Report# : 2001009(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation FIRE DOOR FUSIBLE LINKS.
On February 20, 2001, the licensee determined that the installed fusible link arrangement on roll-up fire Doors 279 and 281, which separated both trains of service water pumps, would not actuate as designed to ensure that the doors would automatically close to provide a 3-hour fire barrier, contrary to 10 CFR Part 50, Appendix R, Section III.G.2.a which required, in part, separation of cables and equipment of redundant trains by a fire barrier having a 3-hour rating.
Inspection Report# : 2001006(pdf)
Significance:        Mar 12, 2001 Identified By: Licensee Item Type: NCV NonCited Violation INADEQUATE SMOKE DETECTOR COVERAGE IN FIRE ZONE TU-95B.
Licensee identified violation of licensee's operating license that the licensee failed to install a detector in each beam pocket in Fire Zone TU-95B.
Inspection Report# : 2001002(pdf)
Significance:        Feb 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO TEST FIRE DOOR IN ACCORDANCE WITH FIRE PLAN.
The inspectors identified a non-cited violation for failure to properly test a fire door in accordance with the facility's fire protection program plan. The finding was of very low safety significance because, although the fire door separated both trains of service water pumps and did not fully close as designed when subsequently tested, the fire loading in the area was insufficient to credibly impact more than two of the four service water pumps in the area.
Inspection Report# : 2001004(pdf)
Significance: N/A Nov 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO COMPLETE COMPONENT RE-TEST REQUIREMENTS PRIOR TO RETURNING EQUIPMENT TO AN OPERABLE STATUS.
No Color. The inspectors identified a Non-Cited Violation for failure to complete component retest requirements following maintenance performed on the B train control room air conditioner compressor condenser. The unit had been returned to an operable status prior to the retest requirements being completed as prescribed in the associated maintenance procedure. This issue was determined to be a violation of the licensee's Operational Quality Assurance Program Manual, Section 8, "Maintenance Planning and Control." Although the risk associated with this finding was very low and did not affect any cornerstones, the inspectors noted that this finding was similar to previous NRC-identified findings and therefore was of greater than minor significance and warranted documentation. (Section 1R19).
Inspection Report# : 2000020(pdf)
 
1Q/2002 Inspection Findings - Kewaunee                                                                                    Page 4 of 12 Significance:        Sep 29, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IDENTIFY CORROSION AS POTENTIAL FAILURE MECHANISM.
The inspectors identified that the licensee failed to identify corrosion as a potential failure mechanism in the operability determination for a carbon steel key in the service water system. Thus, the licensee failed to quantify the corrosion rate and therefore did not adequately evaluate the expected service life of the carbon steel key. One non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was identified.
Inspection Report# : 2000019(pdf)
Significance:        Sep 29, 2000 Identified By: NRC Item Type: NCV NonCited Violation PRESSURE RATING OF AUXILIARY FEEDWATER STEAM TRAPS.
The inspectors identified that a root cause evaluation for a 1996 equipment issue in the turbine-driven auxiliary feedwater system was not completed until 1999. The evaluation stated that the internals of the steam traps were designed to operate at pressures up to a maximum 600 psig but that the traps were exposed to pressures up to 1025 psig. A corrective action item to initiate a design change request to replace the steam traps with a different model rated for the design pressure of the system was described in the evaluation. However, the inspectors identified that the design change request had never been initiated and the KAP ( Kewaunee corrective action document) had been closed. As a result, the corrective action item for this design problem was lost. In addition, operability of the system had never been formally evaluated despite the identification that the system design requirements were not met. The licensee subsequently determined that the steam traps remained operable and was planning to initiate the design change to correct the problem. One non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was identified.
Inspection Report# : 2000019(pdf)
Significance:        Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation DESIGN CONTROL VIOLATION FOR AUXILIARY FEEDWATER STRAINER MESH SIZE.
The inspectors questioned the mesh size of the strainers, which were installed in the suction of the three auxiliary feedwater (AFW) pumps. As a result of the inspectors' questions, license personnel inspected the strainers on August 21, 2000, and found the strainers to have 1/16 inch openings. A note was later found on Figure 10.2-3 of the UFSAR that indicated that the AFW suction strainer size was 1/8 inch. The smaller openings would not support the use of service water as a safety related source for AFW and as a result all three trains of AFW were declared inoperable. This condition had apparently existed for approximately 25 years and was identified as a non-cited violation of Criterion III, "Design Control," of 10 CFR 50, Appendix B.
Inspection Report# : 2000012(pdf)
Significance: N/A Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation RETRIEVAL OF SERVICE WATER SYSTEM DESIGN INFORMATION.
In many cases, design basis information for the service water system was difficult if not impossible to locate. Licensee personnel wrote KAP WO 00-002566 to enter the problem in the corrective action program.
Inspection Report# : 2000012(pdf)
Significance:        Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PROPERLY CONTROL CORRECTED TEST DATA NECESSARY FOR DESIGN CALCULATIONS.
An example of a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," was identified in the handling of service water system flow test data, which was subsequently used in calculations. Gauge readings corrected for post test calibration checks, gauge reading corrections for elevation considerations, and flow values corrected for pump degradation were contained in spreadsheets in the possession of an individual staff member, but not currently packaged with raw test data, and not bearing evidence of a formal review and control process. The connection between the test data, which had been vaulted, and the values used in the calculation, could not be made without use of the uncontrolled spreadsheet.
Inspection Report# : 2000012(pdf)
 
1Q/2002 Inspection Findings - Kewaunee                                                                                  Page 5 of 12 Significance:        Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation IMPROPER DESIGN CALCULATION IDENTIFICATION, NON-CONSERVATIVE ASSUMPTIONS, CALCULATION ERRORS, AND DUPLICATE CALCULATIONS.
An example of a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," was identified because of inadequate control of design calculations. The control failures included improper identification of calculations, non-conservative assumptions, calculation errors, and duplicate or superceded calculations not properly identified or canceled. The failure to follow the established design control process increased the potential for errors in the design and operation of the service water system.
Because the system was subsequently demonstrated to be capable of removing the design heat load, the actual significance was low and this finding screened out as having very low risk significance.
Inspection Report# : 2000012(pdf)
Significance:        Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO DOCUMENT IN THE CORRECTIVE ACTION PROGRAM THE USE OF INCORRECT MATERIAL IN A SERVICE WATER PUMP KEY.
An example of a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was identified because of inadequate corrective action to correct an incorrect coupling adjust nut set screw and a low strength "soft" key material, which had contributed to a pump shaft failure. Licensee personnel had known of the "soft" key material since July 21, 1999. The "soft" key material was found in other service water pumps but had not been removed from all pumps. As of July 25, 2000, licensee personnel had not documented the existence of the "soft" key material in the corrective action program.
Inspection Report# : 2000012(pdf)
Significance:        Jul 07, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO COMPLETE COMPONENT RETEST REQUIREMENTS IN ACCORDANCE WITH PROCEDURE.
On June 26, 2000, during a review of post maintenance testing requirements following maintenance performed on the control room post accident system charcoal filter heat detector, the inspectors identified that maintenance technicians had not completed the component re-test requirements, as required by a preventative maintenance procedure prior to the system being returned to an operable status. On July 7, the inspectors identified a second example of failing to complete component re-test requirements following maintenance on the zone special ventilation system charcoal filter heat detector as required. The issue was considered to be of very low safety significance based on the determination that although the licensee had not completed all of the component retest requirements prior to returning the equipment to service, the subsequent testing determined that the equipment was in an operable status. The failure to complete the component retest requirements in accordance with site procedures was identified as a Non-Cited Violation.
Inspection Report# : 2000014(pdf)
Significance:        Jun 22, 2000 Identified By: NRC Item Type: FIN Finding FAILURE TO MEET SINGLE FAILURE CRITERIA FOR RESIDUAL HEAT REMOVAL VALVE CIRCUITRY.
The licensee identified that the circuitry associated with the residual heat removal system discharge to safety injection system suction isolation valves did not meet single failure criteria. The inspectors noted that this design requirement was identified in the facility's updated safety analysis report. The licensee subsequently implemented a temporary change to the facility. The inspectors reviewed the issue and identified that the facility had been operating outside of its design basis, which was reportable to the NRC.
The licensee subsequently made a one hour non-emergency report to the NRC. Since there was no actual loss of safety function to the system, this issue was screened as very low risk significance.
Inspection Report# : 2000008(pdf)
Significance:        Jun 22, 2000 Identified By: NRC Item Type: FIN Finding MAINTENANCE RULE FAILURES ASSOCIATED WITH REACTOR HEAD VENT VALVE.
The inspectors reviewed the licensee's implementation of the maintenance rule for failures associated with a reactor head vent
 
1Q/2002 Inspection Findings - Kewaunee                                                                                      Page 6 of 12 valve. The licensee's corrective action documents identified a potential maintenance rule functional failure but the completed evaluation of the problem did not document the final determination. However, the inspectors identified that the repeated failures may have been prevented if maintenance activities such as valve disassembly and cleaning had been performed. In this case, maintenance rule reliability goals were not exceeded. The licensee had documented similar maintenance rule program deficiencies and developed a corrective action program to address the deficiencies. Although programmatic deficiencies exist, since no maintenance rule reliability criteria had been exceeded, this issue was considered of very low risk significance.
Inspection Report# : 2000008(pdf)
Significance:        Jun 22, 2000 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO TEST ADDITIONAL RELIEF VALVES IN ACCORDANCE WITH TECHINCAL SPECIFICATIONS.
The licensee identified that the suction relief valve for an auxiliary feedwater pump may have failed its relief test criteria, but did not process the documented deficiency until several weeks later. The licensee then expanded the scope of the relief testing to the suction relief valves associated with the other auxiliary feedwater pumps to meet technical specification requirements. Since any one train of auxiliary feedwater was capable of supplying 100 percent of the decay heat removal requirements, this issue was screened as very low risk significance. However, the time delay in complying with technical specification requirements for testing other relief valves was identified as an NCV.
Inspection Report# : 2000008(pdf)
Significance:        Jun 22, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO INITIATE KEWAUNEE ASSESSMENT PROCESS DOCUMENT REGARDING REFUELING WATER STORAGE TANK LOW-LOW LEVEL ALARM INOPERABILITY.
The inspectors identified that the refueling water storage tank low-low level alarm which was actuating five percent higher than normal had not been documented in a Kewaunee Assessment Process form by the licensee, and therefore had not received an operability evaluation. This failure was identified as contrary to site administrative procedures. Following the licensee's documentation of the problem, the inspectors identified that the associated operability evaluation considered the acceptability of an operator workaround to address the issue, but did not address any safety implications or consequences of the alarm actuating early.
A subsequent operability evaluation by the licensee was evaluated as adequate by the inspectors. Since the subsequent operability evaluation was adequate and it was determined that no safety mitigation equipment was adversely affected by the early actuation of the alarm, this issue was considered of very low risk significance. A non-cited violation (NCV) was identified for failing to document a non-conforming condition, contrary to site administrative procedure requirements.
Inspection Report# : 2000008(pdf)
Significance:        May 22, 2000 Identified By: NRC Item Type: FIN Finding FIRE EXTINGUISHERS NOT LOCATED IN ALL AREAS OF CONTAINMENT BUILDING AT BEGINNING OF REFUELING OUTAGE.
During a walkdown of the containment building, the inspectors identified that portable fire extinguishers were not located in the containment basement at the beginning of the plant refueling outage. Additionally, site fire protection procedures required that responsible fire protection personnel perform inspections of selected plant areas to ensure that the quantity of combustible material was minimized. However, the procedure did not list the containment as an area to be inspected and the procedures did not require the placement and location of portable fire suppression equipment inside containment during the refueling outage. Due to a low number of work activities ongoing at the time, this issue was screened as Green (very low risk significance).
Inspection Report# : 2000007(pdf)
Significance:        May 22, 2000 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO INSTALL RAYCHEM HEAT SHRINK MATERIAL IN ACCORDANCE WITH PROCEDURE REQUIREMENTS.
The licensee identified that two Raychem electrical cable splices utilized in environmentally qualified (EQ) safety-related equipment had not been installed in accordance with EQ requirements. These splices were associated with pressurizer level transmitters and were installed in 1984. The licensee subsequently performed extensive EQ testing of the splices to determine the qualification of the splices' as-found configurations. Test results indicated that the splices would have been able to perform their intended function in a harsh environment inside containment. This issue was considered to be of low safety significance based on the successful EQ testing of the as-found splices' configurations and was screened as Green (very low risk significance). One non-cited violation was
 
1Q/2002 Inspection Findings - Kewaunee                                                                                      Page 7 of 12 identified.
Inspection Report# : 2000007(pdf)
Barrier Integrity Significance:        Jun 22, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation FAILURE TO UPDATE COMPUTER ALARM FOR CURRENT AXIAL FLUX DISTRIBUTION TARGET BAND.
The licensee identified, following plant startup, that a computer alarm had not been updated properly to alarm if axial flux distribution deviated outside of the flux distribution target band. This condition was contrary to technical specification requirements. The licensee reviewed the axial flux distribution history since the startup and determined that at no time was the flux distribution outside of the target band. Since the axial flux distribution was never outside of the target band, this issue was screened as very low risk significance. An NCV was identified for failing to comply with technical specification requirements for monitoring axial flux distributions.
Inspection Report# : 2000008(pdf)
Significance:        May 22, 2000 Identified By: Licensee Item Type: FIN Finding TECHNICAL SPECIFICATIONS INTERPRETATION FOR TESTING REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES.
The NRC determined that the licensee's practice of testing reactor coolant system pressure isolation Valve SI-22B prior to entering the cold shutdown condition was contrary to Technical Specification requirements 4.2.a.3.a. Technical Specification 4.2.a.3.a required that periodic leakage testing of Valve SI-22B be accomplished prior to reaching operating mode after the plant was placed in cold shutdown. This issue was considered to be of low safety significance because of a subsequent successful valve test and was screened as Green (very low risk significance). Enforcement discretion was applied to this item in accordance with Section VII.B.6 of the Enforcement Policy.
Inspection Report# : 2000007(pdf)
Significance:        May 19, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PROPERLY IMPLEMENT FLAW ACCEPTANCE CRITERIA FOR SLEEVE WELD INSPECTIONS.
During 1998 inservice inspection examinations, the licensee failed to properly implement the flaw acceptance criteria for laser welded sleeve inspection within two steam generator tubes. The safety significance was very low based on the absence of adverse consequences, and May 2000 in-situ pressure testing where both welds exhibited zero leakage at normal operating pressure, main steam line break pressure, and three times normal operating differential pressures. As such, this issue was characterized as a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion IX, Control of Special Processes. To correct the error, the licensee plugged both tubes (Section 1RO8).
Inspection Report# : 2000011(pdf)
Emergency Preparedness Significance:        Mar 30, 2001 Identified By: NRC Item Type: VIO Violation FAILURE TO CORRECT SELF-IDENTIFIED ERO AUGMENTATION DRILL DEFICIENCIES.
During a baseline inspection of the emergency preparedness program conducted on August 14 - September 21, 2000, the NRC identified a preliminary White issue and potential violation for the licensees failure to successfully correct deficiencies identified during staff augmentation drills and to demonstrate timely staff augmentation in 1999 and 2000. The issue was unresolved pending the outcome of the NRCs final significance determination. On January 30, 2001, the NRC conducted a regulatory conference with the licensee and subsequently issued the licensee a White finding and Notice of Violation (Enforcement Action No. 00-214)
 
1Q/2002 Inspection Findings - Kewaunee                                                                                    Page 8 of 12 associated with the performance issue.
Inspection Report# : 2001007(pdf)
Significance:        Mar 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation SUPPLEMENTAL INSPECTION OF WHITE ERO AUGMENTATION FINDING AND RESULTING GREEN FINDING.
This supplemental inspection was performed by the NRC to evaluate the licensees evaluation associated with the failure to conduct successful quarterly, off-hours, unannounced staff augmentation drills during the second, third, and fourth quarters of 1999 and the second quarter of 2000. This performance issue was previously characterized as having low to moderate risk significance (White) in NRC Inspection Report No. 50-305/2000015(DRS). During this supplemental inspection, performed in accordance with Inspection Procedure95002, the inspector concluded that the licensee performed a comprehensive evaluation of the unsuccessful staff augmentation drills. The licensees evaluation identifiedtwo root causes which resulted in the unsuccessful drills and in the staffs inability tocorrect thedeficiencies: (1) Management has not effectively acted to provide increased depth and flexibility in the emergency response organization following a reduction in staffing several years ago; and (2) Management has accepted an adverse trend of test failures without requiring investigation into the root causes. The inspector reviewed the licensees corrective actions, both completed and planned, and concluded that the programmatic corrective actions appeared to address the identified root causes. In particular, the licensee assigned certain positions to an on-call rotation to ensure personnel were capable of augmenting in a timely manner, and the licensee was progressing in training additional staff to increase the depth of personnel assigned to key emergency response positions. In addition, the licensee was continuing its efforts in improving its corrective action program. The inspector reviewed the licensees immediate response to the issue and identified that one of the licensees initial corrective actions resulted in a Non-Cited Violation of regulatory requirements. To obtain a timely response of a key emergency response position (severe accident management - core hydraulics), the emergency preparedness staff effectively changed the emergency plan without revising the necessary procedures and without formally assessing the impact of that change. The staff instructed and trained personnel to respond to a location other than the Technical Support Center, which was contrary to the licensees current implementing procedures. While this change enabled the licensee to augment its staff in a timely manner, the change was not performed in accordance with NRC requirements. In order to make such a change, the licensees emergency plan required that the change be formally assessed to ensure that it did not reduce the effectiveness of the plan or any other implementing procedure.
Since the issue did not result in a failure to meet an emergency preparedness planning standard, the failure to adequately implement the emergency plan was determined to be a violation of very low safety significance (Green) (Section 02.3(a)). Due to the licensees acceptable performance in assessing the emergency response augmentation drill deficiencies, the White finding associated with this issue will only be considered in assessing plant performance for a total of four quarters in accordance with the guidance in IMC 0305, Operating Reactor Assessment Program. Implementation of the licensees corrective actions will be reviewed during a future inspection.
Inspection Report# : 2001007(pdf)
Significance:        Mar 23, 2001 Identified By: NRC Item Type: NCV NonCited Violation POST-ACCIDENT SAMPLING SYSTEM CONTAINMENT AIR SAMPLE PANEL TESTING.
A Non-Cited Violation of Technical Specification 6.14 was identified for the failure to implement a program that ensured the capability to obtain and analyze containment atmosphere samples under accident samples using the cantainment air sampling panel (CASP). Although the CASP was installed, as was indicated in the emergency plan, the licensee had neither developed procedures nor had tested its capability to obtain a containment atmosphere sample using the CASP. The licensee could not recall if and when containment air samples were last obtained using the CASP.
Inspection Report# : 2001006(pdf)
Significance: N/A Feb 27, 2001 Identified By: NRC Item Type: FIN Finding SUPPLEMENTAL INSPECTION OF YELLOW ANS PERFORMANCE INDICATOR AND ASSOCIATED CORRECTIVE ACTION PROGRAM DEFICIENCIES.
This supplemental inspection was performed by the NRC to assess the licensee's evaluation associated with a Yellow performance indicator for the Alert and Notification System (ANS) and the associated Yellow finding related to the licensee's corrective action program. During this supplemental inspection, performed in accordance with Inspection Procedure 95002, the inspectors concluded that the licensee performed comprehensive evaluations of the performance problems associated with the ANS and its corrective action program. These evaluations identified primary root causes and contributing causes for both issues. Along with the electronics/hardware problems, the licensee identified the primary root causes for the ANS to be the failure to make changes to the system via a change control process and the failure of the activation procedure to provide for alternate/backup activation methods and to provide clear success criteria. In the case of the corrective action program, the licensee concluded that the primary root causes were plant management's inadequate risk evaluation regarding decisions affecting the corrective action program and the development of a culture in the licensee's organization, which minimized the importance of information from outside organizations. In particular, the licensee focused on low-cost power operation and failed to recognize the value of changes and improvements which
 
1Q/2002 Inspection Findings - Kewaunee                                                                                      Page 9 of 12 had occurred throughout the industry, such as the value and expansion of the corrective action program. In the case of the Yellow ANS performance indicator, the licensee had completed several significant corrective actions to address the root causes and contributing causes identified in its evaluation. The inspectors found that the corrective actions appeared appropriate to address the underlying root causes and that ANS testing data indicated an improving trend in the NRC performance indicator. The licensee also performed comprehensive assessments of the emergency preparedness program, quality assurance program, plant operations, and other plant programs to determine the extent of condition (re. the root causes described above). Based on these evaluations, the licensee began to implement significant actions to correct the deficiencies in the corrective action program and other weaknesses identified. Generally, the inspectors observed progress in the licensee's initial implementation of these corrective actions. Due to the licensee's acceptable performance in assessing the Yellow ANS performance indicator and the associated Yellow finding, the Yellow finding will not be considered in assessing future plant performance.
Inspection Report# : 2001005(pdf)
Significance:        Sep 21, 2000 Identified By: NRC Item Type: FIN Finding INADEQUATE ROOT CAUSE EVALUATION FOR YELLOW ALERT AND NOTIFICATION SYSTEM PERFORMANCE INDICATOR.
The licensee's evaluation of the Yellow Alert and Notification (siren) System Performance Indicator (PI) was inadequate. The inspector concluded that the licensee's evaluation was not performed at the depth necessary to identify the root causes of the siren performance problems and, instead, only identified the symptoms of the root causes. Specifically, the inspector identified the following substantive weaknesses in the licensee's evaluation of the siren system performance, which appeared to result from systemic corrective action program deficiencies within this cornerstone:
* The licensee's evaluation was not of sufficient depth to clearly identify the root causes associated with the decline in siren system performance.
* Licensee management did not provide well-understood and clear guidance/expectations for performing root cause evaluations.
* The licensee's evaluation of the quality assurance program was narrowly focused and was not critical of its role in failing to identify and correct the siren performance problems.
* The licensee did not establish a priority for each of the long-term corrective actions in accordance with the associated significance or risk.
* The licensee did not have any formal provisions for measuring the effectiveness of its corrective actions.
* Within the licensee's evaluation, the licensee had not evaluated common causes or the extent of the condition. Due to the corrective action program performance deficiencies within this cornerstone, we have been unable to conclude that the performance issues that resulted in the yellow PI have been addressed. Therefore, we are issuing a yellow finding that corresponds to the original issues that resulted in a yellow PI. Additional inspection effort will be focused on the licensee's further evaluation of the siren reliability root causes and the continuing corrective action program implementation deficiencies identified during this inspection.
Inspection Report# : 2000017(pdf)
Significance: N/A Apr 05, 2000 Identified By: NRC Item Type: FIN Finding LICENSEE FAILED TO IDENTIFY THE FULL SCOPE OF PROBLEMS WITH THE ALERT AND NOTIFICATION SYSTEM PERFORMANCE.
The inspectors concluded that the licensee's assessment was not sufficiently comprehensive to identify the full scope of problems associated with the Alert and Notification System (ANS) performance program. As a result, licensee corrective actions generally were focused on the equipment problem rather than all root causes and contributing factors. For example, the inspectors identified that: (1) management oversight of the ANS performance program was limited; (2) an audit failed to identify degrading ANS performance as a concern; (3) annual preventive maintenance was not consistently performed on the system; (4) the corrective action program was not used consistently to document ANS problems; and (5) maintenance procedures and records were deficient.
Collectively, these problems indicate that the ANS performance program lacked sufficient structure and oversight.
Inspection Report# : 2000006(pdf)
Occupational Radiation Safety Significance:        Oct 02, 2001 Identified By: NRC Item Type: NCV NonCited Violation HIGH RADIATION AREA ACCESS CONTROLS Non-Cited Violation of Technical Specification 6.13 and an associated Green Finding for failure to 'barricade' three ladders that provided entry to high radiation areas (less than 1000 mrem/hour) located on the steam generator/pressurizer platforms.
Inspection Report# : 2001014(pdf)
 
1Q/2002 Inspection Findings - Kewaunee                                                                                  Page 10 of 12 Significance:      May 19, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO POST A VERY HIGH RADIATION AREA.
The inspectors identified a noncited violation for the failure to post a very high radiation area in accordance with 10 CFR 20.1902(c).
Although the area was not adequately posted, the licensee had provided physical controls and barriers that were consistent with its requirements for a very high radiation area. Based on the adequacy of these controls, the potential for an overexposure from the inadvertent entry of personnel into the area was low. Consequently, this finding was determined to be of very low safety significance (Section 20S1.1).
Inspection Report# : 2000009(pdf)
Public Radiation Safety Physical Protection Significance: N/A Jul 21, 2000 Identified By: NRC Item Type: FIN Finding CORRECTIVE ACTION ON SEARCH ISSUES DID NOT WORK.
The inspector determined that the licensee's effectiveness of implemented corrective actions for a previously identified inspection finding regarding an inadequate vehicle search was not totally effective in preventing recurrence. Previous corrective action was not adequately focused (Section 3PP2.2).
Inspection Report# : 2000013(pdf)
Significance:      Jul 21, 2000 Identified By: NRC Item Type: NCV NonCited Violation SEARCHES INADEQUATE (REPEAT).
The inspector identified a Non-Cited violation by observing that, a security officer failed to search an easily accessible compartment on one vehicle. The failure resulted from human error because the officer did not observe the access panel to the compartment during the vehicle search process. When searched, no prohibited items were found. Corrective actions were implemented. The inspector determined that the licensee's effectiveness of implemented corrective actions for a previously identified inspection finding regarding an inadequate vehicle search was not totally effective in preventing recurrence. Previous corrective action was not adequately focused.
Inspection Report# : 2000013(pdf)
Miscellaneous Significance: N/A Feb 21, 2002 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Implement Required Fire Watch Following Completion of Hot Work Activities A licensee-identified violation was reviewed by the inspectors. Corrective actions taken or planned by the licensee appeared reasonable.
Inspection Report# : 2001017(pdf)
Significance: N/A Aug 24, 2001 Identified By: NRC Item Type: FIN Finding IDENTIFICATION AND RESOLUTION OF PROBLEMS The team concluded that the licensee was generally effective at identifying problems and putting them into the corrective action program. The program itself contained all the necessary attributes of an acceptable corrective action program and was generally successful in correcting identified issues. However, the team noted that, although licensee management had taken efforts to ensure
 
1Q/2002 Inspection Findings - Kewaunee                                                                                  Page 11 of 12 that issues were resolved in accordance with program guidance and requirements, additional efforts appeared necessary to ensure timely resolution of issues. A positive program initiative was the establishment of positions in each of the major plant departments to serve as liaisons between the departments and the corrective action program and to assist with self-assessments. However, examples were identified by the inspectors of problems with the licensee's identification and resolution of problems, prioritization and evaluation of issues, and the effectiveness of corrective actions. Included in these examples were the routine granting of due date extensions for problem evaluation and corrective action implementation, failure to perform a Technical Specification-required biennial surveillance of safety-related procedures, and ineffective corrective actions that resulted in the lack of proper controls over the storage of grease used in safety-related breakers. Based on a review of records and discussions with plant staff, the inspectors concluded that workers at the site felt free to input safety issues into the corrective action program.
Inspection Report# : 2001012(pdf)
Significance: N/A Nov 09, 2000 Identified By: NRC Item Type: FIN Finding FAILURE TO TAKE ADEQUATE CORRECTIVE ACTIONS TO ENSURE COMPONENT RE-TEST REQUIREMENTS COMPLETED ACCORDING TO MAINTENANCE PROCEDURES.
No Color. The inspectors determined that a negative performance trend had developed in the licensee's ability to identify and promptly take appropriate corrective actions to prevent recurrence based on two previously identified examples (NCV 50-305/2000014-01) and one example identified during this inspection period (NCV 50-305/00-20-01). All three examples related to the licensee returning safety-related equipment to service prior to completing all required post-maintenance retesting. While the risk of the individual examples was very low, the licensee had failed to ensure that all retest requirements had been completed before returning safety-related equipment to service. These findings collectively indicated a problem with the licensee's ability to provide timely and adequate corrective actions to prevent recurrence. (Section 4OA2).
Inspection Report# : 2000020(pdf)
Significance: N/A Sep 29, 2000 Identified By: NRC Item Type: FIN Finding EFFECTIVENESS OF CORRECTIVE ACTION PROGRAM.
Based on the results of this inspection, the NRC concluded that the corrective action program at Kewaunee showed significant weaknesses and inconsistencies across all of the procedural elements inspected. These weaknesses existed across departments and affected multiple cornerstones in the strategic performance areas of Reactor Safety, Radiation Safety, and Safeguards. Of particular note was the lack of procedures for determining the significance of conditions adverse to quality and for trending of issues and the complete lack of trending within your corrective action program. We also identified a lack of urgency in correcting issues which resulted in repeat examples occurring and, coupled with a poor tracking system, a tendency for issues to be dropped. While none of the specific examples identified by the team were of high risk significance when looked at in isolation, in the aggregate they were similar in nature to prior issues in the emergency preparedness area that rose to a higher significance level and contributed to a degraded cornerstone. While we concluded that the station had fostered an environment in which personnel freely identified conditions adverse to quality without fear of discrimination or retaliation, we also concluded that significant weaknesses with, and inconsistent implementation of, the station corrective action program resulted in multiple examples where station personnel did not enter deficiencies into the station's formal corrective action program.
Inspection Report# : 2000019(pdf)
Significance: N/A Sep 29, 2000 Identified By: NRC Item Type: NCV NonCited Violation NO PROCEDURAL GUIDANCE FOR DETERMINING IF CONDITIONS ADVERSE TO QUALITY ARE SIGNIFICANT.
The inspectors reviewed the quality assurance (QA) manual requirements against Kewaunee's implementing procedures and identified that two QA manual requirements were not being implemented. Specifically, Program Requirement 3.1.9 which stated that directives and procedures shall provide for the review of conditions adverse to quality to determine if the conditions are significant in nature. This requirement paralleled 10 CFR Part 50, Appendix B, Criterion XVI, which requires that the cause of significant conditions adverse to quality be determined and corrective actions taken to prevent recurrence. The inspectors reviewed the Nuclear Administrative Directive (NAD 11.08) and the procedure (GNP 11.08.01) governing the KAP (Kewaunee corrective action program) process and found no procedure requirements for identifying significant conditions adverse to quality. This finding does not directly affect a cornerstone. As a result, this issue was not evaluated with the Significance Determination Process and was not assigned a color. One example of a non-cited violation of 10 CFR Part 50, Appendix B, Criterion II, "Quality Assurance Programs," was identified.
Inspection Report# : 2000019(pdf)
Significance: N/A Sep 29, 2000 Identified By: NRC Item Type: NCV NonCited Violation NO PROCEDURAL GUIDANCE FOR TRENDING CONDITIONS ADVERSE TO QUALITY.
The inspectors reviewed the quality assurance (QA) manual requirements against Kewaunee's implementing procedures and identified that two QA manual requirements were not being implemented. Specifically, QA Program Requirement, 3.1.10, stated that directives and procedures shall provide for analyzing trends of conditions adverse to quality. Once identified these trends were
 
1Q/2002 Inspection Findings - Kewaunee                                                                                Page 12 of 12 required to be considered significant conditions adverse to quality. The inspectors found that conditions adverse to quality were not defined in the KAP (Kewaunee corrective action program) procedures and that no procedure existed for trending. This finding does not directly affect a cornerstone. As a result, this issue was not evaluated with the Significance Determination Process and was not assigned a color. One example of a non-cited violation of 10 CFR Part 50, Appendix B, Criterion II, "Quality Assurance Program,"
was identified.
Inspection Report# : 2000019(pdf)
Significance: N/A Jun 22, 2000 Identified By: NRC Item Type: FIN Finding CONTROL ROOM OPERATIONS HUMAN PERFORMANCE ISSUES.
The inspectors interviewed operators to evaluate their awareness of degraded control room indications and alarms, and their ability to adequately take manual actions based on degraded alarm functions. The inspectors identified, during interviews, that there was a lack of awareness by operators of a degraded refueling water storage tank low-low level alarm which would be potentially confusing to operators and therefore increase the risk associated with initiating long term sump recirculation.
Inspection Report# : 2000008(pdf)
Last modified : July 22, 2002
 
2Q/2002 Inspection Findings - Kewaunee                                                                        Page 1 of 16 Kewaunee Initiating Events Mitigating Systems Significance:        Jun 30, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to incorporate adquate acceptance criteria in service water procedure Green. The licensee failed to incorporate vendor information in a note contained in an operations procedure. The inaccurate note resulted in a service water pump being inappropriately declared operable. This finding was determined to be a Non-Cited Violation of Technical Specification 6.8.a, "Procedures".
Inspection Report# : 2002003(pdf)
Significance:        Jun 30, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to meet battery surveillance Technical Specification requirements Green. The licensee failed to measure and record safety-related battery cell electrolyte levels on a quarterly basis due to surveillance procedure inadequacies, which inhibited the licensee's ability to monitor and trend battery cell performance. Technical Specifications 4.6.b.2 and 4.6.b.3 required that the licensee measure and record battery cell electrolyte level on a quarterly basis. A Non-Cited Violation was identified.
Inspection Report# : 2002003(pdf)
Significance:        Jun 30, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain design control of component cooling water pumps Green. The licensee failed to adequately maintain design control of the component cooling water pumps, which resulted in the inability of a redundant train component cooling pump to provide cooling of safety-related loads due to the likely failure of the pump following a safety injection actuation. This finding was determined to be a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control.
Inspection Report# : 2002003(pdf)
Significance:        Feb 21, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform Thorough 10 CFR 50.59 Safety Evaluation file://C:\RROP\NRR\OVERSIGHT\ASSESS\KEWA\kewa_pim.html                                                          07/03/2003
 
2Q/2002 Inspection Findings - Kewaunee                                                                            Page 2 of 16 The inspectors identified a Non-Cited Violation for failure to perform an adequate 10 CFR 50.59 safety evaluation associated with emergency operating procedure changes to address component cooling water pump dead-head operational concerns. The safety evaluation did not evaluate the potential for initiating a loss-of-coolant accident via the reactor coolant loop seals during conditions of a complete loss of component cooling water.
Inspection Report# : 2001017(pdf)
Significance: N/A Sep 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ESTABLISH CONTINGENCY PLANS FOR ORANGE RISK CONDITION The inspectors identified the failure to establish contingency plans during a planned high risk plant configuration.
Contrary to administrative requirements, the licensee approved an orange risk condition during a refueling outage with no contingency plans to mitigate the consequences of a loss of spent fuel pool cooling with a full core offload in the pool. A Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings,"
was identified. The finding was of very low safety significance because although the licensee had not approved appropriate contingency actions for the orange risk condition, the licensee subsequently rescheduled the planned maintenance to eliminate the orange risk condition.
Inspection Report# : 2001013(pdf)
Significance: N/A Aug 24, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO CONDUCT BIENNIAL SURVIELLANCE OF SAFETY-RELATED PROCEDURES PER TS 6.8.c A Non-Cited Violation of Technical Specification 6.8.c was identified for the failure to perform a biennial surveillance of safety-related procedues. This issue was more than minor because if left uncorrected, could under the same condition become a more significant safety concern. However, since no specific cornerstone had been impacted, this finding is designated as No Color.
Inspection Report# : 2001012(pdf)
Significance:      Aug 24, 2001 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE CORRECTIVE ACTIONS FOR PREVIOUS PROBLEM WITH THE CONTROL OF SAFETY-RELATED MATERIALS A Non-Cited Violation of Criterion XVI, "Corrective Action," of Appendix B of 10 CFR Part 50 was identified for ineffective corrective actions for a problem with the control of the storage of consumable materials, such as thread sealant, used in safety-related applications. These ineffective actions subsequently resulted in the inadequate control of the storage of grease used in safety-related breakers. This issue was more than minor because if left uncorrected, could under the same condition become a more significant safety concern. In that this issue could credibly affect the operability, availability, reliability, or function of a system or train in a mitigating system, it is a Green finding.
Inspection Report# : 2001012(pdf)
Significance:      Aug 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation NON-RATED FIRE BARRIER A Non-Cited Violation [of 10 CFR Part 50, Appendix R, Section III.G.2.a] was identified for failure to provide a 3-file://C:\RROP\NRR\OVERSIGHT\ASSESS\KEWA\kewa_pim.html                                                              07/03/2003
 
2Q/2002 Inspection Findings - Kewaunee                                                                            Page 3 of 16 hour rated fire barrier to separate redundant trains of safe shutdown equipment. This finding was of very low safety significance because the licensee tested a replica of the fire barrier and demonstrated that the fire barrier provided protection for at least 60 minutes, which was sufficient for the hazards in the area.
Inspection Report# : 2001011(pdf)
Significance:      Jul 20, 2001 Identified By: NRC Item Type: FIN Finding LICENSED OPERATOR REQUALIFICATION EXAMINATION RESULTS The inspectors identified that two of eight crews examined during the licensee's calendar year 2001 licensed operator requalification operating test had failed. The finding was of very low safety significance because both crews that had failed received remedial training prior to being returned to shift, and the results of the licensee's operator licensing requalification operating test given in calendar year 2000 indicated that only one crew, out of a total of eight crews tested, had failed.
Inspection Report# : 2001011(pdf)
Significance: N/A Jun 30, 2001 Identified By: NRC Item Type: FIN Finding FAILURE TO BALANCE RELIABILITY AND AVAILABILITY AS REQUIRED BY 10 CFR 50.65(a)(3)
The inspectors identified a failure to evaluate whether adjustments were necessary such that there would be an appropriate balance between systems' availability and reliability in accordance with 10 CFR 50.65(a)(3) of the maintenance rule. The inspectors identified that the licensee did not have an administrative process to track maintenance rule functional failures and maintenance preventible maintenance functional failures. As a result, reliability and availability could not be balanced as required by the Maintenance Rule periodic evaluation. The safety significance of the specific finding was very low because it did not affect the operability of the systems, and the licensee entered the finding in the corrective action program. However, this finding was considered to be of regulatory concern in the area of maintenance rule implementation due to the extent of the problems with the Maintenance Rule Program.
Inspection Report# : 2001009(pdf)
Significance:      Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IDENTIFY DEFICIENT CONDITION OF VALVE AFW-1B The inspectors identified that the licensee failed to promptly identify and correct the B' train auxiliary feedwater pump discharge check valve which was stuck in an intermediate position. A Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was issued. The finding was of very low safety significance because, although the check valve was stuck in an intermediate position, the time that it was known to have been stuck was less than the technical specification allowed outage time for one train of auxiliary feedwater to be out of service (less than 72 hours). Additionally, the other two trains of auxiliary feedwater were each capable of 100 percent decay heat removal.
Inspection Report# : 2001009(pdf)
Significance:      Jun 30, 2001 Identified By: Licensee file://C:\RROP\NRR\OVERSIGHT\ASSESS\KEWA\kewa_pim.html                                                            07/03/2003
 
2Q/2002 Inspection Findings - Kewaunee                                                                        Page 4 of 16 Item Type: NCV NonCited Violation FAILURE TO TRACK UNAVAILABILITY OF SYSTEMS REQUIRED DURING SHUTDOWN OPERATION 10 CFR 50.65(a)(1), required, in part, that the licensee monitor the performance or condition of SSCs within the scope of the rule as defined by 10 CFR 50.65(b), against licensee-established goals, in a manner sufficient to provide reasonable assurance that such SSCs are capable of fulfilling their intended functions. 10 CFR 50.65(a)(2) stated, in part, that monitoring as specified in 10 CFR 50.65(a)(1) was not required where it had been demonstrated that the performance or condition of an SSC was being effectively controlled through the performance of appropriate preventive maintenance, such that the SSC remained capable of performing its intended function. Contrary to the above, the licensee failed to demonstrate that the performance or condition of systems required to be available during shutdown conditions and within the scope of the rule had been effectively controlled through the performance of appropriate preventive maintenance and did not monitor against licensee-established goals. Specifically, the licensee failed to monitor the unavailability of systems required during shutdown operation.
Inspection Report# : 2001009(pdf)
Significance:        Jun 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO ESTABLISH MAINTENANCE RULE (a)(1) GOALS 10 CFR 50.65(a)(1), requires, in part, that the licensee monitor the performance or condition of SSCs within the scope of the rule as defined by 10 CFR 50.65(b), against licensee-established goals, in a manner sufficient to provide reasonable assurance that such structures, systems, and components, are capable of fulfilling their intended functions.
Such goals shall be established commensurate with safety. When the performance or condition of a structure, system, or component does not meet established goals, appropriate corrective action shall be taken. Contrary to the above, from 1996, the licensee did not take appropriate corrective actions when the performance of those systems in (a)(1) did not meet licensee established goals. Specifically, the licensee determined timely and appropriate corrective actions had not been taken for five systems that had been in (a)(1) category for approximately 3 years to 5 years: component cooling (entered (a)(1) on April 23, 1997), control room air conditioning (July 24, 1996), station and instrument air (July 3, 1997), auxiliary building air ventilation (July 31, 1997), and control rod drive (August 6, 1998). This issue is in the licensee's corrective action system as KAP WO 01-3323. The inspectors evaluated the risk significance of this issue using the Significance Determination Process. The inspectors did not identify where this failure resulted in a total loss of a risk significant SSC. Therefore, this issue was screened as Green (very low risk significance) after a Phase 1 Significance Determination Process review. Although the risk significance of this issue was low, the inspectors concluded that this was more than a minor concern because the failure to recognize and correct ineffective maintenance practices resulted in risk significant systems in (a)(1) for years with no improvement in performance. The NRC tracking number for this issue is 50-305/01-09-02.
Inspection Report# : 2001009(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation FIRE DOOR FUSIBLE LINKS.
On February 20, 2001, the licensee determined that the installed fusible link arrangement on roll-up fire Doors 279 and 281, which separated both trains of service water pumps, would not actuate as designed to ensure that the doors would automatically close to provide a 3-hour fire barrier, contrary to 10 CFR Part 50, Appendix R, Section III.G.2.a which required, in part, separation of cables and equipment of redundant trains by a fire barrier having a 3-hour rating.
Inspection Report# : 2001006(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\KEWA\kewa_pim.html                                                          07/03/2003
 
2Q/2002 Inspection Findings - Kewaunee                                                                            Page 5 of 16 Significance:      Mar 12, 2001 Identified By: Licensee Item Type: NCV NonCited Violation INADEQUATE SMOKE DETECTOR COVERAGE IN FIRE ZONE TU-95B.
Licensee identified violation of licensee's operating license that the licensee failed to install a detector in each beam pocket in Fire Zone TU-95B.
Inspection Report# : 2001002(pdf)
Significance:      Feb 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO TEST FIRE DOOR IN ACCORDANCE WITH FIRE PLAN.
The inspectors identified a non-cited violation for failure to properly test a fire door in accordance with the facility's fire protection program plan. The finding was of very low safety significance because, although the fire door separated both trains of service water pumps and did not fully close as designed when subsequently tested, the fire loading in the area was insufficient to credibly impact more than two of the four service water pumps in the area.
Inspection Report# : 2001004(pdf)
Significance: N/A Nov 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO COMPLETE COMPONENT RE-TEST REQUIREMENTS PRIOR TO RETURNING EQUIPMENT TO AN OPERABLE STATUS.
No Color. The inspectors identified a Non-Cited Violation for failure to complete component retest requirements following maintenance performed on the B train control room air conditioner compressor condenser. The unit had been returned to an operable status prior to the retest requirements being completed as prescribed in the associated maintenance procedure. This issue was determined to be a violation of the licensee's Operational Quality Assurance Program Manual, Section 8, "Maintenance Planning and Control." Although the risk associated with this finding was very low and did not affect any cornerstones, the inspectors noted that this finding was similar to previous NRC-identified findings and therefore was of greater than minor significance and warranted documentation. (Section 1R19).
Inspection Report# : 2000020(pdf)
Significance:      Sep 29, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IDENTIFY CORROSION AS POTENTIAL FAILURE MECHANISM.
The inspectors identified that the licensee failed to identify corrosion as a potential failure mechanism in the operability determination for a carbon steel key in the service water system. Thus, the licensee failed to quantify the corrosion rate and therefore did not adequately evaluate the expected service life of the carbon steel key. One non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was identified.
Inspection Report# : 2000019(pdf)
Significance:      Sep 29, 2000 Identified By: NRC Item Type: NCV NonCited Violation file://C:\RROP\NRR\OVERSIGHT\ASSESS\KEWA\kewa_pim.html                                                            07/03/2003
 
2Q/2002 Inspection Findings - Kewaunee                                                                        Page 6 of 16 PRESSURE RATING OF AUXILIARY FEEDWATER STEAM TRAPS.
The inspectors identified that a root cause evaluation for a 1996 equipment issue in the turbine-driven auxiliary feedwater system was not completed until 1999. The evaluation stated that the internals of the steam traps were designed to operate at pressures up to a maximum 600 psig but that the traps were exposed to pressures up to 1025 psig. A corrective action item to initiate a design change request to replace the steam traps with a different model rated for the design pressure of the system was described in the evaluation. However, the inspectors identified that the design change request had never been initiated and the KAP ( Kewaunee corrective action document) had been closed. As a result, the corrective action item for this design problem was lost. In addition, operability of the system had never been formally evaluated despite the identification that the system design requirements were not met. The licensee subsequently determined that the steam traps remained operable and was planning to initiate the design change to correct the problem. One non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was identified.
Inspection Report# : 2000019(pdf)
Significance: N/A Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation RETRIEVAL OF SERVICE WATER SYSTEM DESIGN INFORMATION.
In many cases, design basis information for the service water system was difficult if not impossible to locate. Licensee personnel wrote KAP WO 00-002566 to enter the problem in the corrective action program.
Inspection Report# : 2000012(pdf)
Significance:        Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PROPERLY CONTROL CORRECTED TEST DATA NECESSARY FOR DESIGN CALCULATIONS.
An example of a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," was identified in the handling of service water system flow test data, which was subsequently used in calculations. Gauge readings corrected for post test calibration checks, gauge reading corrections for elevation considerations, and flow values corrected for pump degradation were contained in spreadsheets in the possession of an individual staff member, but not currently packaged with raw test data, and not bearing evidence of a formal review and control process. The connection between the test data, which had been vaulted, and the values used in the calculation, could not be made without use of the uncontrolled spreadsheet.
Inspection Report# : 2000012(pdf)
Significance:        Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation DESIGN CONTROL VIOLATION FOR AUXILIARY FEEDWATER STRAINER MESH SIZE.
The inspectors questioned the mesh size of the strainers, which were installed in the suction of the three auxiliary feedwater (AFW) pumps. As a result of the inspectors' questions, license personnel inspected the strainers on August 21, 2000, and found the strainers to have 1/16 inch openings. A note was later found on Figure 10.2-3 of the UFSAR that indicated that the AFW suction strainer size was 1/8 inch. The smaller openings would not support the use of service water as a safety related source for AFW and as a result all three trains of AFW were declared inoperable. This condition had apparently existed for approximately 25 years and was identified as a non-cited violation of Criterion III, "Design Control," of 10 CFR 50, Appendix B.
Inspection Report# : 2000012(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\KEWA\kewa_pim.html                                                          07/03/2003
 
2Q/2002 Inspection Findings - Kewaunee                                                                        Page 7 of 16 Significance:      Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation IMPROPER DESIGN CALCULATION IDENTIFICATION, NON-CONSERVATIVE ASSUMPTIONS, CALCULATION ERRORS, AND DUPLICATE CALCULATIONS.
An example of a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," was identified because of inadequate control of design calculations. The control failures included improper identification of calculations, non-conservative assumptions, calculation errors, and duplicate or superceded calculations not properly identified or canceled. The failure to follow the established design control process increased the potential for errors in the design and operation of the service water system. Because the system was subsequently demonstrated to be capable of removing the design heat load, the actual significance was low and this finding screened out as having very low risk significance.
Inspection Report# : 2000012(pdf)
Significance:      Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO DOCUMENT IN THE CORRECTIVE ACTION PROGRAM THE USE OF INCORRECT MATERIAL IN A SERVICE WATER PUMP KEY.
An example of a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was identified because of inadequate corrective action to correct an incorrect coupling adjust nut set screw and a low strength "soft" key material, which had contributed to a pump shaft failure. Licensee personnel had known of the "soft" key material since July 21, 1999. The "soft" key material was found in other service water pumps but had not been removed from all pumps. As of July 25, 2000, licensee personnel had not documented the existence of the "soft" key material in the corrective action program.
Inspection Report# : 2000012(pdf)
Significance:      Jul 07, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO COMPLETE COMPONENT RETEST REQUIREMENTS IN ACCORDANCE WITH PROCEDURE.
On June 26, 2000, during a review of post maintenance testing requirements following maintenance performed on the control room post accident system charcoal filter heat detector, the inspectors identified that maintenance technicians had not completed the component re-test requirements, as required by a preventative maintenance procedure prior to the system being returned to an operable status. On July 7, the inspectors identified a second example of failing to complete component re-test requirements following maintenance on the zone special ventilation system charcoal filter heat detector as required. The issue was considered to be of very low safety significance based on the determination that although the licensee had not completed all of the component retest requirements prior to returning the equipment to service, the subsequent testing determined that the equipment was in an operable status. The failure to complete the component retest requirements in accordance with site procedures was identified as a Non-Cited Violation.
Inspection Report# : 2000014(pdf)
Significance:      Jun 22, 2000 Identified By: NRC file://C:\RROP\NRR\OVERSIGHT\ASSESS\KEWA\kewa_pim.html                                                          07/03/2003
 
2Q/2002 Inspection Findings - Kewaunee                                                                        Page 8 of 16 Item Type: FIN Finding FAILURE TO MEET SINGLE FAILURE CRITERIA FOR RESIDUAL HEAT REMOVAL VALVE CIRCUITRY.
The licensee identified that the circuitry associated with the residual heat removal system discharge to safety injection system suction isolation valves did not meet single failure criteria. The inspectors noted that this design requirement was identified in the facility's updated safety analysis report. The licensee subsequently implemented a temporary change to the facility. The inspectors reviewed the issue and identified that the facility had been operating outside of its design basis, which was reportable to the NRC. The licensee subsequently made a one hour non-emergency report to the NRC. Since there was no actual loss of safety function to the system, this issue was screened as very low risk significance.
Inspection Report# : 2000008(pdf)
Significance:        Jun 22, 2000 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO TEST ADDITIONAL RELIEF VALVES IN ACCORDANCE WITH TECHINCAL SPECIFICATIONS.
The licensee identified that the suction relief valve for an auxiliary feedwater pump may have failed its relief test criteria, but did not process the documented deficiency until several weeks later. The licensee then expanded the scope of the relief testing to the suction relief valves associated with the other auxiliary feedwater pumps to meet technical specification requirements. Since any one train of auxiliary feedwater was capable of supplying 100 percent of the decay heat removal requirements, this issue was screened as very low risk significance. However, the time delay in complying with technical specification requirements for testing other relief valves was identified as an NCV.
Inspection Report# : 2000008(pdf)
Significance:        Jun 22, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO INITIATE KEWAUNEE ASSESSMENT PROCESS DOCUMENT REGARDING REFUELING WATER STORAGE TANK LOW-LOW LEVEL ALARM INOPERABILITY.
The inspectors identified that the refueling water storage tank low-low level alarm which was actuating five percent higher than normal had not been documented in a Kewaunee Assessment Process form by the licensee, and therefore had not received an operability evaluation. This failure was identified as contrary to site administrative procedures.
Following the licensee's documentation of the problem, the inspectors identified that the associated operability evaluation considered the acceptability of an operator workaround to address the issue, but did not address any safety implications or consequences of the alarm actuating early. A subsequent operability evaluation by the licensee was evaluated as adequate by the inspectors. Since the subsequent operability evaluation was adequate and it was determined that no safety mitigation equipment was adversely affected by the early actuation of the alarm, this issue was considered of very low risk significance. A non-cited violation (NCV) was identified for failing to document a non-conforming condition, contrary to site administrative procedure requirements.
Inspection Report# : 2000008(pdf)
Significance:        Jun 22, 2000 Identified By: NRC Item Type: FIN Finding MAINTENANCE RULE FAILURES ASSOCIATED WITH REACTOR HEAD VENT VALVE.
The inspectors reviewed the licensee's implementation of the maintenance rule for failures associated with a reactor file://C:\RROP\NRR\OVERSIGHT\ASSESS\KEWA\kewa_pim.html                                                          07/03/2003
 
2Q/2002 Inspection Findings - Kewaunee                                                                          Page 9 of 16 head vent valve. The licensee's corrective action documents identified a potential maintenance rule functional failure but the completed evaluation of the problem did not document the final determination. However, the inspectors identified that the repeated failures may have been prevented if maintenance activities such as valve disassembly and cleaning had been performed. In this case, maintenance rule reliability goals were not exceeded. The licensee had documented similar maintenance rule program deficiencies and developed a corrective action program to address the deficiencies. Although programmatic deficiencies exist, since no maintenance rule reliability criteria had been exceeded, this issue was considered of very low risk significance.
Inspection Report# : 2000008(pdf)
Significance:        May 22, 2000 Identified By: NRC Item Type: FIN Finding FIRE EXTINGUISHERS NOT LOCATED IN ALL AREAS OF CONTAINMENT BUILDING AT BEGINNING OF REFUELING OUTAGE.
During a walkdown of the containment building, the inspectors identified that portable fire extinguishers were not located in the containment basement at the beginning of the plant refueling outage. Additionally, site fire protection procedures required that responsible fire protection personnel perform inspections of selected plant areas to ensure that the quantity of combustible material was minimized. However, the procedure did not list the containment as an area to be inspected and the procedures did not require the placement and location of portable fire suppression equipment inside containment during the refueling outage. Due to a low number of work activities ongoing at the time, this issue was screened as Green (very low risk significance).
Inspection Report# : 2000007(pdf)
Significance:        May 22, 2000 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO INSTALL RAYCHEM HEAT SHRINK MATERIAL IN ACCORDANCE WITH PROCEDURE REQUIREMENTS.
The licensee identified that two Raychem electrical cable splices utilized in environmentally qualified (EQ) safety-related equipment had not been installed in accordance with EQ requirements. These splices were associated with pressurizer level transmitters and were installed in 1984. The licensee subsequently performed extensive EQ testing of the splices to determine the qualification of the splices' as-found configurations. Test results indicated that the splices would have been able to perform their intended function in a harsh environment inside containment. This issue was considered to be of low safety significance based on the successful EQ testing of the as-found splices' configurations and was screened as Green (very low risk significance). One non-cited violation was identified.
Inspection Report# : 2000007(pdf)
Barrier Integrity Significance:        Jun 22, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation FAILURE TO UPDATE COMPUTER ALARM FOR CURRENT AXIAL FLUX DISTRIBUTION TARGET BAND.
file://C:\RROP\NRR\OVERSIGHT\ASSESS\KEWA\kewa_pim.html                                                            07/03/2003
 
2Q/2002 Inspection Findings - Kewaunee                                                                        Page 10 of 16 The licensee identified, following plant startup, that a computer alarm had not been updated properly to alarm if axial flux distribution deviated outside of the flux distribution target band. This condition was contrary to technical specification requirements. The licensee reviewed the axial flux distribution history since the startup and determined that at no time was the flux distribution outside of the target band. Since the axial flux distribution was never outside of the target band, this issue was screened as very low risk significance. An NCV was identified for failing to comply with technical specification requirements for monitoring axial flux distributions.
Inspection Report# : 2000008(pdf)
Significance:      May 22, 2000 Identified By: Licensee Item Type: FIN Finding TECHNICAL SPECIFICATIONS INTERPRETATION FOR TESTING REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES.
The NRC determined that the licensee's practice of testing reactor coolant system pressure isolation Valve SI-22B prior to entering the cold shutdown condition was contrary to Technical Specification requirements 4.2.a.3.a. Technical Specification 4.2.a.3.a required that periodic leakage testing of Valve SI-22B be accomplished prior to reaching operating mode after the plant was placed in cold shutdown. This issue was considered to be of low safety significance because of a subsequent successful valve test and was screened as Green (very low risk significance). Enforcement discretion was applied to this item in accordance with Section VII.B.6 of the Enforcement Policy.
Inspection Report# : 2000007(pdf)
Significance:      May 19, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PROPERLY IMPLEMENT FLAW ACCEPTANCE CRITERIA FOR SLEEVE WELD INSPECTIONS.
During 1998 inservice inspection examinations, the licensee failed to properly implement the flaw acceptance criteria for laser welded sleeve inspection within two steam generator tubes. The safety significance was very low based on the absence of adverse consequences, and May 2000 in-situ pressure testing where both welds exhibited zero leakage at normal operating pressure, main steam line break pressure, and three times normal operating differential pressures. As such, this issue was characterized as a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion IX, Control of Special Processes. To correct the error, the licensee plugged both tubes (Section 1RO8).
Inspection Report# : 2000011(pdf)
Emergency Preparedness Significance:      Mar 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation SUPPLEMENTAL INSPECTION OF WHITE ERO AUGMENTATION FINDING AND RESULTING GREEN FINDING.
This supplemental inspection was performed by the NRC to evaluate the licensees evaluation associated with the failure to conduct successful quarterly, off-hours, unannounced staff augmentation drills during the second, third, and fourth quarters of 1999 and the second quarter of 2000. This performance issue was previously characterized as having file://C:\RROP\NRR\OVERSIGHT\ASSESS\KEWA\kewa_pim.html                                                          07/03/2003
 
2Q/2002 Inspection Findings - Kewaunee                                                                          Page 11 of 16 low to moderate risk significance (White) in NRC Inspection Report No. 50-305/2000015(DRS). During this supplemental inspection, performed in accordance with Inspection Procedure95002, the inspector concluded that the licensee performed a comprehensive evaluation of the unsuccessful staff augmentation drills. The licensees evaluation identifiedtwo root causes which resulted in the unsuccessful drills and in the staffs inability tocorrect thedeficiencies:
(1) Management has not effectively acted to provide increased depth and flexibility in the emergency response organization following a reduction in staffing several years ago; and (2) Management has accepted an adverse trend of test failures without requiring investigation into the root causes. The inspector reviewed the licensees corrective actions, both completed and planned, and concluded that the programmatic corrective actions appeared to address the identified root causes. In particular, the licensee assigned certain positions to an on-call rotation to ensure personnel were capable of augmenting in a timely manner, and the licensee was progressing in training additional staff to increase the depth of personnel assigned to key emergency response positions. In addition, the licensee was continuing its efforts in improving its corrective action program. The inspector reviewed the licensees immediate response to the issue and identified that one of the licensees initial corrective actions resulted in a Non-Cited Violation of regulatory requirements. To obtain a timely response of a key emergency response position (severe accident management - core hydraulics), the emergency preparedness staff effectively changed the emergency plan without revising the necessary procedures and without formally assessing the impact of that change. The staff instructed and trained personnel to respond to a location other than the Technical Support Center, which was contrary to the licensees current implementing procedures. While this change enabled the licensee to augment its staff in a timely manner, the change was not performed in accordance with NRC requirements. In order to make such a change, the licensees emergency plan required that the change be formally assessed to ensure that it did not reduce the effectiveness of the plan or any other implementing procedure. Since the issue did not result in a failure to meet an emergency preparedness planning standard, the failure to adequately implement the emergency plan was determined to be a violation of very low safety significance (Green) (Section 02.3(a)). Due to the licensees acceptable performance in assessing the emergency response augmentation drill deficiencies, the White finding associated with this issue will only be considered in assessing plant performance for a total of four quarters in accordance with the guidance in IMC 0305, Operating Reactor Assessment Program. Implementation of the licensees corrective actions will be reviewed during a future inspection.
Inspection Report# : 2001007(pdf)
Significance:      Mar 30, 2001 Identified By: NRC Item Type: VIO Violation FAILURE TO CORRECT SELF-IDENTIFIED ERO AUGMENTATION DRILL DEFICIENCIES.
During a baseline inspection of the emergency preparedness program conducted on August 14 - September 21, 2000, the NRC identified a preliminary White issue and potential violation for the licensees failure to successfully correct deficiencies identified during staff augmentation drills and to demonstrate timely staff augmentation in 1999 and 2000.
The issue was unresolved pending the outcome of the NRCs final significance determination. On January 30, 2001, the NRC conducted a regulatory conference with the licensee and subsequently issued the licensee a White finding and Notice of Violation (Enforcement Action No. 00-214) associated with the performance issue.
Inspection Report# : 2001007(pdf)
Significance:      Mar 23, 2001 Identified By: NRC Item Type: NCV NonCited Violation POST-ACCIDENT SAMPLING SYSTEM CONTAINMENT AIR SAMPLE PANEL TESTING.
A Non-Cited Violation of Technical Specification 6.14 was identified for the failure to implement a program that ensured the capability to obtain and analyze containment atmosphere samples under accident samples using the cantainment air sampling panel (CASP). Although the CASP was installed, as was indicated in the emergency plan, the file://C:\RROP\NRR\OVERSIGHT\ASSESS\KEWA\kewa_pim.html                                                            07/03/2003
 
2Q/2002 Inspection Findings - Kewaunee                                                                        Page 12 of 16 licensee had neither developed procedures nor had tested its capability to obtain a containment atmosphere sample using the CASP. The licensee could not recall if and when containment air samples were last obtained using the CASP.
Inspection Report# : 2001006(pdf)
Significance: N/A Feb 27, 2001 Identified By: NRC Item Type: FIN Finding SUPPLEMENTAL INSPECTION OF YELLOW ANS PERFORMANCE INDICATOR AND ASSOCIATED CORRECTIVE ACTION PROGRAM DEFICIENCIES.
This supplemental inspection was performed by the NRC to assess the licensee's evaluation associated with a Yellow performance indicator for the Alert and Notification System (ANS) and the associated Yellow finding related to the licensee's corrective action program. During this supplemental inspection, performed in accordance with Inspection Procedure 95002, the inspectors concluded that the licensee performed comprehensive evaluations of the performance problems associated with the ANS and its corrective action program. These evaluations identified primary root causes and contributing causes for both issues. Along with the electronics/hardware problems, the licensee identified the primary root causes for the ANS to be the failure to make changes to the system via a change control process and the failure of the activation procedure to provide for alternate/backup activation methods and to provide clear success criteria. In the case of the corrective action program, the licensee concluded that the primary root causes were plant management's inadequate risk evaluation regarding decisions affecting the corrective action program and the development of a culture in the licensee's organization, which minimized the importance of information from outside organizations. In particular, the licensee focused on low-cost power operation and failed to recognize the value of changes and improvements which had occurred throughout the industry, such as the value and expansion of the corrective action program. In the case of the Yellow ANS performance indicator, the licensee had completed several significant corrective actions to address the root causes and contributing causes identified in its evaluation. The inspectors found that the corrective actions appeared appropriate to address the underlying root causes and that ANS testing data indicated an improving trend in the NRC performance indicator. The licensee also performed comprehensive assessments of the emergency preparedness program, quality assurance program, plant operations, and other plant programs to determine the extent of condition (re. the root causes described above). Based on these evaluations, the licensee began to implement significant actions to correct the deficiencies in the corrective action program and other weaknesses identified. Generally, the inspectors observed progress in the licensee's initial implementation of these corrective actions. Due to the licensee's acceptable performance in assessing the Yellow ANS performance indicator and the associated Yellow finding, the Yellow finding will not be considered in assessing future plant performance.
Inspection Report# : 2001005(pdf)
Significance:        Sep 21, 2000 Identified By: NRC Item Type: FIN Finding INADEQUATE ROOT CAUSE EVALUATION FOR YELLOW ALERT AND NOTIFICATION SYSTEM PERFORMANCE INDICATOR.
The licensee's evaluation of the Yellow Alert and Notification (siren) System Performance Indicator (PI) was inadequate. The inspector concluded that the licensee's evaluation was not performed at the depth necessary to identify the root causes of the siren performance problems and, instead, only identified the symptoms of the root causes.
Specifically, the inspector identified the following substantive weaknesses in the licensee's evaluation of the siren system performance, which appeared to result from systemic corrective action program deficiencies within this cornerstone:
* The licensee's evaluation was not of sufficient depth to clearly identify the root causes associated with the decline in siren system performance.
* Licensee management did not provide well-understood and clear guidance/expectations for performing root cause evaluations.
* The licensee's evaluation of the quality assurance program was narrowly focused and was not critical of its role in failing to identify and correct the siren performance file://C:\RROP\NRR\OVERSIGHT\ASSESS\KEWA\kewa_pim.html                                                          07/03/2003
 
2Q/2002 Inspection Findings - Kewaunee                                                                        Page 13 of 16 problems.
* The licensee did not establish a priority for each of the long-term corrective actions in accordance with the associated significance or risk.
* The licensee did not have any formal provisions for measuring the effectiveness of its corrective actions.
* Within the licensee's evaluation, the licensee had not evaluated common causes or the extent of the condition. Due to the corrective action program performance deficiencies within this cornerstone, we have been unable to conclude that the performance issues that resulted in the yellow PI have been addressed. Therefore, we are issuing a yellow finding that corresponds to the original issues that resulted in a yellow PI. Additional inspection effort will be focused on the licensee's further evaluation of the siren reliability root causes and the continuing corrective action program implementation deficiencies identified during this inspection.
Inspection Report# : 2000017(pdf)
Significance: N/A Apr 05, 2000 Identified By: NRC Item Type: FIN Finding LICENSEE FAILED TO IDENTIFY THE FULL SCOPE OF PROBLEMS WITH THE ALERT AND NOTIFICATION SYSTEM PERFORMANCE.
The inspectors concluded that the licensee's assessment was not sufficiently comprehensive to identify the full scope of problems associated with the Alert and Notification System (ANS) performance program. As a result, licensee corrective actions generally were focused on the equipment problem rather than all root causes and contributing factors.
For example, the inspectors identified that: (1) management oversight of the ANS performance program was limited; (2) an audit failed to identify degrading ANS performance as a concern; (3) annual preventive maintenance was not consistently performed on the system; (4) the corrective action program was not used consistently to document ANS problems; and (5) maintenance procedures and records were deficient. Collectively, these problems indicate that the ANS performance program lacked sufficient structure and oversight.
Inspection Report# : 2000006(pdf)
Occupational Radiation Safety Significance:        Oct 02, 2001 Identified By: NRC Item Type: NCV NonCited Violation HIGH RADIATION AREA ACCESS CONTROLS Non-Cited Violation of Technical Specification 6.13 and an associated Green Finding for failure to 'barricade' three ladders that provided entry to high radiation areas (less than 1000 mrem/hour) located on the steam generator/pressurizer platforms.
Inspection Report# : 2001014(pdf)
Significance:        May 19, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO POST A VERY HIGH RADIATION AREA.
The inspectors identified a noncited violation for the failure to post a very high radiation area in accordance with 10 CFR 20.1902(c). Although the area was not adequately posted, the licensee had provided physical controls and barriers that were consistent with its requirements for a very high radiation area. Based on the adequacy of these controls, the potential for an overexposure from the inadvertent entry of personnel into the area was low. Consequently, this finding was determined to be of very low safety significance (Section 20S1.1).
file://C:\RROP\NRR\OVERSIGHT\ASSESS\KEWA\kewa_pim.html                                                          07/03/2003
 
2Q/2002 Inspection Findings - Kewaunee                                                                      Page 14 of 16 Inspection Report# : 2000009(pdf)
Public Radiation Safety Physical Protection Significance: N/A Jul 21, 2000 Identified By: NRC Item Type: FIN Finding CORRECTIVE ACTION ON SEARCH ISSUES DID NOT WORK.
The inspector determined that the licensee's effectiveness of implemented corrective actions for a previously identified inspection finding regarding an inadequate vehicle search was not totally effective in preventing recurrence. Previous corrective action was not adequately focused (Section 3PP2.2).
Inspection Report# : 2000013(pdf)
Significance:      Jul 21, 2000 Identified By: NRC Item Type: NCV NonCited Violation SEARCHES INADEQUATE (REPEAT).
The inspector identified a Non-Cited violation by observing that, a security officer failed to search an easily accessible compartment on one vehicle. The failure resulted from human error because the officer did not observe the access panel to the compartment during the vehicle search process. When searched, no prohibited items were found. Corrective actions were implemented. The inspector determined that the licensee's effectiveness of implemented corrective actions for a previously identified inspection finding regarding an inadequate vehicle search was not totally effective in preventing recurrence. Previous corrective action was not adequately focused.
Inspection Report# : 2000013(pdf)
Miscellaneous Significance: N/A Feb 21, 2002 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Implement Required Fire Watch Following Completion of Hot Work Activities A licensee-identified violation was reviewed by the inspectors. Corrective actions taken or planned by the licensee appeared reasonable.
Inspection Report# : 2001017(pdf)
Significance: N/A Aug 24, 2001 Identified By: NRC Item Type: FIN Finding IDENTIFICATION AND RESOLUTION OF PROBLEMS The team concluded that the licensee was generally effective at identifying problems and putting them into the file://C:\RROP\NRR\OVERSIGHT\ASSESS\KEWA\kewa_pim.html                                                          07/03/2003
 
2Q/2002 Inspection Findings - Kewaunee                                                                          Page 15 of 16 corrective action program. The program itself contained all the necessary attributes of an acceptable corrective action program and was generally successful in correcting identified issues. However, the team noted that, although licensee management had taken efforts to ensure that issues were resolved in accordance with program guidance and requirements, additional efforts appeared necessary to ensure timely resolution of issues. A positive program initiative was the establishment of positions in each of the major plant departments to serve as liaisons between the departments and the corrective action program and to assist with self-assessments. However, examples were identified by the inspectors of problems with the licensee's identification and resolution of problems, prioritization and evaluation of issues, and the effectiveness of corrective actions. Included in these examples were the routine granting of due date extensions for problem evaluation and corrective action implementation, failure to perform a Technical Specification-required biennial surveillance of safety-related procedures, and ineffective corrective actions that resulted in the lack of proper controls over the storage of grease used in safety-related breakers. Based on a review of records and discussions with plant staff, the inspectors concluded that workers at the site felt free to input safety issues into the corrective action program.
Inspection Report# : 2001012(pdf)
Significance: N/A Nov 09, 2000 Identified By: NRC Item Type: FIN Finding FAILURE TO TAKE ADEQUATE CORRECTIVE ACTIONS TO ENSURE COMPONENT RE-TEST REQUIREMENTS COMPLETED ACCORDING TO MAINTENANCE PROCEDURES.
No Color. The inspectors determined that a negative performance trend had developed in the licensee's ability to identify and promptly take appropriate corrective actions to prevent recurrence based on two previously identified examples (NCV 50-305/2000014-01) and one example identified during this inspection period (NCV 50-305/00 01). All three examples related to the licensee returning safety-related equipment to service prior to completing all required post-maintenance retesting. While the risk of the individual examples was very low, the licensee had failed to ensure that all retest requirements had been completed before returning safety-related equipment to service. These findings collectively indicated a problem with the licensee's ability to provide timely and adequate corrective actions to prevent recurrence. (Section 4OA2).
Inspection Report# : 2000020(pdf)
Significance: N/A Sep 29, 2000 Identified By: NRC Item Type: FIN Finding EFFECTIVENESS OF CORRECTIVE ACTION PROGRAM.
Based on the results of this inspection, the NRC concluded that the corrective action program at Kewaunee showed significant weaknesses and inconsistencies across all of the procedural elements inspected. These weaknesses existed across departments and affected multiple cornerstones in the strategic performance areas of Reactor Safety, Radiation Safety, and Safeguards. Of particular note was the lack of procedures for determining the significance of conditions adverse to quality and for trending of issues and the complete lack of trending within your corrective action program.
We also identified a lack of urgency in correcting issues which resulted in repeat examples occurring and, coupled with a poor tracking system, a tendency for issues to be dropped. While none of the specific examples identified by the team were of high risk significance when looked at in isolation, in the aggregate they were similar in nature to prior issues in the emergency preparedness area that rose to a higher significance level and contributed to a degraded cornerstone.
While we concluded that the station had fostered an environment in which personnel freely identified conditions adverse to quality without fear of discrimination or retaliation, we also concluded that significant weaknesses with, and inconsistent implementation of, the station corrective action program resulted in multiple examples where station personnel did not enter deficiencies into the station's formal corrective action program.
Inspection Report# : 2000019(pdf)
Significance: N/A Sep 29, 2000 Identified By: NRC file://C:\RROP\NRR\OVERSIGHT\ASSESS\KEWA\kewa_pim.html                                                            07/03/2003
 
2Q/2002 Inspection Findings - Kewaunee                                                                      Page 16 of 16 Item Type: NCV NonCited Violation NO PROCEDURAL GUIDANCE FOR DETERMINING IF CONDITIONS ADVERSE TO QUALITY ARE SIGNIFICANT.
The inspectors reviewed the quality assurance (QA) manual requirements against Kewaunee's implementing procedures and identified that two QA manual requirements were not being implemented. Specifically, Program Requirement 3.1.9 which stated that directives and procedures shall provide for the review of conditions adverse to quality to determine if the conditions are significant in nature. This requirement paralleled 10 CFR Part 50, Appendix B, Criterion XVI, which requires that the cause of significant conditions adverse to quality be determined and corrective actions taken to prevent recurrence. The inspectors reviewed the Nuclear Administrative Directive (NAD 11.08) and the procedure (GNP 11.08.01) governing the KAP (Kewaunee corrective action program) process and found no procedure requirements for identifying significant conditions adverse to quality. This finding does not directly affect a cornerstone. As a result, this issue was not evaluated with the Significance Determination Process and was not assigned a color. One example of a non-cited violation of 10 CFR Part 50, Appendix B, Criterion II, "Quality Assurance Programs," was identified.
Inspection Report# : 2000019(pdf)
Significance: N/A Sep 29, 2000 Identified By: NRC Item Type: NCV NonCited Violation NO PROCEDURAL GUIDANCE FOR TRENDING CONDITIONS ADVERSE TO QUALITY.
The inspectors reviewed the quality assurance (QA) manual requirements against Kewaunee's implementing procedures and identified that two QA manual requirements were not being implemented. Specifically, QA Program Requirement, 3.1.10, stated that directives and procedures shall provide for analyzing trends of conditions adverse to quality. Once identified these trends were required to be considered significant conditions adverse to quality. The inspectors found that conditions adverse to quality were not defined in the KAP (Kewaunee corrective action program) procedures and that no procedure existed for trending. This finding does not directly affect a cornerstone. As a result, this issue was not evaluated with the Significance Determination Process and was not assigned a color. One example of a non-cited violation of 10 CFR Part 50, Appendix B, Criterion II, "Quality Assurance Program," was identified.
Inspection Report# : 2000019(pdf)
Significance: N/A Jun 22, 2000 Identified By: NRC Item Type: FIN Finding CONTROL ROOM OPERATIONS HUMAN PERFORMANCE ISSUES.
The inspectors interviewed operators to evaluate their awareness of degraded control room indications and alarms, and their ability to adequately take manual actions based on degraded alarm functions. The inspectors identified, during interviews, that there was a lack of awareness by operators of a degraded refueling water storage tank low-low level alarm which would be potentially confusing to operators and therefore increase the risk associated with initiating long term sump recirculation.
Inspection Report# : 2000008(pdf)
Last modified : August 29, 2002 file://C:\RROP\NRR\OVERSIGHT\ASSESS\KEWA\kewa_pim.html                                                        07/03/2003
 
3Q/2002 Inspection Findings - Kewaunee                                                                        Page 1 of 17 Kewaunee Initiating Events Mitigating Systems Significance:        Sep 30, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to Provide Appendix R Barriers Between Dedicated and Alternate Fire Zones A finding of very low risk significance was identified by the inspectors for the licensee's failure to provide fire barrier seals on auxiliary building Appendix R walls separating the Dedicated and Alternate fire zones.
Inspection Report# : 2002005(pdf)
Significance: TBD Sep 30, 2002 Identified By: NRC Item Type: URI Unresolved item Adequacy of Medical Examinations The inspectors identified an apparent violation of medical requirement regulations, 10 CFR 55.21, "Medical Examination," and 10 CFR 55.23, "Certification," in that the licensee's medical evaluations appeared to be inadequate in reference to ANSI/ANS 3.4-1983, and failed to adequately implement all the required medical testing. The finding is greater than minor, but is unresolved pending completion of the licensee's investigation into the medical issue, subsequent NRC review, and completion of a significance determination.
Inspection Report# : 2002005(pdf)
Significance: TBD Sep 30, 2002 Identified By: NRC Item Type: URI Unresolved item Adequacy of the Plant-Referenced Simulator to Conform with Simulator Requirements in 10 CFR 55.46 The inspectors identified an apparent violation of the simulator fidelity regulation, 10 CFR 55.46, "Simulation Facilities," in that the licensee's maintenance of simulator core modeling and simulator fidelity appeared to not comply with ANSI/ANS-3.5-1985. The finding is greater than minor, but is unresolved pending completion of the licensee's core modeling testing and investigation, subsequent NRC review of the core testing data, and completion of a significance determination for this issue. On September 19, 2002, the inspectors identified three issues concerning the potential failure to comply with 10 CFR 55.46, "Simulation Facilities." The first issue concerned the licensee's use of the simulator to meet experience requirements for applicants for initial operator and senior operator licenses in accordance with 10 CFR 55.46 (c)(2)(I). The second issue concerned the adequacy of the licensee conducting periodic simulator performance testing throughout the life of the simulator. The third issue concerned the licensee's program for correcting simulator modeling and hardware discrepancies, including discrepancies identified from performance testing in accordance with 10 CFR 55.46 (d)(2).
Inspection Report# : 2002005(pdf)
Significance: TBD Sep 20, 2002 Identified By: NRC Item Type: AV Apparent Violation Failure to Provide Fixed Suppression System in Fire Area TU-95B During performance of follow-up activities in response to a USNRC inspection, the licensee identified that fire area TU-95B had been misclassified in that it should have been classified as required to meet the requirements of Section
 
3Q/2002 Inspection Findings - Kewaunee                                                                            Page 2 of 17 III.G.3 of 10 CFR Part 50, Appendix R. An apparent violation of 10 CFR Part 50, Appendix R, Section III.G.3 was identified for the failure to provide fire area TU-95B with a fixed fire suppression system. This issue has been preliminarily determined to have low to moderate safety significance (White). As a result of failing to have a fixed fire suppression system, there was a greater likelihood that a fire in fire area TU-95B would not be suppressed and redundant trains of cables and equipment required for safe shutdown could be damaged. The corresponding damage could require a shutdown of the plant from outside the control room, significantly increasing the complexity of manual actions required to achieve safe shutdown.
Inspection Report# : 2002006(pdf)
Significance:        Sep 20, 2002 Identified By: NRC Item Type: NCV NonCited Violation Inappropriate Placement of Diesel Generator Room Heat Detectors During performance of a triennial fire protection inspection, USNRC Region III staff identified that heat detectors used for activation of a diesel generator room carbon dioxide (CO2) system were not located and installed in accordance with the applicable National Fire Protection Association (NFPA) code. Specifically, no heat detectors were located at the ceiling level. The failure to appropriately locate and install heat detectors for actuation of the CO2 system is a violation of the Kewaunee Nuclear Power Plant operating license. The finding was greater than minor because it affected the protection against external factors (i.e., fire) attribute for mitigating systems. As a result of the inadequate heat detector placement, actuation of the carbon dioxide system in the diesel generator room could be delayed. The finding was of very low safety significance because the inspector was not able to identify a fire scenario in which safety significant cables would be damaged prior to actuation of the carbon dioxide system.
Inspection Report# : 2002006(pdf)
Significance:        Jun 30, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to incorporate adquate acceptance criteria in service water procedure Green. The licensee failed to incorporate vendor information in a note contained in an operations procedure. The inaccurate note resulted in a service water pump being inappropriately declared operable. This finding was determined to be a Non-Cited Violation of Technical Specification 6.8.a, "Procedures".
Inspection Report# : 2002003(pdf)
Significance:        Jun 30, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to meet battery surveillance Technical Specification requirements Green. The licensee failed to measure and record safety-related battery cell electrolyte levels on a quarterly basis due to surveillance procedure inadequacies, which inhibited the licensee's ability to monitor and trend battery cell performance. Technical Specifications 4.6.b.2 and 4.6.b.3 required that the licensee measure and record battery cell electrolyte level on a quarterly basis. A Non-Cited Violation was identified.
Inspection Report# : 2002003(pdf)
Significance:        Jun 30, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain design control of component cooling water pumps Green. The licensee failed to adequately maintain design control of the component cooling water pumps, which resulted in the inability of a redundant train component cooling pump to provide cooling of safety-related loads due to
 
3Q/2002 Inspection Findings - Kewaunee                                                                            Page 3 of 17 the likely failure of the pump following a safety injection actuation. This finding was determined to be a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control.
Inspection Report# : 2002003(pdf)
Significance:        Feb 21, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform Thorough 10 CFR 50.59 Safety Evaluation The inspectors identified a Non-Cited Violation for failure to perform an adequate 10 CFR 50.59 safety evaluation associated with emergency operating procedure changes to address component cooling water pump dead-head operational concerns. The safety evaluation did not evaluate the potential for initiating a loss-of-coolant accident via the reactor coolant loop seals during conditions of a complete loss of component cooling water.
Inspection Report# : 2001017(pdf)
Significance: N/A Sep 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO ESTABLISH CONTINGENCY PLANS FOR ORANGE RISK CONDITION The inspectors identified the failure to establish contingency plans during a planned high risk plant configuration.
Contrary to administrative requirements, the licensee approved an orange risk condition during a refueling outage with no contingency plans to mitigate the consequences of a loss of spent fuel pool cooling with a full core offload in the pool. A Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings,"
was identified. The finding was of very low safety significance because although the licensee had not approved appropriate contingency actions for the orange risk condition, the licensee subsequently rescheduled the planned maintenance to eliminate the orange risk condition.
Inspection Report# : 2001013(pdf)
Significance: N/A Aug 24, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO CONDUCT BIENNIAL SURVIELLANCE OF SAFETY-RELATED PROCEDURES PER TS 6.8.c A Non-Cited Violation of Technical Specification 6.8.c was identified for the failure to perform a biennial surveillance of safety-related procedues. This issue was more than minor because if left uncorrected, could under the same condition become a more significant safety concern. However, since no specific cornerstone had been impacted, this finding is designated as No Color.
Inspection Report# : 2001012(pdf)
Significance:        Aug 24, 2001 Identified By: NRC Item Type: NCV NonCited Violation INADEQUATE CORRECTIVE ACTIONS FOR PREVIOUS PROBLEM WITH THE CONTROL OF SAFETY-RELATED MATERIALS A Non-Cited Violation of Criterion XVI, "Corrective Action," of Appendix B of 10 CFR Part 50 was identified for ineffective corrective actions for a problem with the control of the storage of consumable materials, such as thread sealant, used in safety-related applications. These ineffective actions subsequently resulted in the inadequate control of the storage of grease used in safety-related breakers. This issue was more than minor because if left uncorrected, could under the same condition become a more significant safety concern. In that this issue could credibly affect the operability, availability, reliability, or function of a system or train in a mitigating system, it is a Green finding.
Inspection Report# : 2001012(pdf)
 
3Q/2002 Inspection Findings - Kewaunee                                                                            Page 4 of 17 Significance:      Aug 09, 2001 Identified By: NRC Item Type: NCV NonCited Violation NON-RATED FIRE BARRIER A Non-Cited Violation [of 10 CFR Part 50, Appendix R, Section III.G.2.a] was identified for failure to provide a 3-hour rated fire barrier to separate redundant trains of safe shutdown equipment. This finding was of very low safety significance because the licensee tested a replica of the fire barrier and demonstrated that the fire barrier provided protection for at least 60 minutes, which was sufficient for the hazards in the area.
Inspection Report# : 2001011(pdf)
Significance:      Jul 20, 2001 Identified By: NRC Item Type: FIN Finding LICENSED OPERATOR REQUALIFICATION EXAMINATION RESULTS The inspectors identified that two of eight crews examined during the licensee's calendar year 2001 licensed operator requalification operating test had failed. The finding was of very low safety significance because both crews that had failed received remedial training prior to being returned to shift, and the results of the licensee's operator licensing requalification operating test given in calendar year 2000 indicated that only one crew, out of a total of eight crews tested, had failed.
Inspection Report# : 2001011(pdf)
Significance: N/A Jun 30, 2001 Identified By: NRC Item Type: FIN Finding FAILURE TO BALANCE RELIABILITY AND AVAILABILITY AS REQUIRED BY 10 CFR 50.65(a)(3)
The inspectors identified a failure to evaluate whether adjustments were necessary such that there would be an appropriate balance between systems' availability and reliability in accordance with 10 CFR 50.65(a)(3) of the maintenance rule. The inspectors identified that the licensee did not have an administrative process to track maintenance rule functional failures and maintenance preventible maintenance functional failures. As a result, reliability and availability could not be balanced as required by the Maintenance Rule periodic evaluation. The safety significance of the specific finding was very low because it did not affect the operability of the systems, and the licensee entered the finding in the corrective action program. However, this finding was considered to be of regulatory concern in the area of maintenance rule implementation due to the extent of the problems with the Maintenance Rule Program.
Inspection Report# : 2001009(pdf)
Significance:      Jun 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IDENTIFY DEFICIENT CONDITION OF VALVE AFW-1B The inspectors identified that the licensee failed to promptly identify and correct the B' train auxiliary feedwater pump discharge check valve which was stuck in an intermediate position. A Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was issued. The finding was of very low safety significance because, although the check valve was stuck in an intermediate position, the time that it was known to have been stuck was less than the technical specification allowed outage time for one train of auxiliary feedwater to be out of service (less than 72 hours). Additionally, the other two trains of auxiliary feedwater were each capable of 100 percent decay heat removal.
Inspection Report# : 2001009(pdf)
 
3Q/2002 Inspection Findings - Kewaunee                                                                        Page 5 of 17 Significance:        Jun 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO TRACK UNAVAILABILITY OF SYSTEMS REQUIRED DURING SHUTDOWN OPERATION 10 CFR 50.65(a)(1), required, in part, that the licensee monitor the performance or condition of SSCs within the scope of the rule as defined by 10 CFR 50.65(b), against licensee-established goals, in a manner sufficient to provide reasonable assurance that such SSCs are capable of fulfilling their intended functions. 10 CFR 50.65(a)(2) stated, in part, that monitoring as specified in 10 CFR 50.65(a)(1) was not required where it had been demonstrated that the performance or condition of an SSC was being effectively controlled through the performance of appropriate preventive maintenance, such that the SSC remained capable of performing its intended function. Contrary to the above, the licensee failed to demonstrate that the performance or condition of systems required to be available during shutdown conditions and within the scope of the rule had been effectively controlled through the performance of appropriate preventive maintenance and did not monitor against licensee-established goals. Specifically, the licensee failed to monitor the unavailability of systems required during shutdown operation.
Inspection Report# : 2001009(pdf)
Significance:        Jun 30, 2001 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO ESTABLISH MAINTENANCE RULE (a)(1) GOALS 10 CFR 50.65(a)(1), requires, in part, that the licensee monitor the performance or condition of SSCs within the scope of the rule as defined by 10 CFR 50.65(b), against licensee-established goals, in a manner sufficient to provide reasonable assurance that such structures, systems, and components, are capable of fulfilling their intended functions.
Such goals shall be established commensurate with safety. When the performance or condition of a structure, system, or component does not meet established goals, appropriate corrective action shall be taken. Contrary to the above, from 1996, the licensee did not take appropriate corrective actions when the performance of those systems in (a)(1) did not meet licensee established goals. Specifically, the licensee determined timely and appropriate corrective actions had not been taken for five systems that had been in (a)(1) category for approximately 3 years to 5 years: component cooling (entered (a)(1) on April 23, 1997), control room air conditioning (July 24, 1996), station and instrument air (July 3, 1997), auxiliary building air ventilation (July 31, 1997), and control rod drive (August 6, 1998). This issue is in the licensee's corrective action system as KAP WO 01-3323. The inspectors evaluated the risk significance of this issue using the Significance Determination Process. The inspectors did not identify where this failure resulted in a total loss of a risk significant SSC. Therefore, this issue was screened as Green (very low risk significance) after a Phase 1 Significance Determination Process review. Although the risk significance of this issue was low, the inspectors concluded that this was more than a minor concern because the failure to recognize and correct ineffective maintenance practices resulted in risk significant systems in (a)(1) for years with no improvement in performance. The NRC tracking number for this issue is 50-305/01-09-02.
Inspection Report# : 2001009(pdf)
Significance:        Mar 31, 2001 Identified By: NRC Item Type: NCV NonCited Violation FIRE DOOR FUSIBLE LINKS.
On February 20, 2001, the licensee determined that the installed fusible link arrangement on roll-up fire Doors 279 and 281, which separated both trains of service water pumps, would not actuate as designed to ensure that the doors would automatically close to provide a 3-hour fire barrier, contrary to 10 CFR Part 50, Appendix R, Section III.G.2.a which required, in part, separation of cables and equipment of redundant trains by a fire barrier having a 3-hour rating.
Inspection Report# : 2001006(pdf)
 
3Q/2002 Inspection Findings - Kewaunee                                                                            Page 6 of 17 Significance:      Mar 12, 2001 Identified By: Licensee Item Type: NCV NonCited Violation INADEQUATE SMOKE DETECTOR COVERAGE IN FIRE ZONE TU-95B.
Licensee identified violation of licensee's operating license that the licensee failed to install a detector in each beam pocket in Fire Zone TU-95B.
Inspection Report# : 2001002(pdf)
Significance:      Feb 13, 2001 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO TEST FIRE DOOR IN ACCORDANCE WITH FIRE PLAN.
The inspectors identified a non-cited violation for failure to properly test a fire door in accordance with the facility's fire protection program plan. The finding was of very low safety significance because, although the fire door separated both trains of service water pumps and did not fully close as designed when subsequently tested, the fire loading in the area was insufficient to credibly impact more than two of the four service water pumps in the area.
Inspection Report# : 2001004(pdf)
Significance: N/A Nov 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO COMPLETE COMPONENT RE-TEST REQUIREMENTS PRIOR TO RETURNING EQUIPMENT TO AN OPERABLE STATUS.
No Color. The inspectors identified a Non-Cited Violation for failure to complete component retest requirements following maintenance performed on the B train control room air conditioner compressor condenser. The unit had been returned to an operable status prior to the retest requirements being completed as prescribed in the associated maintenance procedure. This issue was determined to be a violation of the licensee's Operational Quality Assurance Program Manual, Section 8, "Maintenance Planning and Control." Although the risk associated with this finding was very low and did not affect any cornerstones, the inspectors noted that this finding was similar to previous NRC-identified findings and therefore was of greater than minor significance and warranted documentation. (Section 1R19).
Inspection Report# : 2000020(pdf)
Significance:      Sep 29, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO IDENTIFY CORROSION AS POTENTIAL FAILURE MECHANISM.
The inspectors identified that the licensee failed to identify corrosion as a potential failure mechanism in the operability determination for a carbon steel key in the service water system. Thus, the licensee failed to quantify the corrosion rate and therefore did not adequately evaluate the expected service life of the carbon steel key. One non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was identified.
Inspection Report# : 2000019(pdf)
Significance:      Sep 29, 2000 Identified By: NRC Item Type: NCV NonCited Violation PRESSURE RATING OF AUXILIARY FEEDWATER STEAM TRAPS.
The inspectors identified that a root cause evaluation for a 1996 equipment issue in the turbine-driven auxiliary feedwater system was not completed until 1999. The evaluation stated that the internals of the steam traps were designed to operate at pressures up to a maximum 600 psig but that the traps were exposed to pressures up to 1025
 
3Q/2002 Inspection Findings - Kewaunee                                                                        Page 7 of 17 psig. A corrective action item to initiate a design change request to replace the steam traps with a different model rated for the design pressure of the system was described in the evaluation. However, the inspectors identified that the design change request had never been initiated and the KAP ( Kewaunee corrective action document) had been closed. As a result, the corrective action item for this design problem was lost. In addition, operability of the system had never been formally evaluated despite the identification that the system design requirements were not met. The licensee subsequently determined that the steam traps remained operable and was planning to initiate the design change to correct the problem. One non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was identified.
Inspection Report# : 2000019(pdf)
Significance:        Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PROPERLY CONTROL CORRECTED TEST DATA NECESSARY FOR DESIGN CALCULATIONS.
An example of a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," was identified in the handling of service water system flow test data, which was subsequently used in calculations. Gauge readings corrected for post test calibration checks, gauge reading corrections for elevation considerations, and flow values corrected for pump degradation were contained in spreadsheets in the possession of an individual staff member, but not currently packaged with raw test data, and not bearing evidence of a formal review and control process. The connection between the test data, which had been vaulted, and the values used in the calculation, could not be made without use of the uncontrolled spreadsheet.
Inspection Report# : 2000012(pdf)
Significance: N/A Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation RETRIEVAL OF SERVICE WATER SYSTEM DESIGN INFORMATION.
In many cases, design basis information for the service water system was difficult if not impossible to locate. Licensee personnel wrote KAP WO 00-002566 to enter the problem in the corrective action program.
Inspection Report# : 2000012(pdf)
Significance:        Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation DESIGN CONTROL VIOLATION FOR AUXILIARY FEEDWATER STRAINER MESH SIZE.
The inspectors questioned the mesh size of the strainers, which were installed in the suction of the three auxiliary feedwater (AFW) pumps. As a result of the inspectors' questions, license personnel inspected the strainers on August 21, 2000, and found the strainers to have 1/16 inch openings. A note was later found on Figure 10.2-3 of the UFSAR that indicated that the AFW suction strainer size was 1/8 inch. The smaller openings would not support the use of service water as a safety related source for AFW and as a result all three trains of AFW were declared inoperable. This condition had apparently existed for approximately 25 years and was identified as a non-cited violation of Criterion III, "Design Control," of 10 CFR 50, Appendix B.
Inspection Report# : 2000012(pdf)
Significance:        Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO DOCUMENT IN THE CORRECTIVE ACTION PROGRAM THE USE OF INCORRECT MATERIAL IN A SERVICE WATER PUMP KEY.
An example of a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was
 
3Q/2002 Inspection Findings - Kewaunee                                                                        Page 8 of 17 identified because of inadequate corrective action to correct an incorrect coupling adjust nut set screw and a low strength "soft" key material, which had contributed to a pump shaft failure. Licensee personnel had known of the "soft" key material since July 21, 1999. The "soft" key material was found in other service water pumps but had not been removed from all pumps. As of July 25, 2000, licensee personnel had not documented the existence of the "soft" key material in the corrective action program.
Inspection Report# : 2000012(pdf)
Significance:      Aug 09, 2000 Identified By: NRC Item Type: NCV NonCited Violation IMPROPER DESIGN CALCULATION IDENTIFICATION, NON-CONSERVATIVE ASSUMPTIONS, CALCULATION ERRORS, AND DUPLICATE CALCULATIONS.
An example of a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," was identified because of inadequate control of design calculations. The control failures included improper identification of calculations, non-conservative assumptions, calculation errors, and duplicate or superceded calculations not properly identified or canceled. The failure to follow the established design control process increased the potential for errors in the design and operation of the service water system. Because the system was subsequently demonstrated to be capable of removing the design heat load, the actual significance was low and this finding screened out as having very low risk significance.
Inspection Report# : 2000012(pdf)
Significance:      Jul 07, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO COMPLETE COMPONENT RETEST REQUIREMENTS IN ACCORDANCE WITH PROCEDURE.
On June 26, 2000, during a review of post maintenance testing requirements following maintenance performed on the control room post accident system charcoal filter heat detector, the inspectors identified that maintenance technicians had not completed the component re-test requirements, as required by a preventative maintenance procedure prior to the system being returned to an operable status. On July 7, the inspectors identified a second example of failing to complete component re-test requirements following maintenance on the zone special ventilation system charcoal filter heat detector as required. The issue was considered to be of very low safety significance based on the determination that although the licensee had not completed all of the component retest requirements prior to returning the equipment to service, the subsequent testing determined that the equipment was in an operable status. The failure to complete the component retest requirements in accordance with site procedures was identified as a Non-Cited Violation.
Inspection Report# : 2000014(pdf)
Significance:      Jun 22, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO INITIATE KEWAUNEE ASSESSMENT PROCESS DOCUMENT REGARDING REFUELING WATER STORAGE TANK LOW-LOW LEVEL ALARM INOPERABILITY.
The inspectors identified that the refueling water storage tank low-low level alarm which was actuating five percent higher than normal had not been documented in a Kewaunee Assessment Process form by the licensee, and therefore had not received an operability evaluation. This failure was identified as contrary to site administrative procedures.
Following the licensee's documentation of the problem, the inspectors identified that the associated operability evaluation considered the acceptability of an operator workaround to address the issue, but did not address any safety implications or consequences of the alarm actuating early. A subsequent operability evaluation by the licensee was evaluated as adequate by the inspectors. Since the subsequent operability evaluation was adequate and it was determined that no safety mitigation equipment was adversely affected by the early actuation of the alarm, this issue was considered of very low risk significance. A non-cited violation (NCV) was identified for failing to document a
 
3Q/2002 Inspection Findings - Kewaunee                                                                        Page 9 of 17 non-conforming condition, contrary to site administrative procedure requirements.
Inspection Report# : 2000008(pdf)
Significance:        Jun 22, 2000 Identified By: NRC Item Type: FIN Finding FAILURE TO MEET SINGLE FAILURE CRITERIA FOR RESIDUAL HEAT REMOVAL VALVE CIRCUITRY.
The licensee identified that the circuitry associated with the residual heat removal system discharge to safety injection system suction isolation valves did not meet single failure criteria. The inspectors noted that this design requirement was identified in the facility's updated safety analysis report. The licensee subsequently implemented a temporary change to the facility. The inspectors reviewed the issue and identified that the facility had been operating outside of its design basis, which was reportable to the NRC. The licensee subsequently made a one hour non-emergency report to the NRC. Since there was no actual loss of safety function to the system, this issue was screened as very low risk significance.
Inspection Report# : 2000008(pdf)
Significance:        Jun 22, 2000 Identified By: NRC Item Type: FIN Finding MAINTENANCE RULE FAILURES ASSOCIATED WITH REACTOR HEAD VENT VALVE.
The inspectors reviewed the licensee's implementation of the maintenance rule for failures associated with a reactor head vent valve. The licensee's corrective action documents identified a potential maintenance rule functional failure but the completed evaluation of the problem did not document the final determination. However, the inspectors identified that the repeated failures may have been prevented if maintenance activities such as valve disassembly and cleaning had been performed. In this case, maintenance rule reliability goals were not exceeded. The licensee had documented similar maintenance rule program deficiencies and developed a corrective action program to address the deficiencies. Although programmatic deficiencies exist, since no maintenance rule reliability criteria had been exceeded, this issue was considered of very low risk significance.
Inspection Report# : 2000008(pdf)
Significance:        Jun 22, 2000 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO TEST ADDITIONAL RELIEF VALVES IN ACCORDANCE WITH TECHINCAL SPECIFICATIONS.
The licensee identified that the suction relief valve for an auxiliary feedwater pump may have failed its relief test criteria, but did not process the documented deficiency until several weeks later. The licensee then expanded the scope of the relief testing to the suction relief valves associated with the other auxiliary feedwater pumps to meet technical specification requirements. Since any one train of auxiliary feedwater was capable of supplying 100 percent of the decay heat removal requirements, this issue was screened as very low risk significance. However, the time delay in complying with technical specification requirements for testing other relief valves was identified as an NCV.
Inspection Report# : 2000008(pdf)
Significance:        May 22, 2000 Identified By: NRC Item Type: FIN Finding FIRE EXTINGUISHERS NOT LOCATED IN ALL AREAS OF CONTAINMENT BUILDING AT BEGINNING OF REFUELING OUTAGE.
 
3Q/2002 Inspection Findings - Kewaunee                                                                        Page 10 of 17 During a walkdown of the containment building, the inspectors identified that portable fire extinguishers were not located in the containment basement at the beginning of the plant refueling outage. Additionally, site fire protection procedures required that responsible fire protection personnel perform inspections of selected plant areas to ensure that the quantity of combustible material was minimized. However, the procedure did not list the containment as an area to be inspected and the procedures did not require the placement and location of portable fire suppression equipment inside containment during the refueling outage. Due to a low number of work activities ongoing at the time, this issue was screened as Green (very low risk significance).
Inspection Report# : 2000007(pdf)
Significance:      May 22, 2000 Identified By: Licensee Item Type: NCV NonCited Violation FAILURE TO INSTALL RAYCHEM HEAT SHRINK MATERIAL IN ACCORDANCE WITH PROCEDURE REQUIREMENTS.
The licensee identified that two Raychem electrical cable splices utilized in environmentally qualified (EQ) safety-related equipment had not been installed in accordance with EQ requirements. These splices were associated with pressurizer level transmitters and were installed in 1984. The licensee subsequently performed extensive EQ testing of the splices to determine the qualification of the splices' as-found configurations. Test results indicated that the splices would have been able to perform their intended function in a harsh environment inside containment. This issue was considered to be of low safety significance based on the successful EQ testing of the as-found splices' configurations and was screened as Green (very low risk significance). One non-cited violation was identified.
Inspection Report# : 2000007(pdf)
Barrier Integrity Significance:      Jun 22, 2000 Identified By: Self Disclosing Item Type: NCV NonCited Violation FAILURE TO UPDATE COMPUTER ALARM FOR CURRENT AXIAL FLUX DISTRIBUTION TARGET BAND.
The licensee identified, following plant startup, that a computer alarm had not been updated properly to alarm if axial flux distribution deviated outside of the flux distribution target band. This condition was contrary to technical specification requirements. The licensee reviewed the axial flux distribution history since the startup and determined that at no time was the flux distribution outside of the target band. Since the axial flux distribution was never outside of the target band, this issue was screened as very low risk significance. An NCV was identified for failing to comply with technical specification requirements for monitoring axial flux distributions.
Inspection Report# : 2000008(pdf)
Significance:      May 22, 2000 Identified By: Licensee Item Type: FIN Finding TECHNICAL SPECIFICATIONS INTERPRETATION FOR TESTING REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES.
The NRC determined that the licensee's practice of testing reactor coolant system pressure isolation Valve SI-22B prior to entering the cold shutdown condition was contrary to Technical Specification requirements 4.2.a.3.a. Technical Specification 4.2.a.3.a required that periodic leakage testing of Valve SI-22B be accomplished prior to reaching operating mode after the plant was placed in cold shutdown. This issue was considered to be of low safety significance because of a subsequent successful valve test and was screened as Green (very low risk significance). Enforcement
 
3Q/2002 Inspection Findings - Kewaunee                                                                          Page 11 of 17 discretion was applied to this item in accordance with Section VII.B.6 of the Enforcement Policy.
Inspection Report# : 2000007(pdf)
Significance:      May 19, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PROPERLY IMPLEMENT FLAW ACCEPTANCE CRITERIA FOR SLEEVE WELD INSPECTIONS.
During 1998 inservice inspection examinations, the licensee failed to properly implement the flaw acceptance criteria for laser welded sleeve inspection within two steam generator tubes. The safety significance was very low based on the absence of adverse consequences, and May 2000 in-situ pressure testing where both welds exhibited zero leakage at normal operating pressure, main steam line break pressure, and three times normal operating differential pressures. As such, this issue was characterized as a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion IX, Control of Special Processes. To correct the error, the licensee plugged both tubes (Section 1RO8).
Inspection Report# : 2000011(pdf)
Emergency Preparedness Significance:      Sep 30, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct an Instrument Deficiency Inspection Report# : 2002005(pdf)
Significance:      Mar 30, 2001 Identified By: NRC Item Type: NCV NonCited Violation SUPPLEMENTAL INSPECTION OF WHITE ERO AUGMENTATION FINDING AND RESULTING GREEN FINDING.
This supplemental inspection was performed by the NRC to evaluate the licensees evaluation associated with the failure to conduct successful quarterly, off-hours, unannounced staff augmentation drills during the second, third, and fourth quarters of 1999 and the second quarter of 2000. This performance issue was previously characterized as having low to moderate risk significance (White) in NRC Inspection Report No. 50-305/2000015(DRS). During this supplemental inspection, performed in accordance with Inspection Procedure95002, the inspector concluded that the licensee performed a comprehensive evaluation of the unsuccessful staff augmentation drills. The licensees evaluation identifiedtwo root causes which resulted in the unsuccessful drills and in the staffs inability tocorrect thedeficiencies:
(1) Management has not effectively acted to provide increased depth and flexibility in the emergency response organization following a reduction in staffing several years ago; and (2) Management has accepted an adverse trend of test failures without requiring investigation into the root causes. The inspector reviewed the licensees corrective actions, both completed and planned, and concluded that the programmatic corrective actions appeared to address the identified root causes. In particular, the licensee assigned certain positions to an on-call rotation to ensure personnel were capable of augmenting in a timely manner, and the licensee was progressing in training additional staff to increase the depth of personnel assigned to key emergency response positions. In addition, the licensee was continuing its efforts in improving its corrective action program. The inspector reviewed the licensees immediate response to the issue and identified that one of the licensees initial corrective actions resulted in a Non-Cited Violation of regulatory requirements. To obtain a timely response of a key emergency response position (severe accident management - core hydraulics), the emergency preparedness staff effectively changed the emergency plan without revising the necessary
 
3Q/2002 Inspection Findings - Kewaunee                                                                      Page 12 of 17 procedures and without formally assessing the impact of that change. The staff instructed and trained personnel to respond to a location other than the Technical Support Center, which was contrary to the licensees current implementing procedures. While this change enabled the licensee to augment its staff in a timely manner, the change was not performed in accordance with NRC requirements. In order to make such a change, the licensees emergency plan required that the change be formally assessed to ensure that it did not reduce the effectiveness of the plan or any other implementing procedure. Since the issue did not result in a failure to meet an emergency preparedness planning standard, the failure to adequately implement the emergency plan was determined to be a violation of very low safety significance (Green) (Section 02.3(a)). Due to the licensees acceptable performance in assessing the emergency response augmentation drill deficiencies, the White finding associated with this issue will only be considered in assessing plant performance for a total of four quarters in accordance with the guidance in IMC 0305, Operating Reactor Assessment Program. Implementation of the licensees corrective actions will be reviewed during a future inspection.
Inspection Report# : 2001007(pdf)
Significance:        Mar 30, 2001 Identified By: NRC Item Type: VIO Violation FAILURE TO CORRECT SELF-IDENTIFIED ERO AUGMENTATION DRILL DEFICIENCIES.
During a baseline inspection of the emergency preparedness program conducted on August 14 - September 21, 2000, the NRC identified a preliminary White issue and potential violation for the licensees failure to successfully correct deficiencies identified during staff augmentation drills and to demonstrate timely staff augmentation in 1999 and 2000.
The issue was unresolved pending the outcome of the NRCs final significance determination. On January 30, 2001, the NRC conducted a regulatory conference with the licensee and subsequently issued the licensee a White finding and Notice of Violation (Enforcement Action No. 00-214) associated with the performance issue.
Inspection Report# : 2001007(pdf)
Significance:        Mar 23, 2001 Identified By: NRC Item Type: NCV NonCited Violation POST-ACCIDENT SAMPLING SYSTEM CONTAINMENT AIR SAMPLE PANEL TESTING.
A Non-Cited Violation of Technical Specification 6.14 was identified for the failure to implement a program that ensured the capability to obtain and analyze containment atmosphere samples under accident samples using the cantainment air sampling panel (CASP). Although the CASP was installed, as was indicated in the emergency plan, the licensee had neither developed procedures nor had tested its capability to obtain a containment atmosphere sample using the CASP. The licensee could not recall if and when containment air samples were last obtained using the CASP.
Inspection Report# : 2001006(pdf)
Significance: N/A Feb 27, 2001 Identified By: NRC Item Type: FIN Finding SUPPLEMENTAL INSPECTION OF YELLOW ANS PERFORMANCE INDICATOR AND ASSOCIATED CORRECTIVE ACTION PROGRAM DEFICIENCIES.
This supplemental inspection was performed by the NRC to assess the licensee's evaluation associated with a Yellow performance indicator for the Alert and Notification System (ANS) and the associated Yellow finding related to the licensee's corrective action program. During this supplemental inspection, performed in accordance with Inspection Procedure 95002, the inspectors concluded that the licensee performed comprehensive evaluations of the performance problems associated with the ANS and its corrective action program. These evaluations identified primary root causes and contributing causes for both issues. Along with the electronics/hardware problems, the licensee identified the primary root causes for the ANS to be the failure to make changes to the system via a change control process and the failure of the activation procedure to provide for alternate/backup activation methods and to provide clear success criteria. In the case of the corrective action program, the licensee concluded that the primary root causes were plant management's inadequate risk evaluation regarding decisions affecting the corrective action program and the
 
3Q/2002 Inspection Findings - Kewaunee                                                                        Page 13 of 17 development of a culture in the licensee's organization, which minimized the importance of information from outside organizations. In particular, the licensee focused on low-cost power operation and failed to recognize the value of changes and improvements which had occurred throughout the industry, such as the value and expansion of the corrective action program. In the case of the Yellow ANS performance indicator, the licensee had completed several significant corrective actions to address the root causes and contributing causes identified in its evaluation. The inspectors found that the corrective actions appeared appropriate to address the underlying root causes and that ANS testing data indicated an improving trend in the NRC performance indicator. The licensee also performed comprehensive assessments of the emergency preparedness program, quality assurance program, plant operations, and other plant programs to determine the extent of condition (re. the root causes described above). Based on these evaluations, the licensee began to implement significant actions to correct the deficiencies in the corrective action program and other weaknesses identified. Generally, the inspectors observed progress in the licensee's initial implementation of these corrective actions. Due to the licensee's acceptable performance in assessing the Yellow ANS performance indicator and the associated Yellow finding, the Yellow finding will not be considered in assessing future plant performance.
Inspection Report# : 2001005(pdf)
Significance:        Sep 21, 2000 Identified By: NRC Item Type: FIN Finding INADEQUATE ROOT CAUSE EVALUATION FOR YELLOW ALERT AND NOTIFICATION SYSTEM PERFORMANCE INDICATOR.
The licensee's evaluation of the Yellow Alert and Notification (siren) System Performance Indicator (PI) was inadequate. The inspector concluded that the licensee's evaluation was not performed at the depth necessary to identify the root causes of the siren performance problems and, instead, only identified the symptoms of the root causes.
Specifically, the inspector identified the following substantive weaknesses in the licensee's evaluation of the siren system performance, which appeared to result from systemic corrective action program deficiencies within this cornerstone:
* The licensee's evaluation was not of sufficient depth to clearly identify the root causes associated with the decline in siren system performance.
* Licensee management did not provide well-understood and clear guidance/expectations for performing root cause evaluations.
* The licensee's evaluation of the quality assurance program was narrowly focused and was not critical of its role in failing to identify and correct the siren performance problems.
* The licensee did not establish a priority for each of the long-term corrective actions in accordance with the associated significance or risk.
* The licensee did not have any formal provisions for measuring the effectiveness of its corrective actions.
* Within the licensee's evaluation, the licensee had not evaluated common causes or the extent of the condition. Due to the corrective action program performance deficiencies within this cornerstone, we have been unable to conclude that the performance issues that resulted in the yellow PI have been addressed. Therefore, we are issuing a yellow finding that corresponds to the original issues that resulted in a yellow PI. Additional inspection effort will be focused on the licensee's further evaluation of the siren reliability root causes and the continuing corrective action program implementation deficiencies identified during this inspection.
Inspection Report# : 2000017(pdf)
Significance: N/A Apr 05, 2000 Identified By: NRC Item Type: FIN Finding LICENSEE FAILED TO IDENTIFY THE FULL SCOPE OF PROBLEMS WITH THE ALERT AND NOTIFICATION SYSTEM PERFORMANCE.
The inspectors concluded that the licensee's assessment was not sufficiently comprehensive to identify the full scope of problems associated with the Alert and Notification System (ANS) performance program. As a result, licensee corrective actions generally were focused on the equipment problem rather than all root causes and contributing factors.
For example, the inspectors identified that: (1) management oversight of the ANS performance program was limited; (2) an audit failed to identify degrading ANS performance as a concern; (3) annual preventive maintenance was not consistently performed on the system; (4) the corrective action program was not used consistently to document ANS problems; and (5) maintenance procedures and records were deficient. Collectively, these problems indicate that the ANS performance program lacked sufficient structure and oversight.
Inspection Report# : 2000006(pdf)
 
3Q/2002 Inspection Findings - Kewaunee                                                                        Page 14 of 17 Occupational Radiation Safety Significance:      Oct 02, 2001 Identified By: NRC Item Type: NCV NonCited Violation HIGH RADIATION AREA ACCESS CONTROLS Non-Cited Violation of Technical Specification 6.13 and an associated Green Finding for failure to 'barricade' three ladders that provided entry to high radiation areas (less than 1000 mrem/hour) located on the steam generator/pressurizer platforms.
Inspection Report# : 2001014(pdf)
Significance:      May 19, 2000 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO POST A VERY HIGH RADIATION AREA.
The inspectors identified a noncited violation for the failure to post a very high radiation area in accordance with 10 CFR 20.1902(c). Although the area was not adequately posted, the licensee had provided physical controls and barriers that were consistent with its requirements for a very high radiation area. Based on the adequacy of these controls, the potential for an overexposure from the inadvertent entry of personnel into the area was low. Consequently, this finding was determined to be of very low safety significance (Section 20S1.1).
Inspection Report# : 2000009(pdf)
Public Radiation Safety Physical Protection Significance: N/A Jul 21, 2000 Identified By: NRC Item Type: FIN Finding CORRECTIVE ACTION ON SEARCH ISSUES DID NOT WORK.
The inspector determined that the licensee's effectiveness of implemented corrective actions for a previously identified inspection finding regarding an inadequate vehicle search was not totally effective in preventing recurrence. Previous corrective action was not adequately focused (Section 3PP2.2).
Inspection Report# : 2000013(pdf)
Significance:      Jul 21, 2000 Identified By: NRC Item Type: NCV NonCited Violation SEARCHES INADEQUATE (REPEAT).
The inspector identified a Non-Cited violation by observing that, a security officer failed to search an easily accessible compartment on one vehicle. The failure resulted from human error because the officer did not observe the access panel to the compartment during the vehicle search process. When searched, no prohibited items were found. Corrective actions were implemented. The inspector determined that the licensee's effectiveness of implemented corrective actions
 
3Q/2002 Inspection Findings - Kewaunee                                                                          Page 15 of 17 for a previously identified inspection finding regarding an inadequate vehicle search was not totally effective in preventing recurrence. Previous corrective action was not adequately focused.
Inspection Report# : 2000013(pdf)
Miscellaneous Significance: N/A Feb 21, 2002 Identified By: Licensee Item Type: NCV NonCited Violation Failure to Implement Required Fire Watch Following Completion of Hot Work Activities A licensee-identified violation was reviewed by the inspectors. Corrective actions taken or planned by the licensee appeared reasonable.
Inspection Report# : 2001017(pdf)
Significance: N/A Aug 24, 2001 Identified By: NRC Item Type: FIN Finding IDENTIFICATION AND RESOLUTION OF PROBLEMS The team concluded that the licensee was generally effective at identifying problems and putting them into the corrective action program. The program itself contained all the necessary attributes of an acceptable corrective action program and was generally successful in correcting identified issues. However, the team noted that, although licensee management had taken efforts to ensure that issues were resolved in accordance with program guidance and requirements, additional efforts appeared necessary to ensure timely resolution of issues. A positive program initiative was the establishment of positions in each of the major plant departments to serve as liaisons between the departments and the corrective action program and to assist with self-assessments. However, examples were identified by the inspectors of problems with the licensee's identification and resolution of problems, prioritization and evaluation of issues, and the effectiveness of corrective actions. Included in these examples were the routine granting of due date extensions for problem evaluation and corrective action implementation, failure to perform a Technical Specification-required biennial surveillance of safety-related procedures, and ineffective corrective actions that resulted in the lack of proper controls over the storage of grease used in safety-related breakers. Based on a review of records and discussions with plant staff, the inspectors concluded that workers at the site felt free to input safety issues into the corrective action program.
Inspection Report# : 2001012(pdf)
Significance: N/A Nov 09, 2000 Identified By: NRC Item Type: FIN Finding FAILURE TO TAKE ADEQUATE CORRECTIVE ACTIONS TO ENSURE COMPONENT RE-TEST REQUIREMENTS COMPLETED ACCORDING TO MAINTENANCE PROCEDURES.
No Color. The inspectors determined that a negative performance trend had developed in the licensee's ability to identify and promptly take appropriate corrective actions to prevent recurrence based on two previously identified examples (NCV 50-305/2000014-01) and one example identified during this inspection period (NCV 50-305/00 01). All three examples related to the licensee returning safety-related equipment to service prior to completing all required post-maintenance retesting. While the risk of the individual examples was very low, the licensee had failed to ensure that all retest requirements had been completed before returning safety-related equipment to service. These findings collectively indicated a problem with the licensee's ability to provide timely and adequate corrective actions to prevent recurrence. (Section 4OA2).
Inspection Report# : 2000020(pdf)
Significance: N/A Sep 29, 2000 Identified By: NRC Item Type: FIN Finding
 
3Q/2002 Inspection Findings - Kewaunee                                                                      Page 16 of 17 EFFECTIVENESS OF CORRECTIVE ACTION PROGRAM.
Based on the results of this inspection, the NRC concluded that the corrective action program at Kewaunee showed significant weaknesses and inconsistencies across all of the procedural elements inspected. These weaknesses existed across departments and affected multiple cornerstones in the strategic performance areas of Reactor Safety, Radiation Safety, and Safeguards. Of particular note was the lack of procedures for determining the significance of conditions adverse to quality and for trending of issues and the complete lack of trending within your corrective action program.
We also identified a lack of urgency in correcting issues which resulted in repeat examples occurring and, coupled with a poor tracking system, a tendency for issues to be dropped. While none of the specific examples identified by the team were of high risk significance when looked at in isolation, in the aggregate they were similar in nature to prior issues in the emergency preparedness area that rose to a higher significance level and contributed to a degraded cornerstone.
While we concluded that the station had fostered an environment in which personnel freely identified conditions adverse to quality without fear of discrimination or retaliation, we also concluded that significant weaknesses with, and inconsistent implementation of, the station corrective action program resulted in multiple examples where station personnel did not enter deficiencies into the station's formal corrective action program.
Inspection Report# : 2000019(pdf)
Significance: N/A Sep 29, 2000 Identified By: NRC Item Type: NCV NonCited Violation NO PROCEDURAL GUIDANCE FOR DETERMINING IF CONDITIONS ADVERSE TO QUALITY ARE SIGNIFICANT.
The inspectors reviewed the quality assurance (QA) manual requirements against Kewaunee's implementing procedures and identified that two QA manual requirements were not being implemented. Specifically, Program Requirement 3.1.9 which stated that directives and procedures shall provide for the review of conditions adverse to quality to determine if the conditions are significant in nature. This requirement paralleled 10 CFR Part 50, Appendix B, Criterion XVI, which requires that the cause of significant conditions adverse to quality be determined and corrective actions taken to prevent recurrence. The inspectors reviewed the Nuclear Administrative Directive (NAD 11.08) and the procedure (GNP 11.08.01) governing the KAP (Kewaunee corrective action program) process and found no procedure requirements for identifying significant conditions adverse to quality. This finding does not directly affect a cornerstone. As a result, this issue was not evaluated with the Significance Determination Process and was not assigned a color. One example of a non-cited violation of 10 CFR Part 50, Appendix B, Criterion II, "Quality Assurance Programs," was identified.
Inspection Report# : 2000019(pdf)
Significance: N/A Sep 29, 2000 Identified By: NRC Item Type: NCV NonCited Violation NO PROCEDURAL GUIDANCE FOR TRENDING CONDITIONS ADVERSE TO QUALITY.
The inspectors reviewed the quality assurance (QA) manual requirements against Kewaunee's implementing procedures and identified that two QA manual requirements were not being implemented. Specifically, QA Program Requirement, 3.1.10, stated that directives and procedures shall provide for analyzing trends of conditions adverse to quality. Once identified these trends were required to be considered significant conditions adverse to quality. The inspectors found that conditions adverse to quality were not defined in the KAP (Kewaunee corrective action program) procedures and that no procedure existed for trending. This finding does not directly affect a cornerstone. As a result, this issue was not evaluated with the Significance Determination Process and was not assigned a color. One example of a non-cited violation of 10 CFR Part 50, Appendix B, Criterion II, "Quality Assurance Program," was identified.
Inspection Report# : 2000019(pdf)
Significance: N/A Jun 22, 2000 Identified By: NRC Item Type: FIN Finding CONTROL ROOM OPERATIONS HUMAN PERFORMANCE ISSUES.
The inspectors interviewed operators to evaluate their awareness of degraded control room indications and alarms, and their ability to adequately take manual actions based on degraded alarm functions. The inspectors identified, during interviews, that there was a lack of awareness by operators of a degraded refueling water storage tank low-low level
 
3Q/2002 Inspection Findings - Kewaunee                                                                    Page 17 of 17 alarm which would be potentially confusing to operators and therefore increase the risk associated with initiating long term sump recirculation.
Inspection Report# : 2000008(pdf)
Last modified : December 02, 2002
 
4Q/2002 Inspection Findings - Kewaunee                                                                                                  Page 1 of 3 Kewaunee Initiating Events Mitigating Systems Significance:        Oct 01, 2002 Identified By: NRC Item Type: VIO Violation Failure to Provide Fixed Suppression System in Fire Area TU-95B During performance of follow-up activities in response to a USNRC inspection, the licensee identified that fire area TU-95B had been misclassified in that it should have been classified as required to meet the requirements of Section III.G.3 of 10 CFR Part 50, Appendix R. An apparent violation of 10 CFR Part 50, Appendix R, Section III.G.3 was identified for the failure to provide fire area TU-95B with a fixed fire suppression system. This issue has been preliminarily determined to have low to moderate safety significance (White). As a result of failing to have a fixed fire suppression system, there was a greater likelihood that a fire in fire area TU-95B would not be suppressed and redundant trains of cables and equipment required for safe shutdown could be damaged. The corresponding damage could require a shutdown of the plant from outside the control room, significantly increasing the complexity of manual actions required to achieve safe shutdown.
Inspection Report# : 2002006(pdf)
Significance:        Oct 01, 2002 Identified By: NRC Item Type: NCV NonCited Violation Inappropriate Placement of Diesel Generator Room Heat Detectors During performance of a triennial fire protection inspection, USNRC Region III staff identified that heat detectors used for activation of a diesel generator room carbon dioxide (CO2) system were not located and installed in accordance with the applicable National Fire Protection Association (NFPA) code. Specifically, no heat detectors were located at the ceiling level. The failure to appropriately locate and install heat detectors for actuation of the CO2 system is a violation of the Kewaunee Nuclear Power Plant operating license. The finding was greater than minor because it affected the protection against external factors (i.e., fire) attribute for mitigating systems. As a result of the inadequate heat detector placement, actuation of the carbon dioxide system in the diesel generator room could be delayed. The finding was of very low safety significance because the inspector was not able to identify a fire scenario in which safety significant cables would be damaged prior to actuation of the carbon dioxide system.
Inspection Report# : 2002006(pdf)
Significance:        Sep 30, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to Provide Appendix R Barriers Between Dedicated and Alternate Fire Zones A finding of very low risk significance was identified by the inspectors for the licensee's failure to provide fire barrier seals on auxiliary building Appendix R walls separating the Dedicated and Alternate fire zones.
Inspection Report# : 2002005(pdf)
Significance: TBD Sep 30, 2002 Identified By: NRC Item Type: URI Unresolved item Adequacy of Medical Examinations The inspectors identified an apparent violation of medical requirement regulations, 10 CFR 55.21, "Medical Examination," and 10 CFR 55.23, "Certification," in that the licensee's medical evaluations appeared to be inadequate in reference to ANSI/ANS 3.4-1983, and failed to adequately implement all the required medical testing. The finding is greater than minor, but is unresolved pending completion of the licensee's investigation into the medical issue, subsequent NRC review, and completion of a significance determination.
Inspection Report# : 2002005(pdf)
Significance: TBD Sep 30, 2002
 
4Q/2002 Inspection Findings - Kewaunee                                                                                                Page 2 of 3 Identified By: NRC Item Type: URI Unresolved item Adequacy of the Plant-Referenced Simulator to Conform with Simulator Requirements in 10 CFR 55.46 The inspectors identified an apparent violation of the simulator fidelity regulation, 10 CFR 55.46, "Simulation Facilities," in that the licensee's maintenance of simulator core modeling and simulator fidelity appeared to not comply with ANSI/ANS-3.5-1985. The finding is greater than minor, but is unresolved pending completion of the licensee's core modeling testing and investigation, subsequent NRC review of the core testing data, and completion of a significance determination for this issue. On September 19, 2002, the inspectors identified three issues concerning the potential failure to comply with 10 CFR 55.46, "Simulation Facilities." The first issue concerned the licensee's use of the simulator to meet experience requirements for applicants for initial operator and senior operator licenses in accordance with 10 CFR 55.46 (c)
(2)(I). The second issue concerned the adequacy of the licensee conducting periodic simulator performance testing throughout the life of the simulator. The third issue concerned the licensee's program for correcting simulator modeling and hardware discrepancies, including discrepancies identified from performance testing in accordance with 10 CFR 55.46 (d)(2).
Inspection Report# : 2002005(pdf)
Significance:        Jun 30, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to incorporate adquate acceptance criteria in service water procedure Green. The licensee failed to incorporate vendor information in a note contained in an operations procedure. The inaccurate note resulted in a service water pump being inappropriately declared operable. This finding was determined to be a Non-Cited Violation of Technical Specification 6.8.a, "Procedures".
Inspection Report# : 2002003(pdf)
Significance:        Jun 30, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to meet battery surveillance Technical Specification requirements Green. The licensee failed to measure and record safety-related battery cell electrolyte levels on a quarterly basis due to surveillance procedure inadequacies, which inhibited the licensee's ability to monitor and trend battery cell performance. Technical Specifications 4.6.b.2 and 4.6.b.3 required that the licensee measure and record battery cell electrolyte level on a quarterly basis. A Non-Cited Violation was identified.
Inspection Report# : 2002003(pdf)
Significance:        Jun 30, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain design control of component cooling water pumps Green. The licensee failed to adequately maintain design control of the component cooling water pumps, which resulted in the inability of a redundant train component cooling pump to provide cooling of safety-related loads due to the likely failure of the pump following a safety injection actuation. This finding was determined to be a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control.
Inspection Report# : 2002003(pdf)
Significance:        Feb 21, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform Thorough 10 CFR 50.59 Safety Evaluation The inspectors identified a Non-Cited Violation for failure to perform an adequate 10 CFR 50.59 safety evaluation associated with emergency operating procedure changes to address component cooling water pump dead-head operational concerns. The safety evaluation did not evaluate the potential for initiating a loss-of-coolant accident via the reactor coolant loop seals during conditions of a complete loss of component cooling water.
Inspection Report# : 2001017(pdf)
Barrier Integrity
 
4Q/2002 Inspection Findings - Kewaunee                                                                                                Page 3 of 3 Emergency Preparedness Significance:      Sep 30, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct an Instrument Deficiency A Non-Cited Violation of 10 CFR 50.54(q) was identified for the failure to correct a self-revealing deficiency that was initially identified in June 2002 and that was related to the emergency planning standards of 10 CFR 50.47(b). The deficiency concerned the meteorological monitoring system's instrumentation and the resulting erroneous 10-meter wind direction indications in the Control Room. Correct wind direction information would be required to ensure the capability to provide accurate dose assessments and protective action recommendations under accident conditions, as required by the Kewaunee Emergency Plan. The finding was determined to be of very low safety significance because the erroneous wind direction readings were identified prior to being needed for response to an actual emergency and alternate means were available to obtain accurate meteorological data. Therefore, the issue did not result in the failure to meet a planning standard.
Inspection Report# : 2002005(pdf)
Occupational Radiation Safety Public Radiation Safety Physical Protection Miscellaneous Last modified : March 25, 2003
 
1Q/2003 Inspection Findings - Kewaunee                                                                              Page 1 of 3 Kewaunee 1Q/2003 Plant Inspection Findings Initiating Events Mitigating Systems Significance:        Oct 01, 2002 Identified By: NRC Item Type: VIO Violation Failure to Provide Fixed Suppression System in Fire Area TU-95B During performance of follow-up activities in response to a USNRC inspection, the licensee identified that fire area TU-95B had been misclassified in that it should have been classified as required to meet the requirements of Section III.G.3 of 10 CFR Part 50, Appendix R. An apparent violation of 10 CFR Part 50, Appendix R, Section III.G.3 was identified for the failure to provide fire area TU-95B with a fixed fire suppression system. This issue has been preliminarily determined to have low to moderate safety significance (White). As a result of failing to have a fixed fire suppression system, there was a greater likelihood that a fire in fire area TU-95B would not be suppressed and redundant trains of cables and equipment required for safe shutdown could be damaged. The corresponding damage could require a shutdown of the plant from outside the control room, significantly increasing the complexity of manual actions required to achieve safe shutdown.
Inspection Report# : 2002006(pdf)
Significance:        Oct 01, 2002 Identified By: NRC Item Type: NCV NonCited Violation Inappropriate Placement of Diesel Generator Room Heat Detectors During performance of a triennial fire protection inspection, USNRC Region III staff identified that heat detectors used for activation of a diesel generator room carbon dioxide (CO2) system were not located and installed in accordance with the applicable National Fire Protection Association (NFPA) code. Specifically, no heat detectors were located at the ceiling level. The failure to appropriately locate and install heat detectors for actuation of the CO2 system is a violation of the Kewaunee Nuclear Power Plant operating license. The finding was greater than minor because it affected the protection against external factors (i.e., fire) attribute for mitigating systems. As a result of the inadequate heat detector placement, actuation of the carbon dioxide system in the diesel generator room could be delayed. The finding was of very low safety significance because the inspector was not able to identify a fire scenario in which safety significant cables would be damaged prior to actuation of the carbon dioxide system.
Inspection Report# : 2002006(pdf)
Significance:        Sep 30, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to Provide Appendix R Barriers Between Dedicated and Alternate Fire Zones file://C:\RROP\NRR\OVERSIGHT\ASSESS\KEWA\kewa_pim.html                                                              07/22/2003
 
1Q/2003 Inspection Findings - Kewaunee                                                                          Page 2 of 3 A finding of very low risk significance was identified by the inspectors for the licensee's failure to provide fire barrier seals on auxiliary building Appendix R walls separating the Dedicated and Alternate fire zones.
Inspection Report# : 2002005(pdf)
Significance:        Jun 30, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to incorporate adquate acceptance criteria in service water procedure Green. The licensee failed to incorporate vendor information in a note contained in an operations procedure. The inaccurate note resulted in a service water pump being inappropriately declared operable. This finding was determined to be a Non-Cited Violation of Technical Specification 6.8.a, "Procedures".
Inspection Report# : 2002003(pdf)
Significance:        Jun 30, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to meet battery surveillance Technical Specification requirements Green. The licensee failed to measure and record safety-related battery cell electrolyte levels on a quarterly basis due to surveillance procedure inadequacies, which inhibited the licensee's ability to monitor and trend battery cell performance. Technical Specifications 4.6.b.2 and 4.6.b.3 required that the licensee measure and record battery cell electrolyte level on a quarterly basis. A Non-Cited Violation was identified.
Inspection Report# : 2002003(pdf)
Significance:        Jun 30, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain design control of component cooling water pumps Green. The licensee failed to adequately maintain design control of the component cooling water pumps, which resulted in the inability of a redundant train component cooling pump to provide cooling of safety-related loads due to the likely failure of the pump following a safety injection actuation. This finding was determined to be a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control.
Inspection Report# : 2002003(pdf)
Barrier Integrity Significance:        Mar 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to Log Axial Flux Difference in Accordance with Technical Specifications The inspectors identified a finding of very low risk significance for the licensee's failure to monitor and log axial flux difference after disabling the power range axial flux monitor and computer alarm. The finding was of greater than minor risk significance because the operators failure to log and assess axial flux difference with the alarm disabled as required by Technical Specifications inhibited the operators' ability to trend changing core flux conditions. This failure file://C:\RROP\NRR\OVERSIGHT\ASSESS\KEWA\kewa_pim.html                                                          07/22/2003
 
1Q/2003 Inspection Findings - Kewaunee                                                                          Page 3 of 3 to log and assess axial flux difference could affect fuel cladding performance which is an attribute of the Barrier Integrity Cornerstone. The finding was of very low risk significance because although the finding impacted the Barrier Integrity Cornerstone, it affected the fuel barrier and not the reactor coolant system barrier and no actual abnormal axial flux difference existed during the time that the axial flux monitor alarm was disabled. The finding also affected the cross-cutting area of Human Performance because during the course of establishing a fixed signal in the Process Computer, operators were conducting activities beyond the bounds of approved procedural guidance. This finding was determined to be a Non-Cited Violation of Technical Specification 3.10.b.13.
Inspection Report# : 2003002(pdf)
Emergency Preparedness Significance:        Sep 30, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct an Instrument Deficiency A Non-Cited Violation of 10 CFR 50.54(q) was identified for the failure to correct a self-revealing deficiency that was initially identified in June 2002 and that was related to the emergency planning standards of 10 CFR 50.47(b). The deficiency concerned the meteorological monitoring system's instrumentation and the resulting erroneous 10-meter wind direction indications in the Control Room. Correct wind direction information would be required to ensure the capability to provide accurate dose assessments and protective action recommendations under accident conditions, as required by the Kewaunee Emergency Plan. The finding was determined to be of very low safety significance because the erroneous wind direction readings were identified prior to being needed for response to an actual emergency and alternate means were available to obtain accurate meteorological data. Therefore, the issue did not result in the failure to meet a planning standard.
Inspection Report# : 2002005(pdf)
Occupational Radiation Safety Public Radiation Safety Physical Protection Miscellaneous Last modified : May 30, 2003 file://C:\RROP\NRR\OVERSIGHT\ASSESS\KEWA\kewa_pim.html                                                          07/22/2003
 
2Q/2003 Inspection Findings - Kewaunee                                                                          Page 1 of 4 Kewaunee 2Q/2003 Plant Inspection Findings Initiating Events Significance:      Jun 30, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to Ensure Plant Conditions Appropriate for Tagout Results in Loss of Reactor Coolant System Inventory.
A self-revealed, non-cited violation of 10 CFR 50, Appendix B, Criterion V, was identified for the licensee's failure to properly sequence a tagout in accordance with the licensee's tagout procedure. This resulted in an approximate 100-gallon loss of inventory from the reactor coolant system. A contributing cause of this finding was related to the cross-cutting area of Human Performance. This finding is greater than minor because it affected the Initiating Events 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. The finding is of very low risk significance because none of the checklist attributes of Inspection Manual Chapter 0609, "Shutdown Operations Significance Determination Process," Appendix G, were affected.
Inspection Report# : 2003004(pdf)
Significance:      Jun 30, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to Provide Appropriate Instructions in Refueling Procedure Results in Reactor Vessel Level Indication Perturbation A self-revealed, non-cited violation of 10 CFR 50, Appendix B, Criterion V, was identified for the licensee's failure to ensure that the procedure governing refueling operations and reactor head disassembly had appropriate instructions or cautions to ensure that the reactor head vent remained vented to containment atmosphere. This resulted in the reactor head vent not being vented and affecting the operation of the refueling cavity water level instrument which operators were using to control reactor vessel water level. This finding is greater than minor because it is a configuration control issue which affected the Initiating Events 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. The finding is of very low risk significance because none of the checklist attributes of Inspection Manual Chapter 0609, "Shutdown Operations Significance Determination Process," Appendix G, were affected.
Inspection Report# : 2003004(pdf)
Mitigating Systems Significance:      Jun 30, 2003 Identified By: Self Disclosing file://C:\RROP\NRR\OVERSIGHT\ASSESS\KEWA\kewa_pim.html                                                          10/08/2003
 
2Q/2003 Inspection Findings - Kewaunee                                                                              Page 2 of 4 Item Type: NCV NonCited Violation Failure to Ensure Material of Installed Pipe Plug in RHR System is in Accordance with Design Requirements A self-revealed non-cited violation 10 CFR 50, Appendix B, Criterion V, was identified for the licensee's failure to ensure that the residual heat removal pump recirculation piping material was in accordance with a facility drawing and engineering specifications. This resulted in the corrosion of three pipe plugs, one of which was corroded to the point of leaking. The pipe plugs were installed on each residual heat removal's recirculation pipe pressure breakdown orifice.
The three pipe plugs were made of carbon steel while the residual heat removal system piping, which contained borated water, was required to be made of stainless steel. This finding was greater than minor because it affected the Mitigating System Cornerstone objective of equipment reliability and availability, in that the failure to ensure that the residual heat removal piping materials are in accordance with plant engineering specifications and drawings could result in system leakage significant enough to require taking the system out-of-service. The finding is of very low risk significance because this finding was not a design or qualification deficiency which resulted in a loss of function per Generic Letter 91-18.
Inspection Report# : 2003004(pdf)
Significance:        Oct 01, 2002 Identified By: NRC Item Type: NCV NonCited Violation Inappropriate Placement of Diesel Generator Room Heat Detectors During performance of a triennial fire protection inspection, USNRC Region III staff identified that heat detectors used for activation of a diesel generator room carbon dioxide (CO2) system were not located and installed in accordance with the applicable National Fire Protection Association (NFPA) code. Specifically, no heat detectors were located at the ceiling level. The failure to appropriately locate and install heat detectors for actuation of the CO2 system is a violation of the Kewaunee Nuclear Power Plant operating license. The finding was greater than minor because it affected the protection against external factors (i.e., fire) attribute for mitigating systems. As a result of the inadequate heat detector placement, actuation of the carbon dioxide system in the diesel generator room could be delayed. The finding was of very low safety significance because the inspector was not able to identify a fire scenario in which safety significant cables would be damaged prior to actuation of the carbon dioxide system.
Inspection Report# : 2002006(pdf)
Significance:        Sep 30, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to Provide Appendix R Barriers Between Dedicated and Alternate Fire Zones A finding of very low risk significance was identified by the inspectors for the licensee's failure to provide fire barrier seals on auxiliary building Appendix R walls separating the Dedicated and Alternate fire zones.
Inspection Report# : 2002005(pdf)
Significance:        Apr 04, 2002 Identified By: NRC Item Type: VIO Violation Failure to Provide Fixed Suppression System in Fire Area TU-95B During performance of follow-up activities in response to a USNRC inspection, the licensee identified that fire area TU-95B had been misclassified in that it should have been classified as required to meet the requirements of Section III.G.3 of 10 CFR Part 50, Appendix R. An apparent violation of 10 CFR Part 50, Appendix R, Section III.G.3 was identified for the failure to provide fire area TU-95B with a fixed fire suppression system. This issue has been preliminarily determined to have low to moderate safety significance (White). As a result of failing to have a fixed fire file://C:\RROP\NRR\OVERSIGHT\ASSESS\KEWA\kewa_pim.html                                                              10/08/2003
 
2Q/2003 Inspection Findings - Kewaunee                                                                          Page 3 of 4 suppression system, there was a greater likelihood that a fire in fire area TU-95B would not be suppressed and redundant trains of cables and equipment required for safe shutdown could be damaged. The corresponding damage could require a shutdown of the plant from outside the control room, significantly increasing the complexity of manual actions required to achieve safe shutdown.
Inspection Report# : 2002006(pdf)
Barrier Integrity Significance:        Mar 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to Log Axial Flux Difference in Accordance with Technical Specifications The inspectors identified a finding of very low risk significance for the licensee's failure to monitor and log axial flux difference after disabling the power range axial flux monitor and computer alarm. The finding was of greater than minor risk significance because the operators failure to log and assess axial flux difference with the alarm disabled as required by Technical Specifications inhibited the operators' ability to trend changing core flux conditions. This failure to log and assess axial flux difference could affect fuel cladding performance which is an attribute of the Barrier Integrity Cornerstone. The finding was of very low risk significance because although the finding impacted the Barrier Integrity Cornerstone, it affected the fuel barrier and not the reactor coolant system barrier and no actual abnormal axial flux difference existed during the time that the axial flux monitor alarm was disabled. The finding also affected the cross-cutting area of Human Performance because during the course of establishing a fixed signal in the Process Computer, operators were conducting activities beyond the bounds of approved procedural guidance. This finding was determined to be a Non-Cited Violation of Technical Specification 3.10.b.13.
Inspection Report# : 2003002(pdf)
Emergency Preparedness Significance:        Sep 30, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct an Instrument Deficiency A Non-Cited Violation of 10 CFR 50.54(q) was identified for the failure to correct a self-revealing deficiency that was initially identified in June 2002 and that was related to the emergency planning standards of 10 CFR 50.47(b). The deficiency concerned the meteorological monitoring system's instrumentation and the resulting erroneous 10-meter wind direction indications in the Control Room. Correct wind direction information would be required to ensure the capability to provide accurate dose assessments and protective action recommendations under accident conditions, as required by the Kewaunee Emergency Plan. The finding was determined to be of very low safety significance because the erroneous wind direction readings were identified prior to being needed for response to an actual emergency and alternate means were available to obtain accurate meteorological data. Therefore, the issue did not result in the failure to meet a planning standard.
Inspection Report# : 2002005(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\KEWA\kewa_pim.html                                                          10/08/2003
 
2Q/2003 Inspection Findings - Kewaunee                                                                          Page 4 of 4 Occupational Radiation Safety Public Radiation Safety Physical Protection Significance:      Jun 30, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to Report a Significant Fitness-for-Duty Event in a Timely Manner.
A finding of very low safety significance was identified by the inspectors for a violation of 10 CFR Part 26 Fitness-for-Duty (FFD) reporting requirements. The licensee failed to notify the NRC Operation Center within 24 hours of discovery of an illegal drug found within the licensee's protected area. The licensee failed to report the event because they did not realize this type of event was required to be reported. The finding was determined to be of very low significance because it was a vulnerability in the licensee's Safeguards plan, was not a malevolent act, and similar findings had not occurred in the last four calendar quarters. The finding was determined to be more than minor because illegal drugs located within a licensee's protected area are required to be reported to the NRC in accordance with 10 CFR 26.73(a) requirements.
Inspection Report# : 2003004(pdf)
Miscellaneous Last modified : September 04, 2003 file://C:\RROP\NRR\OVERSIGHT\ASSESS\KEWA\kewa_pim.html                                                          10/08/2003
 
3Q/2003 Inspection Findings - Kewaunee                                                                          Page 1 of 5 Kewaunee 3Q/2003 Plant Inspection Findings Initiating Events Significance:      Jun 30, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to Ensure Plant Conditions Appropriate for Tagout Results in Loss of Reactor Coolant System Inventory.
A self-revealed, non-cited violation of 10 CFR 50, Appendix B, Criterion V, was identified for the licensee's failure to properly sequence a tagout in accordance with the licensee's tagout procedure. This resulted in an approximate 100-gallon loss of inventory from the reactor coolant system. A contributing cause of this finding was related to the cross-cutting area of Human Performance.
This finding is greater than minor because it affected the Initiating Events 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. The finding is of very low risk significance because none of the checklist attributes of Inspection Manual Chapter 0609, "Shutdown Operations Significance Determination Process," Appendix G, were affected.
Inspection Report# : 2003004(pdf)
Significance:      Jun 30, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to Provide Appropriate Instructions in Refueling Procedure Results in Reactor Vessel Level Indication Perturbation A self-revealed, non-cited violation of 10 CFR 50, Appendix B, Criterion V, was identified for the licensee's failure to ensure that the procedure governing refueling operations and reactor head disassembly had appropriate instructions or cautions to ensure that the reactor head vent remained vented to containment atmosphere. This resulted in the reactor head vent not being vented and affecting the operation of the refueling cavity water level instrument which operators were using to control reactor vessel water level.
This finding is greater than minor because it is a configuration control issue which affected the Initiating Events 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. The finding is of very low risk significance because none of the checklist attributes of Inspection Manual Chapter 0609, "Shutdown Operations Significance Determination Process," Appendix G, were affected.
Inspection Report# : 2003004(pdf)
Significance:      Dec 19, 2002 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Adequate Separation of Safety-Related Circuits file://C:\RROP\NRR\OVERSIGHT\ASSESS\KEWA\kewa_pim.html                                                          01/12/2004
 
3Q/2003 Inspection Findings - Kewaunee                                                                            Page 2 of 5 A finding of very low safety significance associated with a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III Design Control was identified that pertained to improper application and use of a common balance-of-plant power supply to feed two redundant safety related circuits, which was not in accordance with the plant engineering specification procedure, the Updated Safety Analysis Report and the applicable Electrical and Electronics Engineers Standards.
This finding was more than minor because the lack of an adequate design for redundant safety related circuits could result in degradation of the component cooling water electrical system and if left uncorrected, could have the potential to upset plant stability, challenge critical safety functions during shutdown as well as power operations, and could potentially affect the reliability and capability of the component cooling water system to respond to initiating events.
This finding was of very low safety significance because it does not represent an actual loss of the component cooling water system's safety function.
Inspection Report# : 2002007(pdf)
Mitigating Systems Significance:      Sep 30, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to Prescribe Instructions or Procedures Appropriate to the Circumstances for the Seismic Storage of Equipment Near the 'A' Auxiliary Feedwater Piping The inspectors identified a Green finding associated with a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures and Drawings," for the failure to prescribe instructions or procedures appropriate to the circumstances for the seismic control of equipment stored near the vicinity of the A' Auxiliary Feedwater (AFW) piping to the A' Steam Generator, an activity affecting quality. The inspectors identified during plant walkdowns that following the 2003 Refueling Outage, portable plant equipment, including two portable 2.5-ton cranes, were stored in close proximity to the AFW piping, without the use of seismic restraints.
Inspection Report# : 2003006(pdf)
Significance:      Sep 30, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to Prescribe Instructions or Procedures Appropriate to the Circumstances for the Installation of the Refueling Cavity Drain Standpipe Following Refueling Activities A Green finding associated with a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures and Drawings," was self-revealed when the licensee, in preparing and verifying the response to NRC Bulletin 2003-01, "Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized Water Reactors," dated June 9, 2003, determined that the containment refueling cavity standpipe had not been installed after the Spring 2003 Refueling Outage. A procedure revision, issued prior to the 2003 Outage, had removed prescribed instructions to install the refueling cavity drain standpipe following reactor vessel refueling activities. The inspectors also concluded that this finding had, as a primary cause, a human performance deficiency.
Inspection Report# : 2003006(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\KEWA\kewa_pim.html                                                            01/12/2004
 
3Q/2003 Inspection Findings - Kewaunee                                                                              Page 3 of 5 Significance:      Jun 30, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to Ensure Material of Installed Pipe Plug in RHR System is in Accordance with Design Requirements A self-revealed non-cited violation 10 CFR 50, Appendix B, Criterion V, was identified for the licensee's failure to ensure that the residual heat removal pump recirculation piping material was in accordance with a facility drawing and engineering specifications. This resulted in the corrosion of three pipe plugs, one of which was corroded to the point of leaking. The pipe plugs were installed on each residual heat removal's recirculation pipe pressure breakdown orifice.
The three pipe plugs were made of carbon steel while the residual heat removal system piping, which contained borated water, was required to be made of stainless steel.
This finding was greater than minor because it affected the Mitigating System Cornerstone objective of equipment reliability and availability, in that the failure to ensure that the residual heat removal piping materials are in accordance with plant engineering specifications and drawings could result in system leakage significant enough to require taking the system out-of-service. The finding is of very low risk significance because this finding was not a design or qualification deficiency which resulted in a loss of function per Generic Letter 91-18.
Inspection Report# : 2003004(pdf)
Significance:      Dec 19, 2002 Identified By: NRC Item Type: NCV NonCited Violation Design Basis Calculations Contained Errors or Did Not Exist A finding of very low safety significance associated with a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III Design Control was identified that related to the control and quality of design basis engineering calculations. Specifically, a number of concerns were identified related to the indexing and control of existing calculations, the lack of available calculations to support some aspects of the current design basis, and errors in existing calculations. As a result of these issues, the current design basis calculations, as well as the existing calculation control processes, may not be adequate to ensure that the design basis will continue to be maintained. Although none of the specific deficiencies identified during the inspection resulted in immediate operability concerns, it was concluded that the component cooling water system design basis was not being adequately controlled by the existing calculations.
This finding was more than minor based on the potential that the lack of adequate control and quality of design basis calculations could result in the ability of the component cooling water system to perform its safety functions to be degraded. Design basis calculations were routinely used in support of design changes, operating procedures, test acceptance criteria, and operability determinations. This finding was of very low safety significance because it did not represent an actual loss of the component cooling water system's safety function.
Inspection Report# : 2002007(pdf)
Significance:      Oct 01, 2002 Identified By: NRC Item Type: NCV NonCited Violation Inappropriate Placement of Diesel Generator Room Heat Detectors During performance of a triennial fire protection inspection, USNRC Region III staff identified that heat detectors used for activation of a diesel generator room carbon dioxide (CO2) system were not located and installed in accordance with the applicable National Fire Protection Association (NFPA) code. Specifically, no heat detectors were located at the ceiling level. The failure to appropriately locate and install heat detectors for actuation of the CO2 system is a violation of the Kewaunee Nuclear Power Plant operating license.
file://C:\RROP\NRR\OVERSIGHT\ASSESS\KEWA\kewa_pim.html                                                              01/12/2004
 
3Q/2003 Inspection Findings - Kewaunee                                                                            Page 4 of 5 The finding was greater than minor because it affected the protection against external factors (i.e., fire) attribute for mitigating systems. As a result of the inadequate heat detector placement, actuation of the carbon dioxide system in the diesel generator room could be delayed. The finding was of very low safety significance because the inspector was not able to identify a fire scenario in which safety significant cables would be damaged prior to actuation of the carbon dioxide system.
Inspection Report# : 2002006(pdf)
Significance:        Apr 04, 2002 Identified By: NRC Item Type: VIO Violation Failure to Provide Fixed Suppression System in Fire Area TU-95B During performance of follow-up activities in response to a USNRC inspection, the licensee identified that fire area TU-95B had been misclassified in that it should have been classified as required to meet the requirements of Section III.G.3 of 10 CFR Part 50, Appendix R. An apparent violation of 10 CFR Part 50, Appendix R, Section III.G.3 was identified for the failure to provide fire area TU-95B with a fixed fire suppression system.
This issue has been preliminarily determined to have low to moderate safety significance (White). As a result of failing to have a fixed fire suppression system, there was a greater likelihood that a fire in fire area TU-95B would not be suppressed and redundant trains of cables and equipment required for safe shutdown could be damaged. The corresponding damage could require a shutdown of the plant from outside the control room, significantly increasing the complexity of manual actions required to achieve safe shutdown.
Due to the licensee failing to conduct a timely root cause evaluation a develop adequate corrective actions, this finding is being held open greater than four quarters until the licensee's root cause evaluation is complete and a supplemental inspection is conducted.
Inspection Report# : 2002006(pdf)
Barrier Integrity Significance:        Mar 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to Log Axial Flux Difference in Accordance with Technical Specifications The inspectors identified a finding of very low risk significance for the licensee's failure to monitor and log axial flux difference after disabling the power range axial flux monitor and computer alarm.
The finding was of greater than minor risk significance because the operators failure to log and assess axial flux difference with the alarm disabled as required by Technical Specifications inhibited the operators' ability to trend changing core flux conditions. This failure to log and assess axial flux difference could affect fuel cladding performance which is an attribute of the Barrier Integrity Cornerstone. The finding was of very low risk significance because although the finding impacted the Barrier Integrity Cornerstone, it affected the fuel barrier and not the reactor coolant system barrier and no actual abnormal axial flux difference existed during the time that the axial flux monitor alarm was disabled. The finding also affected the cross-cutting area of Human Performance because during the course of establishing a fixed signal in the Process Computer, operators were conducting activities beyond the bounds of approved procedural guidance. This finding was determined to be a Non-Cited Violation of Technical Specification file://C:\RROP\NRR\OVERSIGHT\ASSESS\KEWA\kewa_pim.html                                                            01/12/2004
 
3Q/2003 Inspection Findings - Kewaunee                                                                          Page 5 of 5 3.10.b.13.
Inspection Report# : 2003002(pdf)
Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Significance:      Jun 30, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to Report a Significant Fitness-for-Duty Event in a Timely Manner.
A finding of very low safety significance was identified by the inspectors for a violation of 10 CFR Part 26 Fitness-for-Duty (FFD) reporting requirements. The licensee failed to notify the NRC Operation Center within 24 hours of discovery of an illegal drug found within the licensee's protected area. The licensee failed to report the event because they did not realize this type of event was required to be reported.
The finding was determined to be of very low significance because it was a vulnerability in the licensee's Safeguards plan, was not a malevolent act, and similar findings had not occurred in the last four calendar quarters. The finding was determined to be more than minor because illegal drugs located within a licensee's protected area are required to be reported to the NRC in accordance with 10 CFR 26.73(a) requirements.
Inspection Report# : 2003004(pdf)
Miscellaneous Last modified : December 01, 2003 file://C:\RROP\NRR\OVERSIGHT\ASSESS\KEWA\kewa_pim.html                                                          01/12/2004
 
4Q/2003 Inspection Findings - Kewaunee                                                                          Page 1 of 9 Kewaunee 4Q/2003 Plant Inspection Findings Initiating Events Significance:      Jun 30, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to Ensure Plant Conditions Appropriate for Tagout Results in Loss of Reactor Coolant System Inventory.
A self-revealed, non-cited violation of 10 CFR 50, Appendix B, Criterion V, was identified for the licensee's failure to properly sequence a tagout in accordance with the licensee's tagout procedure. This resulted in an approximate 100-gallon loss of inventory from the reactor coolant system. A contributing cause of this finding was related to the cross-cutting area of Human Performance.
This finding is greater than minor because it affected the Initiating Events 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. The finding is of very low risk significance because none of the checklist attributes of Inspection Manual Chapter 0609, "Shutdown Operations Significance Determination Process," Appendix G, were affected.
Inspection Report# : 2003004(pdf)
Significance:      Jun 30, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to Provide Appropriate Instructions in Refueling Procedure Results in Reactor Vessel Level Indication Perturbation A self-revealed, non-cited violation of 10 CFR 50, Appendix B, Criterion V, was identified for the licensee's failure to ensure that the procedure governing refueling operations and reactor head disassembly had appropriate instructions or cautions to ensure that the reactor head vent remained vented to containment atmosphere. This resulted in the reactor head vent not being vented and affecting the operation of the refueling cavity water level instrument which operators were using to control reactor vessel water level.
This finding is greater than minor because it is a configuration control issue which affected the Initiating Events 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. The finding is of very low risk significance because none of the checklist attributes of Inspection Manual Chapter 0609, "Shutdown Operations Significance Determination Process," Appendix G, were affected.
Inspection Report# : 2003004(pdf)
Mitigating Systems Significance: SL-IV Dec 31, 2003 file://C:\RROP\NRR\OVERSIGHT\ASSESS\KEWA\kewa_pim.html                                                          04/22/2004
 
4Q/2003 Inspection Findings - Kewaunee                                                                          Page 2 of 9 Identified By: NRC Item Type: NCV NonCited Violation Non-cited violation of 10 CFR 50.59, for the failure to perform a written evaluation, as required, for a modification to the component cooling water system The inspectors identified a finding of very low safety significance associated with a Non-Cited Violation of 10 CFR 50.59(d)(1) for the licensee's failure to perform a safety evaluation for changes made to the facility. Specifically, the licensee screened out' of the 10 CFR 50.59 process a modification that included the addition of a minimum flow recirculation line to the component cooling water pumps. This modification further cross-connected the suction and discharge piping of both component cooling water pump trains. Subsequently, the inspectors identified and the licensee concurred that a safety evaluation was required for this modification.
Because the Significance Determination Process is not designed to assess the significance of violations that potentially impact or impeded the regulatory process, this issue was dispositioned using the traditional enforcement process in accordance with Section IV of the NRC Enforcement Policy. However, the results of this violation were assessed using the Significance Determination Process. In this case, the licensee failed to perform a safety evaluation in accordance with 10 CFR 50.59 and had placed the new system in service for testing prior to the completion of the required safety evaluation.
The inspectors considered this issue to be of more than minor significance because, if left uncorrected, the issue could become a more significant safety concern. Specifically, the inspectors noted that the licensee's processes for permanent modifications failed to identify this issue at several review levels. The inspectors determined that the issue was of very low significance because the new system was placed in service for a short period of time for testing prior to the completion of the required safety evaluation. In addition, the final safety evaluation completed by the licensee in October 2003 determined that the modification did not require prior NRC approval. The inspectors determined this finding was a Severity Level IV Non-Cited Violation of 10 CFR 50.59. The inspectors also determined that the finding had, as a primary cause, a human performance deficiency which affected the cross-cutting are of Human Performance.
Inspection Report# : 2003008(pdf)
Significance:      Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation 10 CFR 50, Appendix B, Criterion III, "Design Control," the failure to provide for the checking the adequacy of design for temporary mod which changed the CCW system pressure boundary The inspectors identified a finding of very low safety significance associated with a Non-Cited Violation of 10 CFR Part 50 Appendix B, Criterion III, "Design Control," for the licensee's failure to provide for checking of the adequacy of the design in Temporary Change TCR 03-036, in that, the design review failed to confirm the structural integrity of the new pressure boundary established for the studding outlet. Consequently, the licensee performed non-destructive examinations and additional flaw and engineering analyses to confirm the adequacy of the new design.
The inspectors considered this issue of more than minor significance, because if left uncorrected, the issue could become a more significant safety concern. In addition, the inspectors concluded that the finding was greater than minor because the finding involved the design control attribute of the mitigating systems cornerstone and affected the mitigating systems objective of ensuring the capability of the component cooling water system in response to initiating events to prevent undesirable consequences. Specifically, the temporary design change relied on unsupported assumptions that could have impacted the structural integrity of the component cooling water suction line. The inspectors evaluated the finding using the Significance Determination Process Phase 1 screening and determined that the finding was a design or qualification deficiency confirmed not to result in loss of function per Generic Letter 91-18; therefore, the finding was determined to be of very low safety significance.
Inspection Report# : 2003008(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\KEWA\kewa_pim.html                                                          04/22/2004
 
4Q/2003 Inspection Findings - Kewaunee                                                                          Page 3 of 9 Significance:      Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation 10 CFR 50, Appendix B, Criterion XVI, "Corrective Actions," No appropriate immediate corrective actions for reliability issues associated with incorrect cranking cutout relay installed in the EDGs The inspectors identified a finding of very low safety significance associated with a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Actions," for the licensee's failure to take adequate corrective actions in response to the installation of non-conforming cranking cutout relays which prevented energizing of the diesel generator engine start relay. The licensee's corrective actions for this condition adverse to quality addressed routine surveillance procedures, but did not consider the licensee's Emergency Operating Procedures to ensure the Emergency Diesel Generators would remain operable following Diesel Generator Shutdowns as directed by those procedures.
The inspectors considered this issue of more than minor significance, because if left uncorrected, the issue could become a more significant safety concern. In addition, the inspectors concluded that the finding was greater than minor because the finding involved the design control attribute of the mitigating systems cornerstone and affected the mitigating systems objective of ensuring the capability of the diesel generators in response to initiating events to prevent undesirable consequences. Specifically, in part, the licensee's corrective actions included revisions to normal operating procedures to verify continuity across the relay contacts following shutdown of the emergency diesel generators; however, the licensee did not similarly revise its Emergency Operating Procedures to verify continuity across the cranking cutout relay contacts following shutdown of the emergency diesel generators. The inspectors evaluated the finding using the Significance Determination Process Phase 1 screening and determined that the finding was a design or qualification deficiency confirmed not to result in loss of function per Generic Letter 91-18; therefore, the finding was determined to be of very low safety significance.
Inspection Report# : 2003008(pdf)
Significance:      Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation 10 CFR 50, Appendix B, Criterion III, "Design Control," failure to install the appropriate cranking cutout relay in the EDG system in 1998; this resulted in failure of 'B' EDG to start in Feb., 2003 A finding of very low safety significance involving a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," was self-revealing when the "B" emergency diesel generator failed to start on February 26, 2003, during a daily Technical Specification-required test, in response to the "A" emergency diesel generator being out of service for regularly scheduled 18-month periodic maintenance. The generator failed to start due to a pair of electrically open contacts on a cranking cutout relay which prevented energizing of the engine start relay. The cranking cutout relay had been installed during a design change completed in 1998, and the performance ratings of the new relay did not match original design specifications.
The inspectors considered this issue of more than minor significance, because if left uncorrected, the issue could become a more significant safety concern. In addition, the inspectors concluded that the finding was greater than minor because the finding involved the design control attribute of the mitigating systems cornerstone and affected the mitigating systems objective of ensuring the capability of the diesel generators in response to initiating events to prevent undesirable consequences. Specifically, the temporary design change failed to consider inductive electrical loads across the relay contacts, for which the relays were not rated. The inspectors evaluated the finding using the Significance Determination Process Phase 1 screening and determined that the finding was a design or qualification deficiency confirmed not to result in loss of function per Generic Letter 91-18; therefore, the finding was determined to be of very low safety significance.
Inspection Report# : 2003008(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\KEWA\kewa_pim.html                                                          04/22/2004
 
4Q/2003 Inspection Findings - Kewaunee                                                                                Page 4 of 9 Significance:      Dec 12, 2003 Identified By: NRC Item Type: NCV NonCited Violation 10 CFR 50, Appendix B, Criterion XVI NCV for ineffective corrective actions taken to address the implementation of the Boric Acid Leakage Inspection and Tracking Program The team identified a finding of very low significance associated with a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," for the failure to assure that actions were promptly taken to correct deficiencies in the implementation of the boric acid leakage inspection and tracking program for boric acid residue on safety-related components, a condition adverse to quality. Since 2001, approximately 12 condition reports had been initiated concerning the adequacy of the implementation of the licensee's boric acid leakage inspection and tracking program. During the inspection, the team identified approximately 14 safety-related components with various degrees of boric acid, which the licensee had not identified and evaluated in accordance with the boric acid leakage inspection and tracking program.
The team concluded that the licensee's failure to correct previous issues associated with the implementation of the boric acid leak log on safety-related components was greater than minor because if left uncorrected, the issue could become a more significant safety concern. The team evaluated the finding utilizing Inspection Manual Chapter 0609, "Significance Determination Process," Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," Phase 1 screening and determined the finding was of very low significance.
Inspection Report# : 2003010(pdf)
Significance:      Oct 21, 2003 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PERFORM TWO REQUIRED MEDICAL TESTS IN ACCORDANCE WITH 10 CFR 55.21 AND 55.23.
Green. The inspector identified a Non-Cited Violation of 10 CFR 55.21, "Medical Examination," and 10 CFR 55.23, "Certification." The inspector identified that the facility licensee failed to conduct all the medical testing required by American Nuclear Standards Institute/American Nuclear Society (ANSI/ANS) 3.4-1983, "Medical Certification and Monitoring of Personnel Requiring Operator Licenses for Nuclear Power Plants," as committed to by the facility licensee. Specifically, the facility licensee was not testing its operators for nose sensitivity (i.e., ability to detect odor of products of combustion and of tracer or market gases) Section 5.4.2, "Nose," and neurological testing, (i.e., normal central and peripheral nervous system function), including tactile discrimination (Stereognosis) sufficient to distinguish among various shapes of control knobs and hadles by touch, Section 5.4.14, "Neurological."
Inspection Report# : 2003005(pdf)
Significance: SL-IV Oct 21, 2003 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PROVIDE ACCURATE INFORMATION TO THE NRC CONCERNING LICENSED OPERATOR MEDICAL REQUIREMENTS PER NRC FORM 396.
Severity Level IV. The inspector identified a Level IV Non-Cited Violation of 10 CFR 50.9, "Completeness and Accuracy of Information." The inspector identified that the facility licensee, between January 2, 2000, thorugh August 26, 2002, submitted to the NRC, NRC Forms 396 for 13 individuals applying for an initial operator's license and 18 licensed operators applying for renewal of their operator licenses, that were not accurate in all material respects.
Specifically, the NRC Forms 396 certified that each applicant and licensed operator met the medical requirements of ANSI/ANS 3.4-1983. In fact, all the applicants and licensed operators were not adequately examined for all medical tests as required to meet the minimum standards of ANSI/ANS 3.4-1983.
file://C:\RROP\NRR\OVERSIGHT\ASSESS\KEWA\kewa_pim.html                                                                04/22/2004
 
4Q/2003 Inspection Findings - Kewaunee                                                                          Page 5 of 9 Inspection Report# : 2003005(pdf)
Significance:        Oct 21, 2003 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO CONDUCT SIMULATOR PERFORMANCE TESTING THROUGHOUT THE LIFE OF THE SIMULATOR.
Green. The inspector identified a Non-Cited Violation of 10 CFR 55.46, "Simulation Facility." The inspector identified that the facility licensee failed to adequately conduct simulator performance testing throughout the life of the simulator.
In addition, the facility licensee failed to correct modeling and hardware discrepancies and discrepancies identified from scenario validation and from performance testing. In addition, the facility licensee was committed to follow ANSI/ANS 3.5-1985, "Nuclear Power Plant Simulators for Use in Operator Training," as the way they would meet 10 CFR 55.46. Specifically, the licensee failed to conduct performance testing, with regard to normal evolutions core performance tests for Cycle 25, the most recent core load in the actual reactor. The licensee could only provide Cycle 7 normal evolutions core performance tests. No core performance tests had ever been conducted for Cycles 8 through 25, a period of 17 cycles.
Inspection Report# : 2003005(pdf)
Significance: SL-IV Oct 21, 2003 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PROVIDE ACCURATE INFORMATION TO THE NRC CONCERNING ELIGIBILITY REQUIREMENTS FOR OPERATOR LICENSE APPLICATION PER NRC FORM 398.
Severity Level IV. The inspector identified a Level IV Non-Cited Violation of 10 CFR 50.9, "Completeness and Accuracy of Information." The inspector identified that on or about August 13, 2002, a senior facility licensee representative submitted to the NRC, NRC Forms 398 for three individuals, each applying for an initial operator's license, that were not accurate in all material respects. The facility licensee provided inaccurate information by certifying on the NRC Form 398 that the initial operator license applicaitons for three individuals had appropriately met the minimum training requirements for reactivity manipulations on the refrenced facility simulator in accordance with 10 CFR 55.31(a)(5) and 10 CFR 55.46(c)(2).
Inspection Report# : 2003005(pdf)
Significance:        Sep 30, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to Prescribe Instructions or Procedures Appropriate to the Circumstances for the Seismic Storage of Equipment Near the 'A' Auxiliary Feedwater Piping The inspectors identified a Green finding associated with a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures and Drawings," for the failure to prescribe instructions or procedures appropriate to the circumstances for the seismic control of equipment stored near the vicinity of the A' Auxiliary Feedwater (AFW) piping to the A' Steam Generator, an activity affecting quality. The inspectors identified during plant walkdowns that following the 2003 Refueling Outage, portable plant equipment, including two portable 2.5-ton cranes, were stored in close proximity to the AFW piping, without the use of seismic restraints.
Inspection Report# : 2003006(pdf) file://C:\RROP\NRR\OVERSIGHT\ASSESS\KEWA\kewa_pim.html                                                          04/22/2004
 
4Q/2003 Inspection Findings - Kewaunee                                                                              Page 6 of 9 Significance:        Sep 30, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to Prescribe Instructions or Procedures Appropriate to the Circumstances for the Installation of the Refueling Cavity Drain Standpipe Following Refueling Activities A Green finding associated with a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures and Drawings," was self-revealed when the licensee, in preparing and verifying the response to NRC Bulletin 2003-01, "Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized Water Reactors," dated June 9, 2003, determined that the containment refueling cavity standpipe had not been installed after the Spring 2003 Refueling Outage. A procedure revision, issued prior to the 2003 Outage, had removed prescribed instructions to install the refueling cavity drain standpipe following reactor vessel refueling activities. The inspectors also concluded that this finding had, as a primary cause, a human performance deficiency.
Inspection Report# : 2003006(pdf)
Significance:        Jun 30, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to Ensure Material of Installed Pipe Plug in RHR System is in Accordance with Design Requirements A self-revealed non-cited violation 10 CFR 50, Appendix B, Criterion V, was identified for the licensee's failure to ensure that the residual heat removal pump recirculation piping material was in accordance with a facility drawing and engineering specifications. This resulted in the corrosion of three pipe plugs, one of which was corroded to the point of leaking. The pipe plugs were installed on each residual heat removal's recirculation pipe pressure breakdown orifice.
The three pipe plugs were made of carbon steel while the residual heat removal system piping, which contained borated water, was required to be made of stainless steel.
This finding was greater than minor because it affected the Mitigating System Cornerstone objective of equipment reliability and availability, in that the failure to ensure that the residual heat removal piping materials are in accordance with plant engineering specifications and drawings could result in system leakage significant enough to require taking the system out-of-service. The finding is of very low risk significance because this finding was not a design or qualification deficiency which resulted in a loss of function per Generic Letter 91-18.
Inspection Report# : 2003004(pdf)
Significance:        Apr 04, 2002 Identified By: NRC Item Type: VIO Violation Failure to Provide Fixed Suppression System in Fire Area TU-95B During performance of follow-up activities in response to a USNRC inspection, the licensee identified that fire area TU-95B had been misclassified in that it should have been classified as required to meet the requirements of Section III.G.3 of 10 CFR Part 50, Appendix R. An apparent violation of 10 CFR Part 50, Appendix R, Section III.G.3 was identified for the failure to provide fire area TU-95B with a fixed fire suppression system.
This issue has been preliminarily determined to have low to moderate safety significance (White). As a result of failing to have a fixed fire suppression system, there was a greater likelihood that a fire in fire area TU-95B would not be suppressed and redundant trains of cables and equipment required for safe shutdown could be damaged. The corresponding damage could require a shutdown of the plant from outside the control room, significantly increasing the complexity of manual actions required to achieve safe shutdown.
file://C:\RROP\NRR\OVERSIGHT\ASSESS\KEWA\kewa_pim.html                                                              04/22/2004
 
4Q/2003 Inspection Findings - Kewaunee                                                                          Page 7 of 9 Due to the licensee failing to conduct a timely root cause evaluation a develop adequate corrective actions, this finding is being held open greater than four quarters until the licensee's root cause evaluation is complete and a supplemental inspection is conducted.
Inspection Report# : 2002006(pdf)
Inspection Report# : 2003003(pdf)
Barrier Integrity Significance:        Mar 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to Log Axial Flux Difference in Accordance with Technical Specifications The inspectors identified a finding of very low risk significance for the licensee's failure to monitor and log axial flux difference after disabling the power range axial flux monitor and computer alarm.
The finding was of greater than minor risk significance because the operators failure to log and assess axial flux difference with the alarm disabled as required by Technical Specifications inhibited the operators' ability to trend changing core flux conditions. This failure to log and assess axial flux difference could affect fuel cladding performance which is an attribute of the Barrier Integrity Cornerstone. The finding was of very low risk significance because although the finding impacted the Barrier Integrity Cornerstone, it affected the fuel barrier and not the reactor coolant system barrier and no actual abnormal axial flux difference existed during the time that the axial flux monitor alarm was disabled. The finding also affected the cross-cutting area of Human Performance because during the course of establishing a fixed signal in the Process Computer, operators were conducting activities beyond the bounds of approved procedural guidance. This finding was determined to be a Non-Cited Violation of Technical Specification 3.10.b.13.
Inspection Report# : 2003002(pdf)
Emergency Preparedness Significance:        Dec 12, 2003 Identified By: NRC Item Type: FIN Finding Failure to take timely corrective actions to prevent recurrence for a 2001 white finding associated with Emergency Response Organization Augmentation The team identified a Green finding for the failure to take timely corrective actions to prevent recurrence for a White Finding initially identified in September 2000, associated with Emergency Response Organization Augmentation.
While the team determined that corrective actions to date have been effective, as evidenced by only one augmentation drill failure since 2001, three of the eight corrective actions had not been completed.
The team determined that this issue was more than minor because if left uncorrected, the issue could become a more significant safety concern. In addition, the team concluded that the issue affected the emergency preparedness cornerstone performance attribute associated with the emergency response organization augmentation system and emergency response augmentation testing and the objective of implementing adequate measures to protect the health file://C:\RROP\NRR\OVERSIGHT\ASSESS\KEWA\kewa_pim.html                                                          04/22/2004
 
4Q/2003 Inspection Findings - Kewaunee                                                                          Page 8 of 9 and safety of the public in the event of a radiological emergency. The team evaluated the finding utilizing Inspection Manual Chapter 0609, Appendix B, "Emergency Preparedness Significance Determination Process," Section 5.0, "Corrective Actions," dated March 6, 2003, and determined the finding was of very low significance.
Inspection Report# : 2003010(pdf)
Significance:      Dec 12, 2003 Identified By: NRC Item Type: NCV NonCited Violation 10 CFR50.54 and emergency plan NCV for ineffective corrective actions in 2002 which resulted in the failure to make timely notifications for an actual unusual event in February 2003 The team identified a finding of very low significance associated with a Non-Cited Violation of 10 CFR Part 50.54(q) and the licensee's Emergency Plan for the failure to notify the state and local governmental agencies within 15 minutes after the declaration of an actual Unusual Event on February 26, 2003. The team concluded this failure was caused by the licensee's ineffective corrective actions for previously identified weaknesses and problems in the area of Emergency Preparedness.
The team determined that this issue was more than minor because this was an actual event implementation problem and affected the emergency preparedness cornerstone objective of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The team evaluated the finding utilizing Inspection Manual Chapter 0609, Appendix B, "Emergency Preparedness Significance Determination Process," Emergency Preparedness Significance Determination Process Sheet 2, dated March 6, 2003, and determined the finding was of very low significance.
Inspection Report# : 2003010(pdf)
Occupational Radiation Safety Public Radiation Safety Physical Protection Significance:      Jun 30, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to Report a Significant Fitness-for-Duty Event in a Timely Manner.
A finding of very low safety significance was identified by the inspectors for a violation of 10 CFR Part 26 Fitness-for-Duty (FFD) reporting requirements. The licensee failed to notify the NRC Operation Center within 24 hours of discovery of an illegal drug found within the licensee's protected area. The licensee failed to report the event because they did not realize this type of event was required to be reported.
The finding was determined to be of very low significance because it was a vulnerability in the licensee's Safeguards file://C:\RROP\NRR\OVERSIGHT\ASSESS\KEWA\kewa_pim.html                                                          04/22/2004
 
4Q/2003 Inspection Findings - Kewaunee                                                                        Page 9 of 9 plan, was not a malevolent act, and similar findings had not occurred in the last four calendar quarters. The finding was determined to be more than minor because illegal drugs located within a licensee's protected area are required to be reported to the NRC in accordance with 10 CFR 26.73(a) requirements.
Inspection Report# : 2003004(pdf)
Miscellaneous Last modified : March 02, 2004 file://C:\RROP\NRR\OVERSIGHT\ASSESS\KEWA\kewa_pim.html                                                        04/22/2004
 
1Q/2004 Inspection Findings - Kewaunee                                                                                                Page 1 of 6 Kewaunee 1Q/2004 Plant Inspection Findings Initiating Events Significance:        Jun 30, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to Ensure Plant Conditions Appropriate for Tagout Results in Loss of Reactor Coolant System Inventory.
A self-revealed, non-cited violation of 10 CFR 50, Appendix B, Criterion V, was identified for the licensee's failure to properly sequence a tagout in accordance with the licensee's tagout procedure. This resulted in an approximate 100-gallon loss of inventory from the reactor coolant system. A contributing cause of this finding was related to the cross-cutting area of Human Performance.
This finding is greater than minor because it affected the Initiating Events 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. The finding is of very low risk significance because none of the checklist attributes of Inspection Manual Chapter 0609, "Shutdown Operations Significance Determination Process," Appendix G, were affected.
Inspection Report# : 2003004(pdf)
Significance:        Jun 30, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to Provide Appropriate Instructions in Refueling Procedure Results in Reactor Vessel Level Indication Perturbation A self-revealed, non-cited violation of 10 CFR 50, Appendix B, Criterion V, was identified for the licensee's failure to ensure that the procedure governing refueling operations and reactor head disassembly had appropriate instructions or cautions to ensure that the reactor head vent remained vented to containment atmosphere. This resulted in the reactor head vent not being vented and affecting the operation of the refueling cavity water level instrument which operators were using to control reactor vessel water level.
This finding is greater than minor because it is a configuration control issue which affected the Initiating Events 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. The finding is of very low risk significance because none of the checklist attributes of Inspection Manual Chapter 0609, "Shutdown Operations Significance Determination Process," Appendix G, were affected.
Inspection Report# : 2003004(pdf)
Mitigating Systems Significance:        Mar 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO NOTIFY THE NRC OF A CHANGE IN OPERATOR STATUS IN ACCORDANCE WITH 10 CFR 50.74(c)
The inspector identified a violation of 10 CFR 50.74(c), "Notification of Change in Operator or Senior Operator Status." The inspector identified that the facility licensee failed to notify the NRC within 30 days after receiving a change in medical status of a licensed operator from the station's medical examiner. The change in medical status required conditioning of the operator's license by the NRC.
Inspection Report# : 2004002(pdf)
Significance:        Jan 28, 2004 Identified By: NRC Item Type: FIN Finding Failure to appropriately evaluate for potential bypqss flow on service water strainers The inspectors identified a finding of very low safety significance associated with the licensee's failure to appropriately evaluate for potential bypass flow on the service water pump discharge strainers by measuring a critical gap dimension at the bottom of the basket-to-housing 07/14/2004
 
1Q/2004 Inspection Findings - Kewaunee                                                                                                  Page 2 of 6 interface. This finding did not constitute a violation of NRC requirements because the strainers (aside from the pressure boundary) did not fulfill a safety-related function.
The inspectors determined that the finding was of more than minor significance because it would become a more significant safety concern if left uncorrected. Specifically, the failure to appropriately evaluate for potential bypass flow on the service water pump discharge strainers could reasonably result in debris fouling of service water cooled components and degraded or inoperable safety-related equipment. The inspectors concluded that this finding was a licensee performance deficiency of very low safety significance because it did not result in loss of safety function for a service water system train for greater than its Technical Specification allowed outage time. To address this issue, the licensee opened each strainer and measured the gap at the bottom of the basket-to-housing interface.
Inspection Report# : 2004003(pdf)
Significance: SL-IV Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation Non-cited violation of 10 CFR 50.59, for the failure to perform a written evaluation, as required, for a modification to the component cooling water system The inspectors identified a finding of very low safety significance associated with a Non-Cited Violation of 10 CFR 50.59(d)(1) for the licensee's failure to perform a safety evaluation for changes made to the facility. Specifically, the licensee screened out' of the 10 CFR 50.59 process a modification that included the addition of a minimum flow recirculation line to the component cooling water pumps. This modification further cross-connected the suction and discharge piping of both component cooling water pump trains. Subsequently, the inspectors identified and the licensee concurred that a safety evaluation was required for this modification.
Because the Significance Determination Process is not designed to assess the significance of violations that potentially impact or impeded the regulatory process, this issue was dispositioned using the traditional enforcement process in accordance with Section IV of the NRC Enforcement Policy. However, the results of this violation were assessed using the Significance Determination Process. In this case, the licensee failed to perform a safety evaluation in accordance with 10 CFR 50.59 and had placed the new system in service for testing prior to the completion of the required safety evaluation.
The inspectors considered this issue to be of more than minor significance because, if left uncorrected, the issue could become a more significant safety concern. Specifically, the inspectors noted that the licensee's processes for permanent modifications failed to identify this issue at several review levels. The inspectors determined that the issue was of very low significance because the new system was placed in service for a short period of time for testing prior to the completion of the required safety evaluation. In addition, the final safety evaluation completed by the licensee in October 2003 determined that the modification did not require prior NRC approval. The inspectors determined this finding was a Severity Level IV Non-Cited Violation of 10 CFR 50.59. The inspectors also determined that the finding had, as a primary cause, a human performance deficiency which affected the cross-cutting are of Human Performance.
Inspection Report# : 2003008(pdf)
Significance:        Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation 10 CFR 50, Appendix B, Criterion III, "Design Control," the failure to provide for the checking the adequacy of design for temporary mod which changed the CCW system pressure boundary The inspectors identified a finding of very low safety significance associated with a Non-Cited Violation of 10 CFR Part 50 Appendix B, Criterion III, "Design Control," for the licensee's failure to provide for checking of the adequacy of the design in Temporary Change TCR 03-036, in that, the design review failed to confirm the structural integrity of the new pressure boundary established for the studding outlet.
Consequently, the licensee performed non-destructive examinations and additional flaw and engineering analyses to confirm the adequacy of the new design.
The inspectors considered this issue of more than minor significance, because if left uncorrected, the issue could become a more significant safety concern. In addition, the inspectors concluded that the finding was greater than minor because the finding involved the design control attribute of the mitigating systems cornerstone and affected the mitigating systems objective of ensuring the capability of the component cooling water system in response to initiating events to prevent undesirable consequences. Specifically, the temporary design change relied on unsupported assumptions that could have impacted the structural integrity of the component cooling water suction line. The inspectors evaluated the finding using the Significance Determination Process Phase 1 screening and determined that the finding was a design or qualification deficiency confirmed not to result in loss of function per Generic Letter 91-18; therefore, the finding was determined to be of very low safety significance.
Inspection Report# : 2003008(pdf)
Significance:        Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation 10 CFR 50, Appendix B, Criterion XVI, "Corrective Actions," No appropriate immediate corrective actions for reliability issues associated with incorrect cranking cutout relay installed in the EDGs 07/14/2004
 
1Q/2004 Inspection Findings - Kewaunee                                                                                                Page 3 of 6 The inspectors identified a finding of very low safety significance associated with a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Actions," for the licensee's failure to take adequate corrective actions in response to the installation of non-conforming cranking cutout relays which prevented energizing of the diesel generator engine start relay. The licensee's corrective actions for this condition adverse to quality addressed routine surveillance procedures, but did not consider the licensee's Emergency Operating Procedures to ensure the Emergency Diesel Generators would remain operable following Diesel Generator Shutdowns as directed by those procedures.
The inspectors considered this issue of more than minor significance, because if left uncorrected, the issue could become a more significant safety concern. In addition, the inspectors concluded that the finding was greater than minor because the finding involved the design control attribute of the mitigating systems cornerstone and affected the mitigating systems objective of ensuring the capability of the diesel generators in response to initiating events to prevent undesirable consequences. Specifically, in part, the licensee's corrective actions included revisions to normal operating procedures to verify continuity across the relay contacts following shutdown of the emergency diesel generators; however, the licensee did not similarly revise its Emergency Operating Procedures to verify continuity across the cranking cutout relay contacts following shutdown of the emergency diesel generators. The inspectors evaluated the finding using the Significance Determination Process Phase 1 screening and determined that the finding was a design or qualification deficiency confirmed not to result in loss of function per Generic Letter 91-18; therefore, the finding was determined to be of very low safety significance.
Inspection Report# : 2003008(pdf)
Significance:        Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation 10 CFR 50, Appendix B, Criterion III, "Design Control," failure to install the appropriate cranking cutout relay in the EDG system in 1998; this resulted in failure of 'B' EDG to start in Feb., 2003 A finding of very low safety significance involving a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," was self-revealing when the "B" emergency diesel generator failed to start on February 26, 2003, during a daily Technical Specification-required test, in response to the "A" emergency diesel generator being out of service for regularly scheduled 18-month periodic maintenance. The generator failed to start due to a pair of electrically open contacts on a cranking cutout relay which prevented energizing of the engine start relay. The cranking cutout relay had been installed during a design change completed in 1998, and the performance ratings of the new relay did not match original design specifications.
The inspectors considered this issue of more than minor significance, because if left uncorrected, the issue could become a more significant safety concern. In addition, the inspectors concluded that the finding was greater than minor because the finding involved the design control attribute of the mitigating systems cornerstone and affected the mitigating systems objective of ensuring the capability of the diesel generators in response to initiating events to prevent undesirable consequences. Specifically, the temporary design change failed to consider inductive electrical loads across the relay contacts, for which the relays were not rated. The inspectors evaluated the finding using the Significance Determination Process Phase 1 screening and determined that the finding was a design or qualification deficiency confirmed not to result in loss of function per Generic Letter 91-18; therefore, the finding was determined to be of very low safety significance.
Inspection Report# : 2003008(pdf)
Significance:        Dec 12, 2003 Identified By: NRC Item Type: NCV NonCited Violation 10 CFR 50, Appendix B, Criterion XVI NCV for ineffective corrective actions taken to address the implementation of the Boric Acid Leakage Inspection and Tracking Program The team identified a finding of very low significance associated with a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," for the failure to assure that actions were promptly taken to correct deficiencies in the implementation of the boric acid leakage inspection and tracking program for boric acid residue on safety-related components, a condition adverse to quality. Since 2001, approximately 12 condition reports had been initiated concerning the adequacy of the implementation of the licensee's boric acid leakage inspection and tracking program. During the inspection, the team identified approximately 14 safety-related components with various degrees of boric acid, which the licensee had not identified and evaluated in accordance with the boric acid leakage inspection and tracking program.
The team concluded that the licensee's failure to correct previous issues associated with the implementation of the boric acid leak log on safety-related components was greater than minor because if left uncorrected, the issue could become a more significant safety concern. The team evaluated the finding utilizing Inspection Manual Chapter 0609, "Significance Determination Process," Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," Phase 1 screening and determined the finding was of very low significance.
Inspection Report# : 2003010(pdf)
Significance:        Oct 21, 2003 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PERFORM TWO REQUIRED MEDICAL TESTS IN ACCORDANCE WITH 10 CFR 55.21 AND 55.23.
07/14/2004
 
1Q/2004 Inspection Findings - Kewaunee                                                                                                  Page 4 of 6 Green. The inspector identified a Non-Cited Violation of 10 CFR 55.21, "Medical Examination," and 10 CFR 55.23, "Certification." The inspector identified that the facility licensee failed to conduct all the medical testing required by American Nuclear Standards Institute/American Nuclear Society (ANSI/ANS) 3.4-1983, "Medical Certification and Monitoring of Personnel Requiring Operator Licenses for Nuclear Power Plants," as committed to by the facility licensee. Specifically, the facility licensee was not testing its operators for nose sensitivity (i.e., ability to detect odor of products of combustion and of tracer or market gases) Section 5.4.2, "Nose," and neurological testing, (i.e., normal central and peripheral nervous system function), including tactile discrimination (Stereognosis) sufficient to distinguish among various shapes of control knobs and hadles by touch, Section 5.4.14, "Neurological."
Inspection Report# : 2003005(pdf)
Significance: SL-IV Oct 21, 2003 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PROVIDE ACCURATE INFORMATION TO THE NRC CONCERNING LICENSED OPERATOR MEDICAL REQUIREMENTS PER NRC FORM 396.
Severity Level IV. The inspector identified a Level IV Non-Cited Violation of 10 CFR 50.9, "Completeness and Accuracy of Information." The inspector identified that the facility licensee, between January 2, 2000, thorugh August 26, 2002, submitted to the NRC, NRC Forms 396 for 13 individuals applying for an initial operator's license and 18 licensed operators applying for renewal of their operator licenses, that were not accurate in all material respects. Specifically, the NRC Forms 396 certified that each applicant and licensed operator met the medical requirements of ANSI/ANS 3.4-1983. In fact, all the applicants and licensed operators were not adequately examined for all medical tests as required to meet the minimum standards of ANSI/ANS 3.4-1983.
Inspection Report# : 2003005(pdf)
Significance:          Oct 21, 2003 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO CONDUCT SIMULATOR PERFORMANCE TESTING THROUGHOUT THE LIFE OF THE SIMULATOR.
Green. The inspector identified a Non-Cited Violation of 10 CFR 55.46, "Simulation Facility." The inspector identified that the facility licensee failed to adequately conduct simulator performance testing throughout the life of the simulator. In addition, the facility licensee failed to correct modeling and hardware discrepancies and discrepancies identified from scenario validation and from performance testing. In addition, the facility licensee was committed to follow ANSI/ANS 3.5-1985, "Nuclear Power Plant Simulators for Use in Operator Training," as the way they would meet 10 CFR 55.46. Specifically, the licensee failed to conduct performance testing, with regard to normal evolutions core performance tests for Cycle 25, the most recent core load in the actual reactor. The licensee could only provide Cycle 7 normal evolutions core performance tests. No core performance tests had ever been conducted for Cycles 8 through 25, a period of 17 cycles.
Inspection Report# : 2003005(pdf)
Significance: SL-IV Oct 21, 2003 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PROVIDE ACCURATE INFORMATION TO THE NRC CONCERNING ELIGIBILITY REQUIREMENTS FOR OPERATOR LICENSE APPLICATION PER NRC FORM 398.
Severity Level IV. The inspector identified a Level IV Non-Cited Violation of 10 CFR 50.9, "Completeness and Accuracy of Information." The inspector identified that on or about August 13, 2002, a senior facility licensee representative submitted to the NRC, NRC Forms 398 for three individuals, each applying for an initial operator's license, that were not accurate in all material respects. The facility licensee provided inaccurate information by certifying on the NRC Form 398 that the initial operator license applicaitons for three individuals had appropriately met the minimum training requirements for reactivity manipulations on the refrenced facility simulator in accordance with 10 CFR 55.31(a)(5) and 10 CFR 55.46(c)(2).
Inspection Report# : 2003005(pdf)
Significance:          Sep 30, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to Prescribe Instructions or Procedures Appropriate to the Circumstances for the Seismic Storage of Equipment Near the 'A' Auxiliary Feedwater Piping The inspectors identified a Green finding associated with a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures and Drawings," for the failure to prescribe instructions or procedures appropriate to the circumstances for the seismic control of equipment stored near the vicinity of the A' Auxiliary Feedwater (AFW) piping to the A' Steam Generator, an activity affecting quality. The inspectors identified during plant walkdowns that following the 2003 Refueling Outage, portable plant equipment, including two portable 2.5-ton cranes, were stored in close proximity to the AFW piping, without the use of seismic restraints.
Inspection Report# : 2003006(pdf) 07/14/2004
 
1Q/2004 Inspection Findings - Kewaunee                                                                                                  Page 5 of 6 Significance:        Sep 30, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to Prescribe Instructions or Procedures Appropriate to the Circumstances for the Installation of the Refueling Cavity Drain Standpipe Following Refueling Activities A Green finding associated with a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures and Drawings,"
was self-revealed when the licensee, in preparing and verifying the response to NRC Bulletin 2003-01, "Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized Water Reactors," dated June 9, 2003, determined that the containment refueling cavity standpipe had not been installed after the Spring 2003 Refueling Outage. A procedure revision, issued prior to the 2003 Outage, had removed prescribed instructions to install the refueling cavity drain standpipe following reactor vessel refueling activities. The inspectors also concluded that this finding had, as a primary cause, a human performance deficiency.
Inspection Report# : 2003006(pdf)
Significance:        Jun 30, 2003 Identified By: Self Disclosing Item Type: NCV NonCited Violation Failure to Ensure Material of Installed Pipe Plug in RHR System is in Accordance with Design Requirements A self-revealed non-cited violation 10 CFR 50, Appendix B, Criterion V, was identified for the licensee's failure to ensure that the residual heat removal pump recirculation piping material was in accordance with a facility drawing and engineering specifications. This resulted in the corrosion of three pipe plugs, one of which was corroded to the point of leaking. The pipe plugs were installed on each residual heat removal's recirculation pipe pressure breakdown orifice. The three pipe plugs were made of carbon steel while the residual heat removal system piping, which contained borated water, was required to be made of stainless steel.
This finding was greater than minor because it affected the Mitigating System Cornerstone objective of equipment reliability and availability, in that the failure to ensure that the residual heat removal piping materials are in accordance with plant engineering specifications and drawings could result in system leakage significant enough to require taking the system out-of-service. The finding is of very low risk significance because this finding was not a design or qualification deficiency which resulted in a loss of function per Generic Letter 91-18.
Inspection Report# : 2003004(pdf)
Barrier Integrity Emergency Preparedness Significance:        Dec 12, 2003 Identified By: NRC Item Type: FIN Finding Failure to take timely corrective actions to prevent recurrence for a 2001 white finding associated with Emergency Response Organization Augmentation The team identified a Green finding for the failure to take timely corrective actions to prevent recurrence for a White Finding initially identified in September 2000, associated with Emergency Response Organization Augmentation. While the team determined that corrective actions to date have been effective, as evidenced by only one augmentation drill failure since 2001, three of the eight corrective actions had not been completed.
The team determined that this issue was more than minor because if left uncorrected, the issue could become a more significant safety concern.
In addition, the team concluded that the issue affected the emergency preparedness cornerstone performance attribute associated with the emergency response organization augmentation system and emergency response augmentation testing and the objective of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The team evaluated the finding utilizing Inspection Manual Chapter 0609, Appendix B, "Emergency Preparedness Significance Determination Process," Section 5.0, "Corrective Actions," dated March 6, 2003, and determined the finding was of very low significance.
Inspection Report# : 2003010(pdf)
Significance:        Dec 12, 2003 Identified By: NRC 07/14/2004
 
1Q/2004 Inspection Findings - Kewaunee                                                                                                Page 6 of 6 Item Type: NCV NonCited Violation 10 CFR50.54 and emergency plan NCV for ineffective corrective actions in 2002 which resulted in the failure to make timely notifications for an actual unusual event in February 2003 The team identified a finding of very low significance associated with a Non-Cited Violation of 10 CFR Part 50.54(q) and the licensee's Emergency Plan for the failure to notify the state and local governmental agencies within 15 minutes after the declaration of an actual Unusual Event on February 26, 2003. The team concluded this failure was caused by the licensee's ineffective corrective actions for previously identified weaknesses and problems in the area of Emergency Preparedness.
The team determined that this issue was more than minor because this was an actual event implementation problem and affected the emergency preparedness cornerstone objective of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The team evaluated the finding utilizing Inspection Manual Chapter 0609, Appendix B, "Emergency Preparedness Significance Determination Process," Emergency Preparedness Significance Determination Process Sheet 2, dated March 6, 2003, and determined the finding was of very low significance.
Inspection Report# : 2003010(pdf)
Occupational Radiation Safety Public Radiation Safety Physical Protection Significance:        Jun 30, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to Report a Significant Fitness-for-Duty Event in a Timely Manner.
A finding of very low safety significance was identified by the inspectors for a violation of 10 CFR Part 26 Fitness-for-Duty (FFD) reporting requirements. The licensee failed to notify the NRC Operation Center within 24 hours of discovery of an illegal drug found within the licensee's protected area. The licensee failed to report the event because they did not realize this type of event was required to be reported.
The finding was determined to be of very low significance because it was a vulnerability in the licensee's Safeguards plan, was not a malevolent act, and similar findings had not occurred in the last four calendar quarters. The finding was determined to be more than minor because illegal drugs located within a licensee's protected area are required to be reported to the NRC in accordance with 10 CFR 26.73(a) requirements.
Inspection Report# : 2003004(pdf)
Miscellaneous Last modified : May 05, 2004 07/14/2004
 
2Q/2004 Inspection Findings - Kewaunee                                                                                                        Page 1 of 6 Kewaunee 2Q/2004 Plant Inspection Findings Initiating Events Mitigating Systems Significance:        Jun 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify and Correct Issues Associated with Historical Safety Injection Lube Oil Cooler Fouling; 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action" A finding of very low safety significance associated with a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action,"
was self-revealed on January 15, 2004, when licensee inspection of the A' and B' safety injection pump lube oil coolers identified silt and lake grass accumulation at the tube pass inlets. Significant fouling of the safety injection pump lube oil coolers with lake grass had been identified by the licensee as early as 1992 when the coolers were first opened and inspected. The licensee failed to enter the results of those inspections in the corrective action program when fouling was identified, until 2001. When the issue was entered into the corrective action program in 2001, following an inspection by plant personnel, the associated evaluation did not adequately address the issue and corrective actions were not taken in a timely manner to address the issue.
The licensee initiated numerous corrective actions to address the root and contributing causes identified during the root cause evaluation of this event.
Some of those actions included: replacing the old safety injection pump lube oil coolers with coolers of a new design; performing an extent of condition review of other service water systems prior to plant restart in January 2004 to ensure no similar immediate issues existed; sharing lessons learned from this event with all plant staff; and performing a prioritization review of all outstanding plant design modifications.
The inspectors verified the licensee's past operability analysis for the safety injection pumps. The inspectors evaluated the finding using the results of that analysis and Inspection Manual Chapter 0609, "Significance Determination Process," Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," Phase 1 screening, and determined that the finding was of very low safety significance.
Inspection Report# : 2004004(pdf)
Significance:        Jun 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to have Procedures Appropriate to the Circumstances, Including Appropriate Acceptance Criteria for Implementation of the Generic Letter 89-13 Program with Respect to the Safety Injection Lube A finding of very low safety significance associated with a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was self-revealed when the licensee discovered fouling of the safety injection pump lube oil coolers in January 2004. The licensee determined that evidence of the fouling had been present since the first inspection of the coolers in 1992. The licensee performed that first inspection as part of its actions to comply with Generic Letter 89-13, "Service Water System Problems Affecting Safety-Related Equipment." However, no acceptance criteria were included in the licensee's procedures developed to implement the commitments of Generic Letter 89-13 for these coolers to ensure that this activity had been satisfactorily accomplished.
The licensee initiated several corrective actions to address this issue, some of which included: establishing appropriate acceptance criteria for the safety injection lube oil coolers; developing a recovery plan for the licensee's Generic Letter 89-13 program and categorizing the program health in a red status; designating a single program owner to the Generic Letter 89-13 program; and reviewing other procedures utilized to implement the licensee's Generic Letter 89-13 program to verify specific acceptance criteria are contained in the procedures.
The inspectors verified the licensee's past operability analysis for the safety injection pumps. The inspectors evaluated the finding using the results of that analysis and Inspection Manual Chapter 0609, "Significance Determination Process," Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," Phase 1 screening, and determined that the finding was of very low safety significance.
Inspection Report# : 2004004(pdf)
Significance:        Mar 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO NOTIFY THE NRC OF A CHANGE IN OPERATOR STATUS IN ACCORDANCE WITH 10 CFR 50.74(c)
The inspector identified a violation of 10 CFR 50.74(c), "Notification of Change in Operator or Senior Operator Status." The inspector identified that
 
2Q/2004 Inspection Findings - Kewaunee                                                                                                          Page 2 of 6 the facility licensee failed to notify the NRC within 30 days after receiving a change in medical status of a licensed operator from the station's medical examiner. The change in medical status required conditioning of the operator's license by the NRC.
Inspection Report# : 2004002(pdf)
Significance:        Mar 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to have procedures appropriate to the circumstances, including appropriate acceptance criteria for determining important activities have been satisfactorily accomplished for inservice inspect A finding of very low safety significance associated with a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was self-revealed when, on December 10, 2003, licensee personnel discovered evidence of component cooling water (CCW) system leakage from a radiation detector housing integral to the CCW piping. The leakage was determined to be evidence of a through wall leak of an American Society of Mechanical Engineers (ASME) Section XI Class 3 pipe which rendered both trains of CCW inoperable. Therefore, on December 12, 2003, operators declared both trains of the CCW system inoperable, due to the small CCW leak on the CCW radiation detector housing. The licensee subsequently determined that evidence of the leakage had been present for the past 13 years. However, less than adequate procedure acceptance criteria resulted in licensee personnel classifying the leakage during inservice inspections as a non-recordable' indication instead of the required recordable' indication for this type leakage.
The licensee took immediate corrective actions to move the ASME Section XI Class 3 piping boundary to the radiation detector housing. In addition, the licensee implemented corrective actions to prevent recurrence which included: revision of 19 inservice inspection surveillance procedures to incorporate appropriate acceptance criteria; and discussions of the root cause with licensee inservice inspection personnel as a lessons-learned.
This self-revealed finding was greater than minor because, if left uncorrected, the finding would become a more significant safety concern. The inspectors evaluated the finding using the Significance Determination Process, Appendix A, Phase 1 Screening, and determined that the finding was of very low safety significance.
Inspection Report# : 2004002(pdf)
Significance:        Jan 28, 2004 Identified By: NRC Item Type: FIN Finding Failure to appropriately evaluate for potential bypqss flow on service water strainers The inspectors identified a finding of very low safety significance associated with the licensee's failure to appropriately evaluate for potential bypass flow on the service water pump discharge strainers by measuring a critical gap dimension at the bottom of the basket-to-housing interface. This finding did not constitute a violation of NRC requirements because the strainers (aside from the pressure boundary) did not fulfill a safety-related function.
The inspectors determined that the finding was of more than minor significance because it would become a more significant safety concern if left uncorrected. Specifically, the failure to appropriately evaluate for potential bypass flow on the service water pump discharge strainers could reasonably result in debris fouling of service water cooled components and degraded or inoperable safety-related equipment. The inspectors concluded that this finding was a licensee performance deficiency of very low safety significance because it did not result in loss of safety function for a service water system train for greater than its Technical Specification allowed outage time. To address this issue, the licensee opened each strainer and measured the gap at the bottom of the basket-to-housing interface.
Inspection Report# : 2004003(pdf)
Significance: SL-IV Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation Non-cited violation of 10 CFR 50.59, for the failure to perform a written evaluation, as required, for a modification to the component cooling water system The inspectors identified a finding of very low safety significance associated with a Non-Cited Violation of 10 CFR 50.59(d)(1) for the licensee's failure to perform a safety evaluation for changes made to the facility. Specifically, the licensee screened out' of the 10 CFR 50.59 process a modification that included the addition of a minimum flow recirculation line to the component cooling water pumps. This modification further cross-connected the suction and discharge piping of both component cooling water pump trains. Subsequently, the inspectors identified and the licensee concurred that a safety evaluation was required for this modification.
Because the Significance Determination Process is not designed to assess the significance of violations that potentially impact or impeded the regulatory process, this issue was dispositioned using the traditional enforcement process in accordance with Section IV of the NRC Enforcement Policy. However, the results of this violation were assessed using the Significance Determination Process. In this case, the licensee failed to perform a safety evaluation in accordance with 10 CFR 50.59 and had placed the new system in service for testing prior to the completion of the required safety evaluation.
The inspectors considered this issue to be of more than minor significance because, if left uncorrected, the issue could become a more significant safety concern. Specifically, the inspectors noted that the licensee's processes for permanent modifications failed to identify this issue at several review levels.
The inspectors determined that the issue was of very low significance because the new system was placed in service for a short period of time for testing prior to the completion of the required safety evaluation. In addition, the final safety evaluation completed by the licensee in October 2003 determined that the modification did not require prior NRC approval. The inspectors determined this finding was a Severity Level IV Non-Cited Violation of 10 CFR 50.59. The inspectors also determined that the finding had, as a primary cause, a human performance deficiency which affected the cross-cutting are of Human Performance.
 
2Q/2004 Inspection Findings - Kewaunee                                                                                                        Page 3 of 6 Inspection Report# : 2003008(pdf)
Significance:        Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation 10 CFR 50, Appendix B, Criterion III, "Design Control," the failure to provide for the checking the adequacy of design for temporary mod which changed the CCW system pressure boundary The inspectors identified a finding of very low safety significance associated with a Non-Cited Violation of 10 CFR Part 50 Appendix B, Criterion III, "Design Control," for the licensee's failure to provide for checking of the adequacy of the design in Temporary Change TCR 03-036, in that, the design review failed to confirm the structural integrity of the new pressure boundary established for the studding outlet. Consequently, the licensee performed non-destructive examinations and additional flaw and engineering analyses to confirm the adequacy of the new design.
The inspectors considered this issue of more than minor significance, because if left uncorrected, the issue could become a more significant safety concern. In addition, the inspectors concluded that the finding was greater than minor because the finding involved the design control attribute of the mitigating systems cornerstone and affected the mitigating systems objective of ensuring the capability of the component cooling water system in response to initiating events to prevent undesirable consequences. Specifically, the temporary design change relied on unsupported assumptions that could have impacted the structural integrity of the component cooling water suction line. The inspectors evaluated the finding using the Significance Determination Process Phase 1 screening and determined that the finding was a design or qualification deficiency confirmed not to result in loss of function per Generic Letter 91-18; therefore, the finding was determined to be of very low safety significance.
Inspection Report# : 2003008(pdf)
Significance:        Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation 10 CFR 50, Appendix B, Criterion XVI, "Corrective Actions," No appropriate immediate corrective actions for reliability issues associated with incorrect cranking cutout relay installed in the EDGs The inspectors identified a finding of very low safety significance associated with a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Actions," for the licensee's failure to take adequate corrective actions in response to the installation of non-conforming cranking cutout relays which prevented energizing of the diesel generator engine start relay. The licensee's corrective actions for this condition adverse to quality addressed routine surveillance procedures, but did not consider the licensee's Emergency Operating Procedures to ensure the Emergency Diesel Generators would remain operable following Diesel Generator Shutdowns as directed by those procedures.
The inspectors considered this issue of more than minor significance, because if left uncorrected, the issue could become a more significant safety concern. In addition, the inspectors concluded that the finding was greater than minor because the finding involved the design control attribute of the mitigating systems cornerstone and affected the mitigating systems objective of ensuring the capability of the diesel generators in response to initiating events to prevent undesirable consequences. Specifically, in part, the licensee's corrective actions included revisions to normal operating procedures to verify continuity across the relay contacts following shutdown of the emergency diesel generators; however, the licensee did not similarly revise its Emergency Operating Procedures to verify continuity across the cranking cutout relay contacts following shutdown of the emergency diesel generators.
The inspectors evaluated the finding using the Significance Determination Process Phase 1 screening and determined that the finding was a design or qualification deficiency confirmed not to result in loss of function per Generic Letter 91-18; therefore, the finding was determined to be of very low safety significance.
Inspection Report# : 2003008(pdf)
Significance:        Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation 10 CFR 50, Appendix B, Criterion III, "Design Control," failure to install the appropriate cranking cutout relay in the EDG system in 1998; this resulted in failure of 'B' EDG to start in Feb., 2003 A finding of very low safety significance involving a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," was self-revealing when the "B" emergency diesel generator failed to start on February 26, 2003, during a daily Technical Specification-required test, in response to the "A" emergency diesel generator being out of service for regularly scheduled 18-month periodic maintenance. The generator failed to start due to a pair of electrically open contacts on a cranking cutout relay which prevented energizing of the engine start relay. The cranking cutout relay had been installed during a design change completed in 1998, and the performance ratings of the new relay did not match original design specifications.
The inspectors considered this issue of more than minor significance, because if left uncorrected, the issue could become a more significant safety concern. In addition, the inspectors concluded that the finding was greater than minor because the finding involved the design control attribute of the mitigating systems cornerstone and affected the mitigating systems objective of ensuring the capability of the diesel generators in response to initiating events to prevent undesirable consequences. Specifically, the temporary design change failed to consider inductive electrical loads across the relay contacts, for which the relays were not rated. The inspectors evaluated the finding using the Significance Determination Process Phase 1 screening and determined that the finding was a design or qualification deficiency confirmed not to result in loss of function per Generic Letter 91-18; therefore, the finding was determined to be of very low safety significance.
Inspection Report# : 2003008(pdf)
 
2Q/2004 Inspection Findings - Kewaunee                                                                                                                Page 4 of 6 Significance:        Dec 12, 2003 Identified By: NRC Item Type: NCV NonCited Violation 10 CFR 50, Appendix B, Criterion XVI NCV for ineffective corrective actions taken to address the implementation of the Boric Acid Leakage Inspection and Tracking Program The team identified a finding of very low significance associated with a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," for the failure to assure that actions were promptly taken to correct deficiencies in the implementation of the boric acid leakage inspection and tracking program for boric acid residue on safety-related components, a condition adverse to quality. Since 2001, approximately 12 condition reports had been initiated concerning the adequacy of the implementation of the licensee's boric acid leakage inspection and tracking program.
During the inspection, the team identified approximately 14 safety-related components with various degrees of boric acid, which the licensee had not identified and evaluated in accordance with the boric acid leakage inspection and tracking program.
The team concluded that the licensee's failure to correct previous issues associated with the implementation of the boric acid leak log on safety-related components was greater than minor because if left uncorrected, the issue could become a more significant safety concern. The team evaluated the finding utilizing Inspection Manual Chapter 0609, "Significance Determination Process," Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," Phase 1 screening and determined the finding was of very low significance.
Inspection Report# : 2003010(pdf)
Significance:        Oct 21, 2003 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PERFORM TWO REQUIRED MEDICAL TESTS IN ACCORDANCE WITH 10 CFR 55.21 AND 55.23.
Green. The inspector identified a Non-Cited Violation of 10 CFR 55.21, "Medical Examination," and 10 CFR 55.23, "Certification." The inspector identified that the facility licensee failed to conduct all the medical testing required by American Nuclear Standards Institute/American Nuclear Society (ANSI/ANS) 3.4-1983, "Medical Certification and Monitoring of Personnel Requiring Operator Licenses for Nuclear Power Plants," as committed to by the facility licensee. Specifically, the facility licensee was not testing its operators for nose sensitivity (i.e., ability to detect odor of products of combustion and of tracer or market gases) Section 5.4.2, "Nose," and neurological testing, (i.e., normal central and peripheral nervous system function),
including tactile discrimination (Stereognosis) sufficient to distinguish among various shapes of control knobs and hadles by touch, Section 5.4.14, "Neurological."
Inspection Report# : 2003005(pdf)
Significance: SL-IV Oct 21, 2003 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PROVIDE ACCURATE INFORMATION TO THE NRC CONCERNING LICENSED OPERATOR MEDICAL REQUIREMENTS PER NRC FORM 396.
Severity Level IV. The inspector identified a Level IV Non-Cited Violation of 10 CFR 50.9, "Completeness and Accuracy of Information." The inspector identified that the facility licensee, between January 2, 2000, thorugh August 26, 2002, submitted to the NRC, NRC Forms 396 for 13 individuals applying for an initial operator's license and 18 licensed operators applying for renewal of their operator licenses, that were not accurate in all material respects. Specifically, the NRC Forms 396 certified that each applicant and licensed operator met the medical requirements of ANSI/ANS 3.4-1983. In fact, all the applicants and licensed operators were not adequately examined for all medical tests as required to meet the minimum standards of ANSI/ANS 3.4-1983.
Inspection Report# : 2003005(pdf)
Significance:        Oct 21, 2003 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO CONDUCT SIMULATOR PERFORMANCE TESTING THROUGHOUT THE LIFE OF THE SIMULATOR.
Green. The inspector identified a Non-Cited Violation of 10 CFR 55.46, "Simulation Facility." The inspector identified that the facility licensee failed to adequately conduct simulator performance testing throughout the life of the simulator. In addition, the facility licensee failed to correct modeling and hardware discrepancies and discrepancies identified from scenario validation and from performance testing. In addition, the facility licensee was committed to follow ANSI/ANS 3.5-1985, "Nuclear Power Plant Simulators for Use in Operator Training," as the way they would meet 10 CFR 55.46.
Specifically, the licensee failed to conduct performance testing, with regard to normal evolutions core performance tests for Cycle 25, the most recent core load in the actual reactor. The licensee could only provide Cycle 7 normal evolutions core performance tests. No core performance tests had ever been conducted for Cycles 8 through 25, a period of 17 cycles.
Inspection Report# : 2003005(pdf)
Significance: SL-IV Oct 21, 2003 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PROVIDE ACCURATE INFORMATION TO THE NRC CONCERNING ELIGIBILITY REQUIREMENTS FOR OPERATOR LICENSE APPLICATION PER NRC FORM 398.
Severity Level IV. The inspector identified a Level IV Non-Cited Violation of 10 CFR 50.9, "Completeness and Accuracy of Information." The inspector identified that on or about August 13, 2002, a senior facility licensee representative submitted to the NRC, NRC Forms 398 for three
 
2Q/2004 Inspection Findings - Kewaunee                                                                                                            Page 5 of 6 individuals, each applying for an initial operator's license, that were not accurate in all material respects. The facility licensee provided inaccurate information by certifying on the NRC Form 398 that the initial operator license applicaitons for three individuals had appropriately met the minimum training requirements for reactivity manipulations on the refrenced facility simulator in accordance with 10 CFR 55.31(a)(5) and 10 CFR 55.46(c)(2).
Inspection Report# : 2003005(pdf)
Significance:        Sep 30, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to Prescribe Instructions or Procedures Appropriate to the Circumstances for the Seismic Storage of Equipment Near the 'A' Auxiliary Feedwater Piping The inspectors identified a Green finding associated with a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures and Drawings," for the failure to prescribe instructions or procedures appropriate to the circumstances for the seismic control of equipment stored near the vicinity of the A' Auxiliary Feedwater (AFW) piping to the A' Steam Generator, an activity affecting quality. The inspectors identified during plant walkdowns that following the 2003 Refueling Outage, portable plant equipment, including two portable 2.5-ton cranes, were stored in close proximity to the AFW piping, without the use of seismic restraints.
Inspection Report# : 2003006(pdf)
Significance:        Sep 30, 2003 Identified By: NRC Item Type: NCV NonCited Violation Failure to Prescribe Instructions or Procedures Appropriate to the Circumstances for the Installation of the Refueling Cavity Drain Standpipe Following Refueling Activities A Green finding associated with a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures and Drawings," was self-revealed when the licensee, in preparing and verifying the response to NRC Bulletin 2003-01, "Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized Water Reactors," dated June 9, 2003, determined that the containment refueling cavity standpipe had not been installed after the Spring 2003 Refueling Outage. A procedure revision, issued prior to the 2003 Outage, had removed prescribed instructions to install the refueling cavity drain standpipe following reactor vessel refueling activities. The inspectors also concluded that this finding had, as a primary cause, a human performance deficiency.
Inspection Report# : 2003006(pdf)
Barrier Integrity Significance:        Jun 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct historical residual heat removal pump mechanical seal leakage; 10CFR Part 50, Appendix B, Criterion XVI, "Corrective Action" A finding of very low safety significance associated with a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Actions",
was self-revealed on June 16, 2004, when licensee personnel discovered leakage from the B' residual heat removal (RHR) pump seal when the pump was stopped following the performance of a surveillance procedure on the B' RHR train. Plant personnel determined the leakage to be in excess of that specified in the plant's System Integrity Program for leakage from emergency core cooling systems outside containment. The leakage was also in excess of the amount of leakage assumed in the Updated Safety Analysis Report, Chapter 14, for calculation of control room habitability doses and offsite exposures. The inspectors subsequently determined, from interviews with licensee personnel and a review of the licensee's corrective action program and work order history, that excessive RHR seal leakage has occurred since the late 1980s. However, past corrective actions have not been effective to correct this condition adverse to quality.
The licensee performed a prompt engineering review to ensure that no immediate catastrophic failure mechanism for the RHR seal existed. The licensee also performed a prompt engineering review of the impact of the estimated leakage on the control room habitability doses, as well as the offsite doses, and determined no exposure limits would be exceeded. The licensee took actions to immediately stop the leakage and plans to replace the RHR pump seal during the next refueling outage.
This self-revealed finding was more than minor because the finding affected the cornerstone objective of Reactor Safety/Barrier Integrity. The inspectors evaluated the finding using Inspection Manual Chapter 0609, "Significance Determination Process," Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," Phase 1 screening, and determined that the finding was of very low safety significance.
Inspection Report# : 2004004(pdf)
 
2Q/2004 Inspection Findings - Kewaunee                                                                                                        Page 6 of 6 Emergency Preparedness Significance:        Dec 12, 2003 Identified By: NRC Item Type: FIN Finding Failure to take timely corrective actions to prevent recurrence for a 2001 white finding associated with Emergency Response Organization Augmentation The team identified a Green finding for the failure to take timely corrective actions to prevent recurrence for a White Finding initially identified in September 2000, associated with Emergency Response Organization Augmentation. While the team determined that corrective actions to date have been effective, as evidenced by only one augmentation drill failure since 2001, three of the eight corrective actions had not been completed.
The team determined that this issue was more than minor because if left uncorrected, the issue could become a more significant safety concern. In addition, the team concluded that the issue affected the emergency preparedness cornerstone performance attribute associated with the emergency response organization augmentation system and emergency response augmentation testing and the objective of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The team evaluated the finding utilizing Inspection Manual Chapter 0609, Appendix B, "Emergency Preparedness Significance Determination Process," Section 5.0, "Corrective Actions," dated March 6, 2003, and determined the finding was of very low significance.
Inspection Report# : 2003010(pdf)
Significance:        Dec 12, 2003 Identified By: NRC Item Type: NCV NonCited Violation 10 CFR50.54 and emergency plan NCV for ineffective corrective actions in 2002 which resulted in the failure to make timely notifications for an actual unusual event in February 2003 The team identified a finding of very low significance associated with a Non-Cited Violation of 10 CFR Part 50.54(q) and the licensee's Emergency Plan for the failure to notify the state and local governmental agencies within 15 minutes after the declaration of an actual Unusual Event on February 26, 2003. The team concluded this failure was caused by the licensee's ineffective corrective actions for previously identified weaknesses and problems in the area of Emergency Preparedness.
The team determined that this issue was more than minor because this was an actual event implementation problem and affected the emergency preparedness cornerstone objective of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The team evaluated the finding utilizing Inspection Manual Chapter 0609, Appendix B, "Emergency Preparedness Significance Determination Process," Emergency Preparedness Significance Determination Process Sheet 2, dated March 6, 2003, and determined the finding was of very low significance.
Inspection Report# : 2003010(pdf)
Occupational Radiation Safety Public Radiation Safety Physical Protection Physical Protection information not publicly available.
Miscellaneous Last modified : September 08, 2004
 
3Q/2004 Inspection Findings - Kewaunee                                                                                                Page 1 of 9 Kewaunee 3Q/2004 Plant Inspection Findings Initiating Events Mitigating Systems Significance:        Sep 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action." Failure to Promptly Correct Conditions Adverse to Quality, Specifically Associated with Degraded and Nonconforming Conditions A finding of very low safety significance was identified by the inspectors for a violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Actions." During a review of the licensee's list of safety-related equipment designated as degraded or nonconforming, the inspectors identified that the licensee failed to promptly correct three conditions adverse to quality. These conditions adverse to quality included noncompliance of both Residual Heat Removal pump seal coolers with system design requirements, which was previously identified by NRC inspectors in November 2002, but not promptly corrected by the licensee; and two sections of safety-related piping, one associated with the "B" Emergency Diesel Generator fuel oil supply and the other associated with the Component Cooling Water piping from the "B" Residual Heat Removal pump seal cooler and stuffing box, that were identified by the licensee in September and April 2003, respectively, as exceeding Updated Safety Analysis Report stress criteria but not promptly corrected by the licensee. The primary cause of this finding was related to the cross-cutting area of problem identification and resolution. The licensee failed to prioritize and promptly correct these conditions adverse to quality in accordance with the guidelines in the corrective action program. Once these conditions were identified, the licensee restored the following conditions to operable: the A' RHR Pump Seal Cooler; the CCW piping expansion loop from the B' RHR pump seal cooler; and the fuel oil supply piping to the B' EDG. The licensee planned to restore the B' RHR Pump Seal Cooler during the upcoming Fall 2004 Refueling Outage.
This issue was more than minor because it affected the Mitigating System cornerstone attribute of design control for initial design and plant modifications and affected the associated cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was of very low safety significance because it was not a design or qualification deficiency that has been confirmed to result in a loss of function per Generic Letter 91-18. This issue was a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Actions."
Inspection Report# : 2004007(pdf)
Significance:        Sep 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings." Failure to Have Procedures Appropriate to the Circumstances for Preventive Maintenance of the TDAFW Pump Turbine A finding of very low safety significance was self-revealed during the licensee's review of high oil particulate in the Turbine Driven Auxiliary Feedwater Pump Turbine, which resulted in a violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings."
The licensee determined that high oil particulate in the Turbine Driven Auxiliary Feedwater Pump Turbine was caused by damage to the journal bearing. Maintenance procedures did not specify appropriate acceptance criteria for oil sampling, did not specify an appropriate inspection frequency and criteria for the turbine bearings and bearing cavities, and allowed the reuse of bearings in different locations during maintenance of the Turbine, which were not acceptable maintenance practices. The reuse of the upper inboard bearing in a different location contributed to the journal bearing damage. The licensee took immediate remedial corrective actions to replace the bearings, clean the housing and return the pump to service. In addition, the licensee revised its maintenance procedures to include appropriate instructions for turbine and pump maintenance activities.
This self-revealed finding was more than minor because, if left uncorrected, the issue would have become a more significant safety concern. In addition, it affected the Mitigating Systems attributes of equipment performance reliability and the Mitigating Systems cornerstone objective of ensuring the reliability of systems. The finding was of very low safety significance because it was not a design or qualification deficiency that has been confirmed to result in a loss of function per Generic Letter 91-18. This issue was a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, And Drawings."
Inspection Report# : 2004007(pdf)
 
3Q/2004 Inspection Findings - Kewaunee                                                                                                    Page 2 of 9 Significance:        Sep 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings." Failure to Have Acceptance Criteria for Flushing of the 1ARHR Fan Coil Unit A finding of very low safety significance was identified by the inspectors for a violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, And Drawings." This finding was associated with the licensee's failure to implement an appropriate inspection and cleaning procedure containing quantitative or qualitative acceptance criteria for the 1A RHR pump pit Fan Coil Unit to ensure that cleaning was satisfactorily accomplished. Following discovery, the licensee entered the issue into its corrective action program and conducted an immediate operability assessment that determined the involved fan coil units were operable.
This issue was more than minor because it involved the procedure quality 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. The finding was of very low safety significance because it was not a design or qualification deficiency that has been confirmed to result in a loss of function per Generic Letter 91-18. This issue was a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, And Drawings."
Inspection Report# : 2004007(pdf)
Significance:        Sep 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation 10 CFR 50, Appendix B, Criterion III, "Design Control." Failure to Verify the Acceptability of a Single Failure Vulnerability Introduced During a System Modification 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." This finding was associated with the licensee's failure to perform a design verification to demonstrate that the diesel generator lube oil cooler service water outlet valve actuators, installed under Design Change 3357, would not result in a failure of the valve stems under conditions in which the valve ball froze nor had the licensee provided sufficient justification to show that valve ball freezing was not credible.
Following discovery, the licensee entered the issue into its corrective action program and performed an operability assessment which provided additional justification to demonstrate that the stem failure was considered not credible.
This issue was more than minor because it involved 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. The finding was of very low safety significance because it was not a design or qualification deficiency that has been confirmed to result in a loss of function per Generic Letter 91-18. This issue was a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control."
Inspection Report# : 2004007(pdf)
Significance:        Jul 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Repair the Deluge System Heat Detectors in a Timely Manner (1R05.10.b.1)
The team identified a Non-Cited Violation NCV of License Condition 2.C(3) having very low safety significance (Green) for the failure to repair a deluge sprinkler system in a timely manner.
Inspection Report# : 2004005(pdf)
Significance:        Jul 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Acceptable (Quality Related) Pre-Fire Strategies (1R05.10.b.2)
The team identified a NCV of License Condition 2.C.(3) having very low safety significance (Green) for failing to include pertinent information in fire strategies. Specifically, the licensee failed to include information about the potential unavailability of certain fire hose stations and identify hydrogen and propane piping hazards in a fire zone.
Inspection Report# : 2004005(pdf)
Significance:        Jul 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Meet the Fire Protection Program Requirements for Hose Lengths to Maintain an Acceptable Water Pressure and Flow at Hose Stations (1R05.10.b.3)
The team identified an NCV of the Kewaunee Fire Protection Program Plan, License Condition 2.C(3), having very low safety significance
 
3Q/2004 Inspection Findings - Kewaunee                                                                                                Page 3 of 9 (Green) for failure to meet the fire protection program requirement to maintain an acceptable water pressure and flow at hose stations.
Inspection Report# : 2004005(pdf)
Significance:        Jul 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Meet the NFPA Code Requirements for Extinguisher Placement (1R05.10.b.4)
The team identified an NCV of License Condition 2.C.(3), which requires the licensee to implement all provisions of their approved fire protection program. The licensee failed to maintain NFPA Code requirements for the number of Class A extinguishers.
Inspection Report# : 2004005(pdf)
Significance:        Jul 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Transient Combustibles Not Adequately Controlled Within Fire Area AX-32 (1R05.10.b.5(1))
The team identified a Non-Cited Violation (NCV) of License Condition 2.C.(3) having very low safety significance (Green) for failure to adequately control transient combustibles in Fire Area AX-32.
Inspection Report# : 2004005(pdf)
Significance:        Jul 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Transient Combustibles Not Adequately Controlled Within Fire Area AX-24 (1R05.10.b.5(2))
The team identified a Non-Cited Violation (NCV) of License Condition 2.C.(3) having very low safety significance (Green) for failure to adequately control transient combustibles in Fire Area AX-24.
Inspection Report# : 2004005(pdf)
Significance:        Jun 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify and Correct Issues Associated with Historical Safety Injection Lube Oil Cooler Fouling; 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action" A finding of very low safety significance associated with a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was self-revealed on January 15, 2004, when licensee inspection of the A' and B' safety injection pump lube oil coolers identified silt and lake grass accumulation at the tube pass inlets. Significant fouling of the safety injection pump lube oil coolers with lake grass had been identified by the licensee as early as 1992 when the coolers were first opened and inspected. The licensee failed to enter the results of those inspections in the corrective action program when fouling was identified, until 2001. When the issue was entered into the corrective action program in 2001, following an inspection by plant personnel, the associated evaluation did not adequately address the issue and corrective actions were not taken in a timely manner to address the issue.
The licensee initiated numerous corrective actions to address the root and contributing causes identified during the root cause evaluation of this event. Some of those actions included: replacing the old safety injection pump lube oil coolers with coolers of a new design; performing an extent of condition review of other service water systems prior to plant restart in January 2004 to ensure no similar immediate issues existed; sharing lessons learned from this event with all plant staff; and performing a prioritization review of all outstanding plant design modifications.
The inspectors verified the licensee's past operability analysis for the safety injection pumps. The inspectors evaluated the finding using the results of that analysis and Inspection Manual Chapter 0609, "Significance Determination Process," Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," Phase 1 screening, and determined that the finding was of very low safety significance.
Inspection Report# : 2004004(pdf)
Significance:        Jun 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to have Procedures Appropriate to the Circumstances, Including Appropriate Acceptance Criteria for Implementation of the Generic Letter 89-13 Program with Respect to the Safety Injection Lube A finding of very low safety significance associated with a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions,
 
3Q/2004 Inspection Findings - Kewaunee                                                                                                Page 4 of 9 Procedures, and Drawings," was self-revealed when the licensee discovered fouling of the safety injection pump lube oil coolers in January 2004. The licensee determined that evidence of the fouling had been present since the first inspection of the coolers in 1992. The licensee performed that first inspection as part of its actions to comply with Generic Letter 89-13, "Service Water System Problems Affecting Safety-Related Equipment." However, no acceptance criteria were included in the licensee's procedures developed to implement the commitments of Generic Letter 89-13 for these coolers to ensure that this activity had been satisfactorily accomplished.
The licensee initiated several corrective actions to address this issue, some of which included: establishing appropriate acceptance criteria for the safety injection lube oil coolers; developing a recovery plan for the licensee's Generic Letter 89-13 program and categorizing the program health in a red status; designating a single program owner to the Generic Letter 89-13 program; and reviewing other procedures utilized to implement the licensee's Generic Letter 89-13 program to verify specific acceptance criteria are contained in the procedures.
The inspectors verified the licensee's past operability analysis for the safety injection pumps. The inspectors evaluated the finding using the results of that analysis and Inspection Manual Chapter 0609, "Significance Determination Process," Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," Phase 1 screening, and determined that the finding was of very low safety significance.
Inspection Report# : 2004004(pdf)
Significance:        Mar 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO NOTIFY THE NRC OF A CHANGE IN OPERATOR STATUS IN ACCORDANCE WITH 10 CFR 50.74(c)
The inspector identified a violation of 10 CFR 50.74(c), "Notification of Change in Operator or Senior Operator Status." The inspector identified that the facility licensee failed to notify the NRC within 30 days after receiving a change in medical status of a licensed operator from the station's medical examiner. The change in medical status required conditioning of the operator's license by the NRC.
Inspection Report# : 2004002(pdf)
Significance:        Mar 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to have procedures appropriate to the circumstances, including appropriate acceptance criteria for determining important activities have been satisfactorily accomplished for inservice inspect A finding of very low safety significance associated with a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was self-revealed when, on December 10, 2003, licensee personnel discovered evidence of component cooling water (CCW) system leakage from a radiation detector housing integral to the CCW piping. The leakage was determined to be evidence of a through wall leak of an American Society of Mechanical Engineers (ASME) Section XI Class 3 pipe which rendered both trains of CCW inoperable. Therefore, on December 12, 2003, operators declared both trains of the CCW system inoperable, due to the small CCW leak on the CCW radiation detector housing. The licensee subsequently determined that evidence of the leakage had been present for the past 13 years.
However, less than adequate procedure acceptance criteria resulted in licensee personnel classifying the leakage during inservice inspections as a non-recordable' indication instead of the required recordable' indication for this type leakage.
The licensee took immediate corrective actions to move the ASME Section XI Class 3 piping boundary to the radiation detector housing. In addition, the licensee implemented corrective actions to prevent recurrence which included: revision of 19 inservice inspection surveillance procedures to incorporate appropriate acceptance criteria; and discussions of the root cause with licensee inservice inspection personnel as a lessons-learned.
This self-revealed finding was greater than minor because, if left uncorrected, the finding would become a more significant safety concern. The inspectors evaluated the finding using the Significance Determination Process, Appendix A, Phase 1 Screening, and determined that the finding was of very low safety significance.
Inspection Report# : 2004002(pdf)
Significance:        Jan 28, 2004 Identified By: NRC Item Type: FIN Finding Failure to appropriately evaluate for potential bypqss flow on service water strainers The inspectors identified a finding of very low safety significance associated with the licensee's failure to appropriately evaluate for potential bypass flow on the service water pump discharge strainers by measuring a critical gap dimension at the bottom of the basket-to-housing interface. This finding did not constitute a violation of NRC requirements because the strainers (aside from the pressure boundary) did not fulfill a safety-related function.
The inspectors determined that the finding was of more than minor significance because it would become a more significant safety concern if left uncorrected. Specifically, the failure to appropriately evaluate for potential bypass flow on the service water pump discharge strainers could reasonably result in debris fouling of service water cooled components and degraded or inoperable safety-related equipment. The inspectors concluded that this finding was a licensee performance deficiency of very low safety significance because it did not result in loss of safety
 
3Q/2004 Inspection Findings - Kewaunee                                                                                                  Page 5 of 9 function for a service water system train for greater than its Technical Specification allowed outage time. To address this issue, the licensee opened each strainer and measured the gap at the bottom of the basket-to-housing interface.
Inspection Report# : 2004003(pdf)
Significance: SL-IV Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation Non-cited violation of 10 CFR 50.59, for the failure to perform a written evaluation, as required, for a modification to the component cooling water system The inspectors identified a finding of very low safety significance associated with a Non-Cited Violation of 10 CFR 50.59(d)(1) for the licensee's failure to perform a safety evaluation for changes made to the facility. Specifically, the licensee screened out' of the 10 CFR 50.59 process a modification that included the addition of a minimum flow recirculation line to the component cooling water pumps. This modification further cross-connected the suction and discharge piping of both component cooling water pump trains. Subsequently, the inspectors identified and the licensee concurred that a safety evaluation was required for this modification.
Because the Significance Determination Process is not designed to assess the significance of violations that potentially impact or impeded the regulatory process, this issue was dispositioned using the traditional enforcement process in accordance with Section IV of the NRC Enforcement Policy. However, the results of this violation were assessed using the Significance Determination Process. In this case, the licensee failed to perform a safety evaluation in accordance with 10 CFR 50.59 and had placed the new system in service for testing prior to the completion of the required safety evaluation.
The inspectors considered this issue to be of more than minor significance because, if left uncorrected, the issue could become a more significant safety concern. Specifically, the inspectors noted that the licensee's processes for permanent modifications failed to identify this issue at several review levels. The inspectors determined that the issue was of very low significance because the new system was placed in service for a short period of time for testing prior to the completion of the required safety evaluation. In addition, the final safety evaluation completed by the licensee in October 2003 determined that the modification did not require prior NRC approval. The inspectors determined this finding was a Severity Level IV Non-Cited Violation of 10 CFR 50.59. The inspectors also determined that the finding had, as a primary cause, a human performance deficiency which affected the cross-cutting are of Human Performance.
Inspection Report# : 2003008(pdf)
Significance:        Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation 10 CFR 50, Appendix B, Criterion III, "Design Control," the failure to provide for the checking the adequacy of design for temporary mod which changed the CCW system pressure boundary The inspectors identified a finding of very low safety significance associated with a Non-Cited Violation of 10 CFR Part 50 Appendix B, Criterion III, "Design Control," for the licensee's failure to provide for checking of the adequacy of the design in Temporary Change TCR 03-036, in that, the design review failed to confirm the structural integrity of the new pressure boundary established for the studding outlet.
Consequently, the licensee performed non-destructive examinations and additional flaw and engineering analyses to confirm the adequacy of the new design.
The inspectors considered this issue of more than minor significance, because if left uncorrected, the issue could become a more significant safety concern. In addition, the inspectors concluded that the finding was greater than minor because the finding involved the design control attribute of the mitigating systems cornerstone and affected the mitigating systems objective of ensuring the capability of the component cooling water system in response to initiating events to prevent undesirable consequences. Specifically, the temporary design change relied on unsupported assumptions that could have impacted the structural integrity of the component cooling water suction line. The inspectors evaluated the finding using the Significance Determination Process Phase 1 screening and determined that the finding was a design or qualification deficiency confirmed not to result in loss of function per Generic Letter 91-18; therefore, the finding was determined to be of very low safety significance.
Inspection Report# : 2003008(pdf)
Significance:        Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation 10 CFR 50, Appendix B, Criterion XVI, "Corrective Actions," No appropriate immediate corrective actions for reliability issues associated with incorrect cranking cutout relay installed in the EDGs The inspectors identified a finding of very low safety significance associated with a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Actions," for the licensee's failure to take adequate corrective actions in response to the installation of non-conforming cranking cutout relays which prevented energizing of the diesel generator engine start relay. The licensee's corrective actions for this condition adverse to quality addressed routine surveillance procedures, but did not consider the licensee's Emergency Operating Procedures to ensure the Emergency Diesel Generators would remain operable following Diesel Generator Shutdowns as directed by those procedures.
The inspectors considered this issue of more than minor significance, because if left uncorrected, the issue could become a more significant safety concern. In addition, the inspectors concluded that the finding was greater than minor because the finding involved the design control
 
3Q/2004 Inspection Findings - Kewaunee                                                                                                  Page 6 of 9 attribute of the mitigating systems cornerstone and affected the mitigating systems objective of ensuring the capability of the diesel generators in response to initiating events to prevent undesirable consequences. Specifically, in part, the licensee's corrective actions included revisions to normal operating procedures to verify continuity across the relay contacts following shutdown of the emergency diesel generators; however, the licensee did not similarly revise its Emergency Operating Procedures to verify continuity across the cranking cutout relay contacts following shutdown of the emergency diesel generators. The inspectors evaluated the finding using the Significance Determination Process Phase 1 screening and determined that the finding was a design or qualification deficiency confirmed not to result in loss of function per Generic Letter 91-18; therefore, the finding was determined to be of very low safety significance.
Inspection Report# : 2003008(pdf)
Significance:          Dec 31, 2003 Identified By: NRC Item Type: NCV NonCited Violation 10 CFR 50, Appendix B, Criterion III, "Design Control," failure to install the appropriate cranking cutout relay in the EDG system in 1998; this resulted in failure of 'B' EDG to start in Feb., 2003 A finding of very low safety significance involving a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," was self-revealing when the "B" emergency diesel generator failed to start on February 26, 2003, during a daily Technical Specification-required test, in response to the "A" emergency diesel generator being out of service for regularly scheduled 18-month periodic maintenance. The generator failed to start due to a pair of electrically open contacts on a cranking cutout relay which prevented energizing of the engine start relay. The cranking cutout relay had been installed during a design change completed in 1998, and the performance ratings of the new relay did not match original design specifications.
The inspectors considered this issue of more than minor significance, because if left uncorrected, the issue could become a more significant safety concern. In addition, the inspectors concluded that the finding was greater than minor because the finding involved the design control attribute of the mitigating systems cornerstone and affected the mitigating systems objective of ensuring the capability of the diesel generators in response to initiating events to prevent undesirable consequences. Specifically, the temporary design change failed to consider inductive electrical loads across the relay contacts, for which the relays were not rated. The inspectors evaluated the finding using the Significance Determination Process Phase 1 screening and determined that the finding was a design or qualification deficiency confirmed not to result in loss of function per Generic Letter 91-18; therefore, the finding was determined to be of very low safety significance.
Inspection Report# : 2003008(pdf)
Significance:          Dec 12, 2003 Identified By: NRC Item Type: NCV NonCited Violation 10 CFR 50, Appendix B, Criterion XVI NCV for ineffective corrective actions taken to address the implementation of the Boric Acid Leakage Inspection and Tracking Program The team identified a finding of very low significance associated with a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," for the failure to assure that actions were promptly taken to correct deficiencies in the implementation of the boric acid leakage inspection and tracking program for boric acid residue on safety-related components, a condition adverse to quality. Since 2001, approximately 12 condition reports had been initiated concerning the adequacy of the implementation of the licensee's boric acid leakage inspection and tracking program. During the inspection, the team identified approximately 14 safety-related components with various degrees of boric acid, which the licensee had not identified and evaluated in accordance with the boric acid leakage inspection and tracking program.
The team concluded that the licensee's failure to correct previous issues associated with the implementation of the boric acid leak log on safety-related components was greater than minor because if left uncorrected, the issue could become a more significant safety concern. The team evaluated the finding utilizing Inspection Manual Chapter 0609, "Significance Determination Process," Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," Phase 1 screening and determined the finding was of very low significance.
Inspection Report# : 2003010(pdf)
Significance:          Oct 21, 2003 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PERFORM TWO REQUIRED MEDICAL TESTS IN ACCORDANCE WITH 10 CFR 55.21 AND 55.23.
Green. The inspector identified a Non-Cited Violation of 10 CFR 55.21, "Medical Examination," and 10 CFR 55.23, "Certification." The inspector identified that the facility licensee failed to conduct all the medical testing required by American Nuclear Standards Institute/American Nuclear Society (ANSI/ANS) 3.4-1983, "Medical Certification and Monitoring of Personnel Requiring Operator Licenses for Nuclear Power Plants," as committed to by the facility licensee. Specifically, the facility licensee was not testing its operators for nose sensitivity (i.e., ability to detect odor of products of combustion and of tracer or market gases) Section 5.4.2, "Nose," and neurological testing, (i.e., normal central and peripheral nervous system function), including tactile discrimination (Stereognosis) sufficient to distinguish among various shapes of control knobs and hadles by touch, Section 5.4.14, "Neurological."
Inspection Report# : 2003005(pdf)
Significance: SL-IV Oct 21, 2003
 
3Q/2004 Inspection Findings - Kewaunee                                                                                                  Page 7 of 9 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PROVIDE ACCURATE INFORMATION TO THE NRC CONCERNING LICENSED OPERATOR MEDICAL REQUIREMENTS PER NRC FORM 396.
Severity Level IV. The inspector identified a Level IV Non-Cited Violation of 10 CFR 50.9, "Completeness and Accuracy of Information." The inspector identified that the facility licensee, between January 2, 2000, thorugh August 26, 2002, submitted to the NRC, NRC Forms 396 for 13 individuals applying for an initial operator's license and 18 licensed operators applying for renewal of their operator licenses, that were not accurate in all material respects. Specifically, the NRC Forms 396 certified that each applicant and licensed operator met the medical requirements of ANSI/ANS 3.4-1983. In fact, all the applicants and licensed operators were not adequately examined for all medical tests as required to meet the minimum standards of ANSI/ANS 3.4-1983.
Inspection Report# : 2003005(pdf)
Significance:        Oct 21, 2003 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO CONDUCT SIMULATOR PERFORMANCE TESTING THROUGHOUT THE LIFE OF THE SIMULATOR.
Green. The inspector identified a Non-Cited Violation of 10 CFR 55.46, "Simulation Facility." The inspector identified that the facility licensee failed to adequately conduct simulator performance testing throughout the life of the simulator. In addition, the facility licensee failed to correct modeling and hardware discrepancies and discrepancies identified from scenario validation and from performance testing. In addition, the facility licensee was committed to follow ANSI/ANS 3.5-1985, "Nuclear Power Plant Simulators for Use in Operator Training," as the way they would meet 10 CFR 55.46. Specifically, the licensee failed to conduct performance testing, with regard to normal evolutions core performance tests for Cycle 25, the most recent core load in the actual reactor. The licensee could only provide Cycle 7 normal evolutions core performance tests. No core performance tests had ever been conducted for Cycles 8 through 25, a period of 17 cycles.
Inspection Report# : 2003005(pdf)
Significance: SL-IV Oct 21, 2003 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO PROVIDE ACCURATE INFORMATION TO THE NRC CONCERNING ELIGIBILITY REQUIREMENTS FOR OPERATOR LICENSE APPLICATION PER NRC FORM 398.
Severity Level IV. The inspector identified a Level IV Non-Cited Violation of 10 CFR 50.9, "Completeness and Accuracy of Information." The inspector identified that on or about August 13, 2002, a senior facility licensee representative submitted to the NRC, NRC Forms 398 for three individuals, each applying for an initial operator's license, that were not accurate in all material respects. The facility licensee provided inaccurate information by certifying on the NRC Form 398 that the initial operator license applicaitons for three individuals had appropriately met the minimum training requirements for reactivity manipulations on the refrenced facility simulator in accordance with 10 CFR 55.31(a)(5) and 10 CFR 55.46(c)(2).
Inspection Report# : 2003005(pdf)
Barrier Integrity Significance:        Sep 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings." Failure to Implement Procedures for Work on Safety-Related Equipment A finding of very low safety significance was identified by the inspectors for a violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, And Drawings." The licensee conducted corrective maintenance to fix a deficient condition on the containment personnel hatch seal, a safety-related component, under the toolpouch maintenance' process rather than with the use of a work request or a work order, contrary to procedural requirements. The primary cause of this finding was related to the cross-cutting area of human performance.
Licensee personnel failed to appropriately implement licensee procedures for conducting work on safety-related components. Once this was identified, the licensee performed an extent of condition evaluation on the work control process and identified that, since July 2002, approximately 14 percent of the work performed under toolpouch maintenance' had been performed on safety-related components without a work order. The licensee also implemented a number of corrective actions to ensure work on safety-related equipment is conducted according to procedural requirements.
This issue was more than minor because it affected the Barrier Integrity Cornerstone attribute of reactor containment integrity, and, if left uncorrected, the finding could become a more significant safety concern. The finding was of very low safety significance because it did not represent an actual open pathway in the physical integrity of the reactor containment and none of the work conducted on safety-related equipment without a work order resulted in an operability concern. This issue was a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, And Drawings."
Inspection Report# : 2004007(pdf)
 
3Q/2004 Inspection Findings - Kewaunee                                                                                              Page 8 of 9 Significance:        Jun 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct historical residual heat removal pump mechanical seal leakage; 10CFR Part 50, Appendix B, Criterion XVI, "Corrective Action" A finding of very low safety significance associated with a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Actions", was self-revealed on June 16, 2004, when licensee personnel discovered leakage from the B' residual heat removal (RHR) pump seal when the pump was stopped following the performance of a surveillance procedure on the B' RHR train. Plant personnel determined the leakage to be in excess of that specified in the plant's System Integrity Program for leakage from emergency core cooling systems outside containment. The leakage was also in excess of the amount of leakage assumed in the Updated Safety Analysis Report, Chapter 14, for calculation of control room habitability doses and offsite exposures. The inspectors subsequently determined, from interviews with licensee personnel and a review of the licensee's corrective action program and work order history, that excessive RHR seal leakage has occurred since the late 1980s. However, past corrective actions have not been effective to correct this condition adverse to quality.
The licensee performed a prompt engineering review to ensure that no immediate catastrophic failure mechanism for the RHR seal existed. The licensee also performed a prompt engineering review of the impact of the estimated leakage on the control room habitability doses, as well as the offsite doses, and determined no exposure limits would be exceeded. The licensee took actions to immediately stop the leakage and plans to replace the RHR pump seal during the next refueling outage.
This self-revealed finding was more than minor because the finding affected the cornerstone objective of Reactor Safety/Barrier Integrity. The inspectors evaluated the finding using Inspection Manual Chapter 0609, "Significance Determination Process," Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," Phase 1 screening, and determined that the finding was of very low safety significance.
Inspection Report# : 2004004(pdf)
Emergency Preparedness Significance:        Dec 12, 2003 Identified By: NRC Item Type: FIN Finding Failure to take timely corrective actions to prevent recurrence for a 2001 white finding associated with Emergency Response Organization Augmentation The team identified a Green finding for the failure to take timely corrective actions to prevent recurrence for a White Finding initially identified in September 2000, associated with Emergency Response Organization Augmentation. While the team determined that corrective actions to date have been effective, as evidenced by only one augmentation drill failure since 2001, three of the eight corrective actions had not been completed.
The team determined that this issue was more than minor because if left uncorrected, the issue could become a more significant safety concern.
In addition, the team concluded that the issue affected the emergency preparedness cornerstone performance attribute associated with the emergency response organization augmentation system and emergency response augmentation testing and the objective of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The team evaluated the finding utilizing Inspection Manual Chapter 0609, Appendix B, "Emergency Preparedness Significance Determination Process," Section 5.0, "Corrective Actions," dated March 6, 2003, and determined the finding was of very low significance.
Inspection Report# : 2003010(pdf)
Significance:        Dec 12, 2003 Identified By: NRC Item Type: NCV NonCited Violation 10 CFR50.54 and emergency plan NCV for ineffective corrective actions in 2002 which resulted in the failure to make timely notifications for an actual unusual event in February 2003 The team identified a finding of very low significance associated with a Non-Cited Violation of 10 CFR Part 50.54(q) and the licensee's Emergency Plan for the failure to notify the state and local governmental agencies within 15 minutes after the declaration of an actual Unusual Event on February 26, 2003. The team concluded this failure was caused by the licensee's ineffective corrective actions for previously identified weaknesses and problems in the area of Emergency Preparedness.
The team determined that this issue was more than minor because this was an actual event implementation problem and affected the emergency preparedness cornerstone objective of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The team evaluated the finding utilizing Inspection Manual Chapter 0609, Appendix B, "Emergency Preparedness Significance Determination Process," Emergency Preparedness Significance Determination Process Sheet 2, dated March 6, 2003, and determined the finding was of very low significance.
 
3Q/2004 Inspection Findings - Kewaunee                  Page 9 of 9 Inspection Report# : 2003010(pdf)
Occupational Radiation Safety Public Radiation Safety Physical Protection Physical Protection information not publicly available.
Miscellaneous Last modified : December 29, 2004
 
4Q/2004 Inspection Findings - Kewaunee                                                                                                Page 1 of 6 Kewaunee 4Q/2004 Plant Inspection Findings Initiating Events Mitigating Systems Significance:        Sep 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings." Failure to Have Procedures Appropriate to the Circumstances for Preventive Maintenance of the TDAFW Pump Turbine A finding of very low safety significance was self-revealed during the licensee's review of high oil particulate in the Turbine Driven Auxiliary Feedwater Pump Turbine, which resulted in a violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings."
The licensee determined that high oil particulate in the Turbine Driven Auxiliary Feedwater Pump Turbine was caused by damage to the journal bearing. Maintenance procedures did not specify appropriate acceptance criteria for oil sampling, did not specify an appropriate inspection frequency and criteria for the turbine bearings and bearing cavities, and allowed the reuse of bearings in different locations during maintenance of the Turbine, which were not acceptable maintenance practices. The reuse of the upper inboard bearing in a different location contributed to the journal bearing damage. The licensee took immediate remedial corrective actions to replace the bearings, clean the housing and return the pump to service. In addition, the licensee revised its maintenance procedures to include appropriate instructions for turbine and pump maintenance activities.
This self-revealed finding was more than minor because, if left uncorrected, the issue would have become a more significant safety concern. In addition, it affected the Mitigating Systems attributes of equipment performance reliability and the Mitigating Systems cornerstone objective of ensuring the reliability of systems. The finding was of very low safety significance because it was not a design or qualification deficiency that has been confirmed to result in a loss of function per Generic Letter 91-18. This issue was a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, And Drawings."
Inspection Report# : 2004007(pdf)
Significance:        Sep 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action." Failure to Promptly Correct Conditions Adverse to Quality, Specifically Associated with Degraded and Nonconforming Conditions A finding of very low safety significance was identified by the inspectors for a violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Actions." During a review of the licensee's list of safety-related equipment designated as degraded or nonconforming, the inspectors identified that the licensee failed to promptly correct three conditions adverse to quality. These conditions adverse to quality included noncompliance of both Residual Heat Removal pump seal coolers with system design requirements, which was previously identified by NRC inspectors in November 2002, but not promptly corrected by the licensee; and two sections of safety-related piping, one associated with the "B" Emergency Diesel Generator fuel oil supply and the other associated with the Component Cooling Water piping from the "B" Residual Heat Removal pump seal cooler and stuffing box, that were identified by the licensee in September and April 2003, respectively, as exceeding Updated Safety Analysis Report stress criteria but not promptly corrected by the licensee. The primary cause of this finding was related to the cross-cutting area of problem identification and resolution. The licensee failed to prioritize and promptly correct these conditions adverse to quality in accordance with the guidelines in the corrective action program. Once these conditions were identified, the licensee restored the following conditions to operable: the A' RHR Pump Seal Cooler; the CCW piping expansion loop from the B' RHR pump seal cooler; and the fuel oil supply piping to the B' EDG. The licensee planned to restore the B' RHR Pump Seal Cooler during the upcoming Fall 2004 Refueling Outage.
This issue was more than minor because it affected the Mitigating System cornerstone attribute of design control for initial design and plant modifications and affected the associated cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was of very low safety significance because it was not a design or qualification deficiency that has been confirmed to result in a loss of function per Generic Letter 91-18. This issue was a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Actions."
Inspection Report# : 2004007(pdf)
 
4Q/2004 Inspection Findings - Kewaunee                                                                                                  Page 2 of 6 Significance:        Sep 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings." Failure to Have Acceptance Criteria for Flushing of the 1ARHR Fan Coil Unit A finding of very low safety significance was identified by the inspectors for a violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, And Drawings." This finding was associated with the licensee's failure to implement an appropriate inspection and cleaning procedure containing quantitative or qualitative acceptance criteria for the 1A RHR pump pit Fan Coil Unit to ensure that cleaning was satisfactorily accomplished. Following discovery, the licensee entered the issue into its corrective action program and conducted an immediate operability assessment that determined the involved fan coil units were operable.
This issue was more than minor because it involved the procedure quality 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. The finding was of very low safety significance because it was not a design or qualification deficiency that has been confirmed to result in a loss of function per Generic Letter 91-18. This issue was a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, And Drawings."
Inspection Report# : 2004007(pdf)
Significance:        Sep 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation 10 CFR 50, Appendix B, Criterion III, "Design Control." Failure to Verify the Acceptability of a Single Failure Vulnerability Introduced During a System Modification 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." This finding was associated with the licensee's failure to perform a design verification to demonstrate that the diesel generator lube oil cooler service water outlet valve actuators, installed under Design Change 3357, would not result in a failure of the valve stems under conditions in which the valve ball froze nor had the licensee provided sufficient justification to show that valve ball freezing was not credible.
Following discovery, the licensee entered the issue into its corrective action program and performed an operability assessment which provided additional justification to demonstrate that the stem failure was considered not credible.
This issue was more than minor because it involved 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. The finding was of very low safety significance because it was not a design or qualification deficiency that has been confirmed to result in a loss of function per Generic Letter 91-18. This issue was a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control."
Inspection Report# : 2004007(pdf)
Significance:        Jul 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Repair the Deluge System Heat Detectors in a Timely Manner The team identified a Non-Cited Violation of a license condition for fire protection. The licensee failed to take timely corrective actions to repair several maintenance storage area deluge system rate-of-rise heat detectors which were inoperable for an extended period of time. At the time of this inspection, the detectors had been repaired and returned to operability.
The finding was greater than minor because it affected the mitigating systems cornerstone attribute of protection against external factors (fire).
Specially, a partially inoperable deluge system can increase the likelihood of a fire which could challenge safe shutdown. The finding was of very low safety significance because this fire area has Pyr-A-Larm ionization detectors located at the ceiling level. These detectors would alarm in the control room and the fire brigade would respond to a fire in this area. In addition, other defense-in-depth fire protection elements remained unaffected and fire in this area would not result in a loss of dedicated safe shutdown systems.
Inspection Report# : 2004005(pdf)
Significance:        Jul 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Acceptable (Quality Related) Pre-Fire Strategies A finding of very low safety significance was identified by the team for a violation of a license condition for fire protection. The licensee failed to include pertinent information in their fire strategies. Specifically, the licensee failed to include information about the potential unavailability of certain fire hose stations and identify hydrogen and propane piping hazards in a fire zone. Once the issues were identified, the licensee entered the issue into their corrective action program and planned to revise their fire strategies to include the pertinent information.
 
4Q/2004 Inspection Findings - Kewaunee                                                                                                  Page 3 of 6 The issue was greater than minor because the failure to include pertinent information relating to the water supply used for manual fire fighting and hazards associated with hydrogen and propane piping in fire strategies could adversely impact fire fighting strategies used by the fire brigade in fighting a fire. The issue was of very low safety significance because of the extensive training provided to fire brigade members to deal with unexpected contingencies. The issue was a Non-Cited Violation of License Condition 2.C(3) which required, in part, that fire area strategies provide the fire brigade pertinent information on a given plan area to help the brigade to be better prepared for fire fighting within that area.
Inspection Report# : 2004005(pdf)
Significance:        Jul 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Meet the Fire Protection Program Requirements for Hose Lengths to Maintain an Acceptable Water Pressure and Flow at Hose Stations The team identified a Non-Cited Violation of License Condition 2.C(3), which requires the licensee to implement all provisions of their NRC approved fire protection program. The licensee failed to meet the fire protection program requirements for hose lengths to maintain an acceptable water pressure and flow to hose stations. The licensee's corrective actions included replacing hoses to increase water flow at hose stations The finding was greater than minor because it affected the mitigating systems cornerstone attribute of protection against external factors (fire).
Specifically, the failure to maintain acceptable water pressure and water flow to hose stations can hamper the brigade's ability to fight a fire, thereby, potentially endangering mitigating systems. The finding was of very low safety significance because the problem only impacts the effectiveness of the fire brigade while other fire protection features, such as fire barriers and physical separation remain available.
Inspection Report# : 2004005(pdf)
Significance:        Jul 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Meet the NFPA Code Requirements for Extinguisher Placement The team identified a Non-Cited Violation of License Condition 2.C.(3), which requires the licensee to implement all provisions of their approved fire protection program. Amendment No. 23 to Facility Operating License Safety Evaluation Report dated December 12, 1978, required fire extinguishers in accordance with the National Fire Protection Association Code. The licensee failed to meet the Code requirements for extinguisher placement in Fire Area AX-32. Once identified, the licensee initiated corrective actions to meet the Code requirements.
The finding was greater than minor because it affected the mitigating systems cornerstone attribute of protection against external factors (fire).
Specially, not having an extinguisher to put out a small fire can increase the likelihood of a fire which could challenge safe shutdown. The finding was of very low safety significance because this fire area has fire detectors that would alarm in the control room and the fire brigade would respond to a fire in this area. In addition, other defense-in-depth fire protection elements remained unaffected and fire in this area would not result in a loss of dedicated safe shutdown systems.
Inspection Report# : 2004005(pdf)
Significance:        Jul 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Transient Combustibles Not Adequately Controlled Within Fire Area AX-32 The team identified a Non-Cited Violation of License Condition 2.C(3) for failure to adequately control transient combustibles in fire area AX-
: 32. Specifically, authorization for the storage and use of combustibles in safety-related areas was not obtained. Once uncontrolled transient combustibles were identified, the materials were either included in the transient combustible permit system or removed from the area.
The issue was greater than minor because the failure to adequately control combustible materials could result in a more significant safety issue.
Uncontrolled combustibles could result in the greater likelihood or severity of a fire which affects equipment important to safety. The finding was of very low safety significance because of mitigation capability available in the event of a fire in fire area AX-32.
Inspection Report# : 2004005(pdf)
Significance:        Jul 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Transient Combustibles Not Adequately Controlled Within Fire Area AX-24 The team identified a Non-Cited Violation of License Condition 2.C(3), in that a hazardous quantity of transient combustibles was present in fire area AX-24. The hazardous quantity of transient combustibles present exceeded the quantity of combustibles allowed with no fire detection systems in this fire area.
 
4Q/2004 Inspection Findings - Kewaunee                                                                                                Page 4 of 6 The finding was greater than minor because it affected the mitigating systems cornerstone attribute of protection against external factors (fire).
Specifically, the presence of transient combustibles beyond what was approved by the NRC could result in the increased likelihood of a fire which could challenge safe shutdown. The finding was of very low safety significance because a fire from the observed transient combustibles would not result in a loss of the alternate shutdown systems.
Inspection Report# : 2004005(pdf)
Significance:        Jun 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify and Correct Issues Associated with Historical Safety Injection Lube Oil Cooler Fouling; 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action" A finding of very low safety significance associated with a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was self-revealed on January 15, 2004, when licensee inspection of the A' and B' safety injection pump lube oil coolers identified silt and lake grass accumulation at the tube pass inlets. Significant fouling of the safety injection pump lube oil coolers with lake grass had been identified by the licensee as early as 1992 when the coolers were first opened and inspected. The licensee failed to enter the results of those inspections in the corrective action program when fouling was identified, until 2001. When the issue was entered into the corrective action program in 2001, following an inspection by plant personnel, the associated evaluation did not adequately address the issue and corrective actions were not taken in a timely manner to address the issue.
The licensee initiated numerous corrective actions to address the root and contributing causes identified during the root cause evaluation of this event. Some of those actions included: replacing the old safety injection pump lube oil coolers with coolers of a new design; performing an extent of condition review of other service water systems prior to plant restart in January 2004 to ensure no similar immediate issues existed; sharing lessons learned from this event with all plant staff; and performing a prioritization review of all outstanding plant design modifications.
The inspectors verified the licensee's past operability analysis for the safety injection pumps. The inspectors evaluated the finding using the results of that analysis and Inspection Manual Chapter 0609, "Significance Determination Process," Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," Phase 1 screening, and determined that the finding was of very low safety significance.
Inspection Report# : 2004004(pdf)
Significance:        Jun 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to have Procedures Appropriate to the Circumstances, Including Appropriate Acceptance Criteria for Implementation of the Generic Letter 89-13 Program with Respect to the Safety Injection Lube A finding of very low safety significance associated with a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was self-revealed when the licensee discovered fouling of the safety injection pump lube oil coolers in January 2004. The licensee determined that evidence of the fouling had been present since the first inspection of the coolers in 1992. The licensee performed that first inspection as part of its actions to comply with Generic Letter 89-13, "Service Water System Problems Affecting Safety-Related Equipment." However, no acceptance criteria were included in the licensee's procedures developed to implement the commitments of Generic Letter 89-13 for these coolers to ensure that this activity had been satisfactorily accomplished.
The licensee initiated several corrective actions to address this issue, some of which included: establishing appropriate acceptance criteria for the safety injection lube oil coolers; developing a recovery plan for the licensee's Generic Letter 89-13 program and categorizing the program health in a red status; designating a single program owner to the Generic Letter 89-13 program; and reviewing other procedures utilized to implement the licensee's Generic Letter 89-13 program to verify specific acceptance criteria are contained in the procedures.
The inspectors verified the licensee's past operability analysis for the safety injection pumps. The inspectors evaluated the finding using the results of that analysis and Inspection Manual Chapter 0609, "Significance Determination Process," Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," Phase 1 screening, and determined that the finding was of very low safety significance.
Inspection Report# : 2004004(pdf)
Significance:        Mar 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation FAILURE TO NOTIFY THE NRC OF A CHANGE IN OPERATOR STATUS IN ACCORDANCE WITH 10 CFR 50.74(c)
The inspector identified a violation of 10 CFR 50.74(c), "Notification of Change in Operator or Senior Operator Status." The inspector identified that the facility licensee failed to notify the NRC within 30 days after receiving a change in medical status of a licensed operator from the station's medical examiner. The change in medical status required conditioning of the operator's license by the NRC.
Inspection Report# : 2004002(pdf)
 
4Q/2004 Inspection Findings - Kewaunee                                                                                                Page 5 of 6 Significance:        Mar 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to have procedures appropriate to the circumstances, including appropriate acceptance criteria for determining important activities have been satisfactorily accomplished for inservice inspect A finding of very low safety significance associated with a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was self-revealed when, on December 10, 2003, licensee personnel discovered evidence of component cooling water (CCW) system leakage from a radiation detector housing integral to the CCW piping. The leakage was determined to be evidence of a through wall leak of an American Society of Mechanical Engineers (ASME) Section XI Class 3 pipe which rendered both trains of CCW inoperable. Therefore, on December 12, 2003, operators declared both trains of the CCW system inoperable, due to the small CCW leak on the CCW radiation detector housing. The licensee subsequently determined that evidence of the leakage had been present for the past 13 years.
However, less than adequate procedure acceptance criteria resulted in licensee personnel classifying the leakage during inservice inspections as a non-recordable' indication instead of the required recordable' indication for this type leakage.
The licensee took immediate corrective actions to move the ASME Section XI Class 3 piping boundary to the radiation detector housing. In addition, the licensee implemented corrective actions to prevent recurrence which included: revision of 19 inservice inspection surveillance procedures to incorporate appropriate acceptance criteria; and discussions of the root cause with licensee inservice inspection personnel as a lessons-learned.
This self-revealed finding was greater than minor because, if left uncorrected, the finding would become a more significant safety concern. The inspectors evaluated the finding using the Significance Determination Process, Appendix A, Phase 1 Screening, and determined that the finding was of very low safety significance.
Inspection Report# : 2004002(pdf)
Significance:        Jan 28, 2004 Identified By: NRC Item Type: FIN Finding Failure to appropriately evaluate for potential bypqss flow on service water strainers The inspectors identified a finding of very low safety significance associated with the licensee's failure to appropriately evaluate for potential bypass flow on the service water pump discharge strainers by measuring a critical gap dimension at the bottom of the basket-to-housing interface. This finding did not constitute a violation of NRC requirements because the strainers (aside from the pressure boundary) did not fulfill a safety-related function.
The inspectors determined that the finding was of more than minor significance because it would become a more significant safety concern if left uncorrected. Specifically, the failure to appropriately evaluate for potential bypass flow on the service water pump discharge strainers could reasonably result in debris fouling of service water cooled components and degraded or inoperable safety-related equipment. The inspectors concluded that this finding was a licensee performance deficiency of very low safety significance because it did not result in loss of safety function for a service water system train for greater than its Technical Specification allowed outage time. To address this issue, the licensee opened each strainer and measured the gap at the bottom of the basket-to-housing interface.
Inspection Report# : 2004003(pdf)
Barrier Integrity Significance:        Sep 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings." Failure to Implement Procedures for Work on Safety-Related Equipment A finding of very low safety significance was identified by the inspectors for a violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, And Drawings." The licensee conducted corrective maintenance to fix a deficient condition on the containment personnel hatch seal, a safety-related component, under the toolpouch maintenance' process rather than with the use of a work request or a work order, contrary to procedural requirements. The primary cause of this finding was related to the cross-cutting area of human performance.
Licensee personnel failed to appropriately implement licensee procedures for conducting work on safety-related components. Once this was identified, the licensee performed an extent of condition evaluation on the work control process and identified that, since July 2002, approximately 14 percent of the work performed under toolpouch maintenance' had been performed on safety-related components without a work order. The licensee also implemented a number of corrective actions to ensure work on safety-related equipment is conducted according to procedural requirements.
This issue was more than minor because it affected the Barrier Integrity Cornerstone attribute of reactor containment integrity, and, if left uncorrected, the finding could become a more significant safety concern. The finding was of very low safety significance because it did not
 
4Q/2004 Inspection Findings - Kewaunee                                                                                            Page 6 of 6 represent an actual open pathway in the physical integrity of the reactor containment and none of the work conducted on safety-related equipment without a work order resulted in an operability concern. This issue was a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, And Drawings."
Inspection Report# : 2004007(pdf)
Significance:        Jun 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct historical residual heat removal pump mechanical seal leakage; 10CFR Part 50, Appendix B, Criterion XVI, "Corrective Action" A finding of very low safety significance associated with a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Actions", was self-revealed on June 16, 2004, when licensee personnel discovered leakage from the B' residual heat removal (RHR) pump seal when the pump was stopped following the performance of a surveillance procedure on the B' RHR train. Plant personnel determined the leakage to be in excess of that specified in the plant's System Integrity Program for leakage from emergency core cooling systems outside containment. The leakage was also in excess of the amount of leakage assumed in the Updated Safety Analysis Report, Chapter 14, for calculation of control room habitability doses and offsite exposures. The inspectors subsequently determined, from interviews with licensee personnel and a review of the licensee's corrective action program and work order history, that excessive RHR seal leakage has occurred since the late 1980s. However, past corrective actions have not been effective to correct this condition adverse to quality.
The licensee performed a prompt engineering review to ensure that no immediate catastrophic failure mechanism for the RHR seal existed. The licensee also performed a prompt engineering review of the impact of the estimated leakage on the control room habitability doses, as well as the offsite doses, and determined no exposure limits would be exceeded. The licensee took actions to immediately stop the leakage and plans to replace the RHR pump seal during the next refueling outage.
This self-revealed finding was more than minor because the finding affected the cornerstone objective of Reactor Safety/Barrier Integrity. The inspectors evaluated the finding using Inspection Manual Chapter 0609, "Significance Determination Process," Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," Phase 1 screening, and determined that the finding was of very low safety significance.
Inspection Report# : 2004004(pdf)
Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Physical Protection information not publicly available.
Miscellaneous Last modified : March 09, 2005
 
1Q/2005 Inspection Findings - Kewaunee                                                                                              Page 1 of 11 Kewaunee 1Q/2005 Plant Inspection Findings Initiating Events Significance:        Feb 18, 2005 Identified By: NRC Item Type: FIN Finding Safety Buses Relay Sensitivity to External Electrical Distubrances The team identified a finding of very low safety significance for a failure to provide adequate relay setpoint calibration tolerances on safety buses 1-5 and 1-6 loss of voltage relays. The existing relay setting calibration tolerances would have allowed the loss of voltage relays to actuate spuriously during certain offsite electrical system disturbances and un-necessarily separate the safety buses from the offsite power system and result in a plant transient. The licensee implemented corrective actions to revise the appropriate loss of voltage relay surveillance procedures.
The finding was more than minor because the failure to provide adequate relay setting tolerances could result in an unnecessary separation of the safety buses from the electrical grid and an ensuing plant transient. The finding was of very low safety significance because the issue would not preclude the safety buses from being re-energized by the emergency power sources. The finding was a not a violation of regulatory requirements.
Inspection Report# : 2005002(pdf)
Significance:        Feb 18, 2005 Identified By: NRC Item Type: NCV NonCited Violation Operator Actions Following Station Blackout - Lack of Procedure Guidance The team identified a Non-Cited Violation of 10 CFR 50.63, "Loss of All Alternating Current Power," for a failure to maintain procedural steps that minimized the likelihood and duration of a Station Blackout (SBO) event. The deleted procedural steps allowed for the cross-connection of the plant's two redundant safety buses should both the Reserve Auxiliary Transformer and the 1B Emergency Diesel Generator fail. These procedural steps, as originally employed, served to lessen the likelihood of the SBO occurring, and/or reduce the time of the SBO. The licensee implemented corrective actions to revise the appropriate operations procedure.
This finding was more than minor, because it was associated with the likelihood of an initiating event and the reliability of a safety bus that responds to an initiating event. The finding was of very low safety significance, because multiple sources of both onsite and offsite power remained available to supply the two safety buses.
Inspection Report# : 2005002(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Control Of Combustible Matrials A finding of very low safety significance was identified by the inspectors for a violation of a fire protection License Condition. The inspectors identified multiple examples of combustible materials either stored or in use without specific authorization. Specifically, the licensee stored and used lubricating oil in an emergency diesel generator room beyond that authorized by the Fire Protection Program Analysis, the licensee stored unauthorized combustible materials above the shelves in the working materials storage area and on top of cabinets nearby, and the licensee stored compressed flammable gas cylinders in the auxiliary building without authorization. Once these issues were identified, the licensee removed the unauthorized materials. This finding was related to the cross-cutting area of problem identification and resolution in that the NRC had previously identified issues relating to control of transient combustible materials above and near the working materials storage area but adequate corrective actions were not put in place to prevent recurrence of this isse.
The finding was more than minor because the failure to adequately control combustible materials, if left uncorrected, could become a more safety significant concern. The finding was of very low safety significance because the issue was a low degradation of fire prevention and administrative controls. The finding was a Non-Cited Violation of License Condition 2.C(3) which required specific authorization for the storage and use of combustibles in safety-related areas.
Inspection Report# : 2004009(pdf)
Significance:        Dec 31, 2004 Identified By: NRC
 
1Q/2005 Inspection Findings - Kewaunee                                                                                                Page 2 of 11 Item Type: NCV NonCited Violation Inadequate Corrective Action to Preclude Storage of Oxygen Cylinders Next to Flammable Gas Cylinders A finding of very low safety significance was identified by the inspectors for a violation of a fire protection License Condition. The inspectors identified the storage of compressed oxygen cylinders near compressed flammable gas cylinders. Once this issue was identified, the licensee removed the stored compressed oxygen cylinders from the area.
The finding was more than minor because the inappropriate storage of compressed oxygen cylinders could result in greater severity of a fire affecting equipment important to safety. The finding was of very low safety significance because the issue was a low degradation of fire prevention and administrative controls. The finding was a Non-Cited Violation of License Condition 2.C(3) which required the bulk storage of compressed oxygen cylinders to be separated from compressed flammable gas cylinders and corrective action of conditions significantly adverse to quality to preclude recurrence.
Inspection Report# : 2004009(pdf)
Mitigating Systems Significance:        Feb 18, 2005 Identified By: NRC Item Type: NCV NonCited Violation Lack of Allowance for Manual Actions in Establishing Setpoint to Transfer AFW Pump Suction Source The team identified a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for failure to establish the condensate storage tank (CST) level setpoint to transfer the auxiliary feedwater (AFW) pump suction supply from the CST to service water.
The team determined that the calculation setpoint did not include an allowance for the manual operator actions required by emergency operations procedures. The licensee revised the plant procedure to perform the operator actions earlier in the procedure.
This finding was more than minor because it affected the mitigating systems cornerstone objective of equipment reliability, in that failure to align the AFW pump suctions to service water prior to the CSTs being depleted could have resulted in damage to the AFW pumps. The finding was determined to be of very low safety significance because it was a design deficiency that did not result in a loss of function.
Inspection Report# : 2005002(pdf)
Significance:        Feb 18, 2005 Identified By: NRC Item Type: FIN Finding Lack of 4160 Vac Bus 1-5 Ovewrcurrent and Loss of Voltage Relay Coordination The team identified a finding of very low safety significance for a failure to provide adequate electrical coordination of protective devices thereby ensuring that postulated electrical faults would be isolated upon detection. Specifically, the team identified that the lack of adequate electrical systems coordination between the undervoltage and overcurrent protection on 4160 Vac safety bus 1-5 would result in the loss of voltage relays actuating before the bus over-current relays. This design deficiency results in the failure to lock out safety bus 1- 5 upon postulated electrical faults and subjects the postulated faulted safety bus 1-5 to be re-energized via an alternate offsite source. This design introduced a challenge to the safety equipment availability and reliability. The licensee planned to develop changes to the affected relays.
The finding was more than minor because the failure to provide adequate electrical coordination of electrical devices provided an unnecessary challenge to safety-related equipment, and if left uncorrected, could become a more safety significant concern. The finding was of very low safety significance because it was a design deficiency that did not result in the loss of system function. The finding was a not a violation of regulatory requirements.
Inspection Report# : 2005002(pdf)
Significance:        Feb 18, 2005 Identified By: NRC Item Type: NCV NonCited Violation Short Circuit Duty of Buses Exceeded - Impact on Safe Shutdown Analysis The team identified a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion II, "Quality Assurance Program," for a failure to identify potentially adverse conditions to the plant's fire protection safe shutdown analysis caused by known overduty conditions on non-safety related buses 1-1, 1-2, 1-3, and 1-4. While the overduty condition was known to have existed at least since 1992, the licensee never entered the issue into the plant's corrective action program, where a proper evaluation should have addressed 10 CFR Part 50, Appendix R, safe shutdown related effects. The licensee planned to continue efforts to identify additional evaluations and corrective actions.
This finding was more than minor, because it was associated with the degradation of a fire protection feature. The finding was of very low safety significance because after extensive evaluation of the deficiency, the licensee was able to determine that the plant could still safely shut down the plant during a postulated fire event.
Inspection Report# : 2005002(pdf)
 
1Q/2005 Inspection Findings - Kewaunee                                                                                                Page 3 of 11 Significance:        Feb 18, 2005 Identified By: NRC Item Type: NCV NonCited Violation Battery Sizing Deficiencies The team identified a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for failure to implement adequate design controls of documents, inputs, and assumptions in the design of the two safety-related batteries. Specifically, the licensee did not perform and control battery sizing calculations, including consideration of temperature effects, to ensure that the batteries maintained sufficient capacity to perform the intended design function. The team determined that the failure to appropriately evaluate effects of battery room and cell temperatures also affected the cross-cutting area of Problem Identification and Resolution because the subject of battery capacity versus battery temperature had been previously identified in a 1992 NRC inspection. The licensee planned to perform battery sizing calculations as part of an overall electrical systems analysis improvement project.
This finding was more than minor because it affected the mitigating systems cornerstone objective of ensuring the availability and reliability of the 125 Volts direct current battery system to respond to initiating events to prevent undesirable consequences. The finding is of very low safety significance because the battery remained operable. The licensee planned to develop formal battery sizing calculations.
Inspection Report# : 2005002(pdf)
Significance: SL-IV Feb 18, 2005 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Evaluation of Procedure Changes to Address AFW Design Deficiencies The team identified a finding involving a Non-Cited Violation of 10 CFR 50.59, "Changes, Tests, and Experiments." The finding involved a failure to perform an adequate review of operations procedure changes in accordance with 10 CFR 50.59 associated with the operation of motor-operated valves for the auxiliary feedwater suction source from the service water system. The team determined that the licensee's approval of changes to Procedure E-0-05, with the introduction of adverse effects, and a determination that 10 CFR 50.59 was not applicable was a violation of 10 CFR 50.59. The licensee subsequently performed additional evaluations of the procedure changes.
Because the issue affected the NRC's ability to perform its regulatory function, this finding was evaluated with the traditional enforcement process. The finding was determined to be of very low safety significance since the design basis safety-related function of the AFW system, to remove reactor decay heat following a loss of normal feedwater, was not adversely affected. This was determined to be a Severity Level IV NCV of 10 CFR 50.59.
Inspection Report# : 2005002(pdf)
Significance:        Feb 18, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Ensure that Calculation Assumption was Based on Valid Times for Manual Operator Actions The team identified a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control." The finding involved the condensate storage tank (CST) level setpoint to transfer the auxiliary feedwater (AFW) pump suction from the CSTs to service water. A calculation assumption stated that a flow would drain from the CSTs to the condenser for 10 minutes until the operators isolated the flow by closing manual valve MU-2A. The team determined that the actions could not be completed in the time assumed by the calculation. The licensee initiated corrective actions to revise the appropriate operations procedure and calculation.
This finding was greater than minor because it affected the mitigating system cornerstone objective of equipment reliability, in that failure to align the AFW pump suctions to service water prior to the CSTs being depleted could have resulted in damage to the AFW pumps. The finding was determined to be of very low safety significance because it was a design deficiency that was not found to result in a loss of function. The team concluded that it was unlikely that the operators would allow the CST level to reach the EOP setpoint without attempting to refill the tanks from other sources, and that the operators would be aware of the CST levels.
Inspection Report# : 2005002(pdf)
Significance:        Feb 18, 2005 Identified By: NRC Item Type: NCV NonCited Violation TSC DG Target Reliability Methodology Inadequate The team identified a Non-Cited Violation of 10 CFR 50.63, "Loss of All Alternating Current Power." The finding involved the failure to establish a target reliability for the plant's alternate power source consistent with the reliability approved by the NRC staff in the licensee's Station Blackout submittal for 10 CFR 50.63. The non-conservative target reliability employed by the licensee resulted in the failure of the licensee to increase efforts to restore the Technical Support Center (TSC) Diesel Generator (DG) to its approved target reliability at an earlier date. The licensee subsequently initiated a corrective action to change the TSC DG reliability methodology.
This finding was more than minor, because it affected the reliability of a support system required for the mitigation of an Station Blackout event. The finding was of very low safety significance, because the finding did not directly affect the immediate operability of the TSC DG.
Inspection Report# : 2005002(pdf)
 
1Q/2005 Inspection Findings - Kewaunee                                                                                                  Page 4 of 11 Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Scaffolding Erected too Close to Safety-Related Equipment Required to be Operable A finding of very low safety significance was identified by the inspectors for a violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings." The finding was associated with the licensee's failure to adequately implement scaffold control requirements contained in Procedure GMP-127, "Requirements and Guidelines for Scaffold Construction and Inspection," which required that scaffolding be no closer than 2 inches from any safety-related equipment unless otherwise evaluated and approved by Engineering. Specifically, scaffolding was erected within 2 inches of safety-related piping for the Service Water outlet from the jacket water heat exchangers for Diesel Generator B, the piping for the Emergency Borate MOV (CVC-440), and Safety Injection Pump A, without engineering evaluation and approval. Upon discovery of this condition, the licensee took immediate action to bring all noted scaffolding problems into compliance with licensee procedures and initiated a CAP document for the issue.
The finding was more than minor because, if left uncorrected, the issue may have resulted in a more significant safety concern. Specifically, the failure of scaffolding having adequate spacing in the vicinity of safety-related equipment during a seismic event could result in damage to mitigating equipment. The finding was of very low safety significance because it did not result in the actual loss of the safety function of the train or system. The finding was a Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings."
Inspection Report# : 2004009(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify Inadequate Pre-Fire Strategies A finding of very low safety significance was identified by the inspectors for a violation of a fire protection License Condition. The inspectors identified that the licensee failed to identify pertinent information, such as the presence of compressed flammable gas cylinders, on a fire area strategy for fire brigade personnel. Once this issue was identified, the licensee revised the fire area strategy for the affected area.
The finding was more than minor because the failure to provide adequate warnings and guidance relating to hazards associated with compressed flammable gas cylinders in fire strategies could adversely impact fire fighting strategies used by the fire brigade in fighting a fire.
The finding was of very low safety significance due to extensive training provided to fire brigade members to deal with unexpected contingencies. The finding was a Non-Cited Violation of License Condition 2.C(3) which required that fire area strategies provide pertinent information to help the fire brigade to be better prepared for fire fighting within that area.
Inspection Report# : 2004009(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Non-conforming Condition on the Safety-Related Containment Sump A finding of very low safety significance was identified by the inspectors for a violation of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Actions." The original licensing and design basis of the containment sump screens was to prevent any particles greater than 1/8 inch from entering the sump. The inspectors determined that the screen size was 1/8-inch by 15/32-inch which allowed particles greater than 1/8-inch to enter the sump. The inspectors subsequently determined that this issue had been identified and entered into the licensee's corrective action program in 1997. However, adequate corrective actions were not taken to correct this condition adverse to quality. Once this issue was identified, the licensee conducted an operability determination and concluded that there were no immediate operability issues with the containment sump. The licensee determined that the sump screens were nonconforming in accordance with Generic Letter 91-18, and planned long term corrective actions to be developed in conjunction with the resolution of Generic Safety Issue 191 and NRC Generic Letter 2004-02.
The inspectors concluded that the primary cause of this finding was related to the performance characteristic of corrective actions in the cross-cutting area of problem identification and resolution.
This finding was more than minor because the issue affected the Mitigating System cornerstone attribute of design control for initial design and equipment performance reliability and affected the associated cornerstone objective to ensure the reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was of very low safety significance because it was not a design or qualification deficiency that has been confirmed to result in a loss of function per Generic Letter 91-18. This finding was a Non-Cited Violation of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Actions."
Inspection Report# : 2004009(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Instructions and Procedures for Inspections and Cleaning of the Safety-Related Containment Sump A finding of very low safety significance was identified by the inspectors for a violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, And Drawings," regarding licensee instructions and procedures for containment sump inspections. Specifically, the inspectors
 
1Q/2005 Inspection Findings - Kewaunee                                                                                              Page 5 of 11 identified that current licensee procedures did not require inspection or cleaning when boric acid or small debris may be present in the containment sump. The licensee's procedures for containment coatings did not require inspection of the coating located inside the containment sump which had not been inspected since initial application; and the licensee's procedure for containment sump gap inspections did not specify acceptance criteria to ensure this activity was satisfactorily accomplished. The licensee subsequently initiated several corrective actions to address these issues which included, but are not limited to: immediate inspection and cleaning of the safety-related containment sump; immediate inspection and assessment of the safety-related sump concrete coating; revision of preventive maintenance activities to require inspection and cleaning of the safety-related containment sump every refueling outage; revision of procedures to include inspection of the safety-related containment sump concrete coating every refueling outage; and revision of procedures to include appropriate acceptance criteria for determining that important activities were satisfactorily accomplished.
This finding was more than minor because if left uncorrected the finding could become a more significant safety concern and the issue affected the Mitigating System cornerstone attributes of equipment performance reliability and procedure quality and affected the associated cornerstone objective to ensure the reliability and capability of systems that respond to initiating events to prevent undesirable consequences.
The finding was of very low safety significance because it was not a design or qualification deficiency that has been confirmed to result in a loss of function per Generic Letter 91-18. This finding was a Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings."
Inspection Report# : 2004009(pdf)
Significance:        Sep 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action." Failure to Promptly Correct Conditions Adverse to Quality, Specifically Associated with Degraded and Nonconforming Conditions A finding of very low safety significance was identified by the inspectors for a violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Actions." During a review of the licensee's list of safety-related equipment designated as degraded or nonconforming, the inspectors identified that the licensee failed to promptly correct three conditions adverse to quality. These conditions adverse to quality included noncompliance of both Residual Heat Removal pump seal coolers with system design requirements, which was previously identified by NRC inspectors in November 2002, but not promptly corrected by the licensee; and two sections of safety-related piping, one associated with the "B" Emergency Diesel Generator fuel oil supply and the other associated with the Component Cooling Water piping from the "B" Residual Heat Removal pump seal cooler and stuffing box, that were identified by the licensee in September and April 2003, respectively, as exceeding Updated Safety Analysis Report stress criteria but not promptly corrected by the licensee. The primary cause of this finding was related to the cross-cutting area of problem identification and resolution. The licensee failed to prioritize and promptly correct these conditions adverse to quality in accordance with the guidelines in the corrective action program. Once these conditions were identified, the licensee restored the following conditions to operable: the A' RHR Pump Seal Cooler; the CCW piping expansion loop from the B' RHR pump seal cooler; and the fuel oil supply piping to the B' EDG. The licensee planned to restore the B' RHR Pump Seal Cooler during the upcoming Fall 2004 Refueling Outage.
This issue was more than minor because it affected the Mitigating System cornerstone attribute of design control for initial design and plant modifications and affected the associated cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was of very low safety significance because it was not a design or qualification deficiency that has been confirmed to result in a loss of function per Generic Letter 91-18. This issue was a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Actions."
Inspection Report# : 2004007(pdf)
Significance:        Sep 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings." Failure to Have Procedures Appropriate to the Circumstances for Preventive Maintenance of the TDAFW Pump Turbine A finding of very low safety significance was self-revealed during the licensee's review of high oil particulate in the Turbine Driven Auxiliary Feedwater Pump Turbine, which resulted in a violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings."
The licensee determined that high oil particulate in the Turbine Driven Auxiliary Feedwater Pump Turbine was caused by damage to the journal bearing. Maintenance procedures did not specify appropriate acceptance criteria for oil sampling, did not specify an appropriate inspection frequency and criteria for the turbine bearings and bearing cavities, and allowed the reuse of bearings in different locations during maintenance of the Turbine, which were not acceptable maintenance practices. The reuse of the upper inboard bearing in a different location contributed to the journal bearing damage. The licensee took immediate remedial corrective actions to replace the bearings, clean the housing and return the pump to service. In addition, the licensee revised its maintenance procedures to include appropriate instructions for turbine and pump maintenance activities.
This self-revealed finding was more than minor because, if left uncorrected, the issue would have become a more significant safety concern. In addition, it affected the Mitigating Systems attributes of equipment performance reliability and the Mitigating Systems cornerstone objective of ensuring the reliability of systems. The finding was of very low safety significance because it was not a design or qualification deficiency that has been confirmed to result in a loss of function per Generic Letter 91-18. This issue was a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, And Drawings."
 
1Q/2005 Inspection Findings - Kewaunee                                                                                                  Page 6 of 11 Inspection Report# : 2004007(pdf)
Significance:        Sep 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings." Failure to Have Acceptance Criteria for Flushing of the 1ARHR Fan Coil Unit A finding of very low safety significance was identified by the inspectors for a violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, And Drawings." This finding was associated with the licensee's failure to implement an appropriate inspection and cleaning procedure containing quantitative or qualitative acceptance criteria for the 1A RHR pump pit Fan Coil Unit to ensure that cleaning was satisfactorily accomplished. Following discovery, the licensee entered the issue into its corrective action program and conducted an immediate operability assessment that determined the involved fan coil units were operable.
This issue was more than minor because it involved the procedure quality 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. The finding was of very low safety significance because it was not a design or qualification deficiency that has been confirmed to result in a loss of function per Generic Letter 91-18. This issue was a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, And Drawings."
Inspection Report# : 2004007(pdf)
Significance:        Sep 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation 10 CFR 50, Appendix B, Criterion III, "Design Control." Failure to Verify the Acceptability of a Single Failure Vulnerability Introduced During a System Modification 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." This finding was associated with the licensee's failure to perform a design verification to demonstrate that the diesel generator lube oil cooler service water outlet valve actuators, installed under Design Change 3357, would not result in a failure of the valve stems under conditions in which the valve ball froze nor had the licensee provided sufficient justification to show that valve ball freezing was not credible.
Following discovery, the licensee entered the issue into its corrective action program and performed an operability assessment which provided additional justification to demonstrate that the stem failure was considered not credible.
This issue was more than minor because it involved 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. The finding was of very low safety significance because it was not a design or qualification deficiency that has been confirmed to result in a loss of function per Generic Letter 91-18. This issue was a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control."
Inspection Report# : 2004007(pdf)
Significance:        Jul 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Repair the Deluge System Heat Detectors in a Timely Manner The team identified a Non-Cited Violation of a license condition for fire protection. The licensee failed to take timely corrective actions to repair several maintenance storage area deluge system rate-of-rise heat detectors which were inoperable for an extended period of time. At the time of this inspection, the detectors had been repaired and returned to operability.
The finding was greater than minor because it affected the mitigating systems cornerstone attribute of protection against external factors (fire).
Specially, a partially inoperable deluge system can increase the likelihood of a fire which could challenge safe shutdown. The finding was of very low safety significance because this fire area has Pyr-A-Larm ionization detectors located at the ceiling level. These detectors would alarm in the control room and the fire brigade would respond to a fire in this area. In addition, other defense-in-depth fire protection elements remained unaffected and fire in this area would not result in a loss of dedicated safe shutdown systems.
Inspection Report# : 2004005(pdf)
Significance:        Jul 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Acceptable (Quality Related) Pre-Fire Strategies A finding of very low safety significance was identified by the team for a violation of a license condition for fire protection. The licensee failed to include pertinent information in their fire strategies. Specifically, the licensee failed to include information about the potential unavailability
 
1Q/2005 Inspection Findings - Kewaunee                                                                                                Page 7 of 11 of certain fire hose stations and identify hydrogen and propane piping hazards in a fire zone. Once the issues were identified, the licensee entered the issue into their corrective action program and planned to revise their fire strategies to include the pertinent information.
The issue was greater than minor because the failure to include pertinent information relating to the water supply used for manual fire fighting and hazards associated with hydrogen and propane piping in fire strategies could adversely impact fire fighting strategies used by the fire brigade in fighting a fire. The issue was of very low safety significance because of the extensive training provided to fire brigade members to deal with unexpected contingencies. The issue was a Non-Cited Violation of License Condition 2.C(3) which required, in part, that fire area strategies provide the fire brigade pertinent information on a given plan area to help the brigade to be better prepared for fire fighting within that area.
Inspection Report# : 2004005(pdf)
Significance:        Jul 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Meet the Fire Protection Program Requirements for Hose Lengths to Maintain an Acceptable Water Pressure and Flow at Hose Stations The team identified a Non-Cited Violation of License Condition 2.C(3), which requires the licensee to implement all provisions of their NRC approved fire protection program. The licensee failed to meet the fire protection program requirements for hose lengths to maintain an acceptable water pressure and flow to hose stations. The licensee's corrective actions included replacing hoses to increase water flow at hose stations The finding was greater than minor because it affected the mitigating systems cornerstone attribute of protection against external factors (fire).
Specifically, the failure to maintain acceptable water pressure and water flow to hose stations can hamper the brigade's ability to fight a fire, thereby, potentially endangering mitigating systems. The finding was of very low safety significance because the problem only impacts the effectiveness of the fire brigade while other fire protection features, such as fire barriers and physical separation remain available.
Inspection Report# : 2004005(pdf)
Significance:        Jul 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Meet the NFPA Code Requirements for Extinguisher Placement The team identified a Non-Cited Violation of License Condition 2.C.(3), which requires the licensee to implement all provisions of their approved fire protection program. Amendment No. 23 to Facility Operating License Safety Evaluation Report dated December 12, 1978, required fire extinguishers in accordance with the National Fire Protection Association Code. The licensee failed to meet the Code requirements for extinguisher placement in Fire Area AX-32. Once identified, the licensee initiated corrective actions to meet the Code requirements.
The finding was greater than minor because it affected the mitigating systems cornerstone attribute of protection against external factors (fire).
Specially, not having an extinguisher to put out a small fire can increase the likelihood of a fire which could challenge safe shutdown. The finding was of very low safety significance because this fire area has fire detectors that would alarm in the control room and the fire brigade would respond to a fire in this area. In addition, other defense-in-depth fire protection elements remained unaffected and fire in this area would not result in a loss of dedicated safe shutdown systems.
Inspection Report# : 2004005(pdf)
Significance:        Jul 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Transient Combustibles Not Adequately Controlled Within Fire Area AX-32 The team identified a Non-Cited Violation of License Condition 2.C(3) for failure to adequately control transient combustibles in fire area AX-
: 32. Specifically, authorization for the storage and use of combustibles in safety-related areas was not obtained. Once uncontrolled transient combustibles were identified, the materials were either included in the transient combustible permit system or removed from the area.
The issue was greater than minor because the failure to adequately control combustible materials could result in a more significant safety issue.
Uncontrolled combustibles could result in the greater likelihood or severity of a fire which affects equipment important to safety. The finding was of very low safety significance because of mitigation capability available in the event of a fire in fire area AX-32.
Inspection Report# : 2004005(pdf)
Significance:        Jul 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Transient Combustibles Not Adequately Controlled Within Fire Area AX-24
 
1Q/2005 Inspection Findings - Kewaunee                                                                                              Page 8 of 11 The team identified a Non-Cited Violation of License Condition 2.C(3), in that a hazardous quantity of transient combustibles was present in fire area AX-24. The hazardous quantity of transient combustibles present exceeded the quantity of combustibles allowed with no fire detection systems in this fire area.
The finding was greater than minor because it affected the mitigating systems cornerstone attribute of protection against external factors (fire).
Specifically, the presence of transient combustibles beyond what was approved by the NRC could result in the increased likelihood of a fire which could challenge safe shutdown. The finding was of very low safety significance because a fire from the observed transient combustibles would not result in a loss of the alternate shutdown systems.
Inspection Report# : 2004005(pdf)
Significance:        Jun 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify and Correct Issues Associated with Historical Safety Injection Lube Oil Cooler Fouling; 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action" A finding of very low safety significance associated with a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was self-revealed on January 15, 2004, when licensee inspection of the A' and B' safety injection pump lube oil coolers identified silt and lake grass accumulation at the tube pass inlets. Significant fouling of the safety injection pump lube oil coolers with lake grass had been identified by the licensee as early as 1992 when the coolers were first opened and inspected. The licensee failed to enter the results of those inspections in the corrective action program when fouling was identified, until 2001. When the issue was entered into the corrective action program in 2001, following an inspection by plant personnel, the associated evaluation did not adequately address the issue and corrective actions were not taken in a timely manner to address the issue.
The licensee initiated numerous corrective actions to address the root and contributing causes identified during the root cause evaluation of this event. Some of those actions included: replacing the old safety injection pump lube oil coolers with coolers of a new design; performing an extent of condition review of other service water systems prior to plant restart in January 2004 to ensure no similar immediate issues existed; sharing lessons learned from this event with all plant staff; and performing a prioritization review of all outstanding plant design modifications.
The inspectors verified the licensee's past operability analysis for the safety injection pumps. The inspectors evaluated the finding using the results of that analysis and Inspection Manual Chapter 0609, "Significance Determination Process," Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," Phase 1 screening, and determined that the finding was of very low safety significance.
Inspection Report# : 2004004(pdf)
Significance:        Jun 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to have Procedures Appropriate to the Circumstances, Including Appropriate Acceptance Criteria for Implementation of the Generic Letter 89-13 Program with Respect to the Safety Injection Lube A finding of very low safety significance associated with a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was self-revealed when the licensee discovered fouling of the safety injection pump lube oil coolers in January 2004. The licensee determined that evidence of the fouling had been present since the first inspection of the coolers in 1992. The licensee performed that first inspection as part of its actions to comply with Generic Letter 89-13, "Service Water System Problems Affecting Safety-Related Equipment." However, no acceptance criteria were included in the licensee's procedures developed to implement the commitments of Generic Letter 89-13 for these coolers to ensure that this activity had been satisfactorily accomplished.
The licensee initiated several corrective actions to address this issue, some of which included: establishing appropriate acceptance criteria for the safety injection lube oil coolers; developing a recovery plan for the licensee's Generic Letter 89-13 program and categorizing the program health in a red status; designating a single program owner to the Generic Letter 89-13 program; and reviewing other procedures utilized to implement the licensee's Generic Letter 89-13 program to verify specific acceptance criteria are contained in the procedures.
The inspectors verified the licensee's past operability analysis for the safety injection pumps. The inspectors evaluated the finding using the results of that analysis and Inspection Manual Chapter 0609, "Significance Determination Process," Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," Phase 1 screening, and determined that the finding was of very low safety significance.
Inspection Report# : 2004004(pdf)
Barrier Integrity
 
1Q/2005 Inspection Findings - Kewaunee                                                                                                Page 9 of 11 Significance:        Mar 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Reactor Operation Above LIcensed Power Limit A finding of very low safety significance associated with a Non-Cited Violation of the plant operating license was self-revealed during normal plant operations. The Kewaunee Nuclear Power Plant Facility Operating License, as amended stated, "The Nuclear Management Company (NMC) is authorized to operate the facility at steady-state reactor core power levels not in excess of 1772 megawatts (thermal)." Contrary to this, on January 31, 2005, the 8-hour average thermal power peaked at 1772.07 MWt before being restored to below 1772 MWt. Reactor power was allowed to rise above 1772 MWt because the 8-hr average reactor thermal power indicator on the plant process computer system was not reliable, and the site operating philosophy allowed the 1-minute average and the 15-minute average reactor thermal power indications to exceed 1772 Mwt. Once the 8-hour average was discovered to be in excess of that allowed in the Operating License, operators immediately lowered power to within the licensed limit and entered this issue into the corrective action program. This violation of the plant operating license was considered greater than minor, because it could affect the barrier integrity cornerstone objective of protecting the integrity of the fuel cladding and was associated with the barrier integrity cornerstone attributes of thermal limits and reactivity control. The finding also involved the crosscutting area of human performance. In accordance with Inspection Manual Chaper (IMC) 0609, Appendix A, Phase 1, the finding was of very low safety significance.
Inspection Report# : 2005003(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: VIO Violation Inability to Close Containment Equipment Hatch The inspectors identified an apparent violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, And Drawings," having potential safety significance greater than green. The finding was associated with the licensee's inability to close the containment equipment hatch in an expeditious manner while the plant was in the refueling shutdown mode, fuel was in the reactor vessel, the time to boil was estimated to be less than 30 minutes, and the reactor coolant system was open to the containment atmosphere. The inability to close the containment equipment hatch was caused by a design error in a large steel rail system installed inside the containment which was to be used to bring heavy equipment into the containment. This large steel rail system obstructed closure of the containment equipment hatch. The inability to close the hatch in an expeditious manner violated the licensee's procedure requirements to do so.
This finding was more than minor because it affected the Barrier Integrity Cornerstone objective and was associated with the Barrier Integrity Cornerstone attribute of containment boundary preservation. Since this finding was determined to be potentially greater than Green using the SDP Phase 2 Process, this finding is of a to-be-determined (TBD) safety significance pending review by the NRC Significance Determination Process/Enforcement Review Panel (SERP).
On May 5, 2005, a final significance determination letter was issued for a WHITE finding (IR 2005009).
Inspection Report# : 2004009(pdf)
Inspection Report# : 2005009(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Reactor Building Ventilation Isolation Function Not Available When Required A finding of very low safety significance associated with Technical Specification 3.8 a.1.b., "Refueling Operations - Containment Closure,"
was self-revealed during required daily surveillance testing of reactor building ventilation system isolation. During the surveillance test, plant operators discovered that radiation monitors would not cause a Reactor Building Ventilation System Isolation to occur as designed. The cause of this failure was that other engineered safeguards testing was in progress that disabled the Reactor Building Ventilation System Isolation function, which was required to be operable at the time. Once this issue was identified, the licensee restored the Reactor Building Ventilation System and entered this issue into the corrective action program.
This finding was more than minor, because it represented a degradation of the Barrier Integrity Cornerstone objective and was associated with Barrier Integrity cornerstone attribute of safety system and component performance. The finding was of very low safety significance because it did not result in the actual release of radioactive material. This finding was a Non-Cited Violation of Plant Technical Specification 3.8.a.1.b.,
"Refueling Operations-Containment Closure."
Inspection Report# : 2004009(pdf)
Significance:        Sep 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings." Failure to Implement Procedures for Work on Safety-Related Equipment A finding of very low safety significance was identified by the inspectors for a violation of 10 CFR Part 50, Appendix B, Criterion V,
 
1Q/2005 Inspection Findings - Kewaunee                                                                                            Page 10 of 11 "Instructions, Procedures, And Drawings." The licensee conducted corrective maintenance to fix a deficient condition on the containment personnel hatch seal, a safety-related component, under the toolpouch maintenance' process rather than with the use of a work request or a work order, contrary to procedural requirements. The primary cause of this finding was related to the cross-cutting area of human performance.
Licensee personnel failed to appropriately implement licensee procedures for conducting work on safety-related components. Once this was identified, the licensee performed an extent of condition evaluation on the work control process and identified that, since July 2002, approximately 14 percent of the work performed under toolpouch maintenance' had been performed on safety-related components without a work order. The licensee also implemented a number of corrective actions to ensure work on safety-related equipment is conducted according to procedural requirements.
This issue was more than minor because it affected the Barrier Integrity Cornerstone attribute of reactor containment integrity, and, if left uncorrected, the finding could become a more significant safety concern. The finding was of very low safety significance because it did not represent an actual open pathway in the physical integrity of the reactor containment and none of the work conducted on safety-related equipment without a work order resulted in an operability concern. This issue was a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, And Drawings."
Inspection Report# : 2004007(pdf)
Significance:        Jun 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to correct historical residual heat removal pump mechanical seal leakage; 10CFR Part 50, Appendix B, Criterion XVI, "Corrective Action" A finding of very low safety significance associated with a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Actions", was self-revealed on June 16, 2004, when licensee personnel discovered leakage from the B' residual heat removal (RHR) pump seal when the pump was stopped following the performance of a surveillance procedure on the B' RHR train. Plant personnel determined the leakage to be in excess of that specified in the plant's System Integrity Program for leakage from emergency core cooling systems outside containment. The leakage was also in excess of the amount of leakage assumed in the Updated Safety Analysis Report, Chapter 14, for calculation of control room habitability doses and offsite exposures. The inspectors subsequently determined, from interviews with licensee personnel and a review of the licensee's corrective action program and work order history, that excessive RHR seal leakage has occurred since the late 1980s. However, past corrective actions have not been effective to correct this condition adverse to quality.
The licensee performed a prompt engineering review to ensure that no immediate catastrophic failure mechanism for the RHR seal existed. The licensee also performed a prompt engineering review of the impact of the estimated leakage on the control room habitability doses, as well as the offsite doses, and determined no exposure limits would be exceeded. The licensee took actions to immediately stop the leakage and plans to replace the RHR pump seal during the next refueling outage.
This self-revealed finding was more than minor because the finding affected the cornerstone objective of Reactor Safety/Barrier Integrity. The inspectors evaluated the finding using Inspection Manual Chapter 0609, "Significance Determination Process," Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," Phase 1 screening, and determined that the finding was of very low safety significance.
Inspection Report# : 2004004(pdf)
Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Physical Protection information not publicly available.
 
1Q/2005 Inspection Findings - Kewaunee Page 11 of 11 Miscellaneous Last modified : June 17, 2005
 
2Q/2005 Inspection Findings - Kewaunee                                                                                              Page 1 of 10 Kewaunee 2Q/2005 Plant Inspection Findings Initiating Events Significance:        Jun 30, 2005 Identified By: NRC Item Type: FIN Finding Inadequate controls for loose material in substation A finding of very low safety significance was identified by the inspectors for failure to control loose materials in the protected area and substation. No violation of NRC requirements occurred. Once identified, the licensee initiated a condition report (CAP) to develop a surveillance procedure to remove loose materials before summer months where potential adverse weather was apparent.
The issue was more than minor because, if left uncontrolled, the loose items adjacent to the auxiliary transformers and in the substation would become a more significant safety concern. The issue was of very low safety significance because the finding did not contribute to the likelihood of a primary or secondary system loss of coolant accident initiator; the finding did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available; and the finding did not increase the likelihood of a fire or internal or external flooding. The issue was not considered a violation of regulatory requirements because it did not affect safety-related structures, systems, or components.
Inspection Report# : 2005008(pdf)
Significance:        Jun 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to manage risk duing periods where the grid condition was defined as unstable.
A finding of very low safety significance was identified by the inspectors for a Non-Cited Violation (NCV) of Title 10 CFR Part 50.65(a)(4).
The licensee failed to adequately assess shutdown risk during degraded grid conditions. Once identified, the licensee initiated a CAP to modify shutdown safety assessment and operating procedures to include grid conditions in risk assessments. The finding was more than minor because the licensee's risk assessment had incorrect assumptions that had the potential to change the outcome of the assessment. The inspectors determined that the finding could not be evaluated using the Significance Determination Process because the finding was associated with an inadequate qualitative risk assessment. The inspectors determined that this issue was of very low safety significance which was verified by the regional branch chief.
Inspection Report# : 2005008(pdf)
Significance:        Feb 18, 2005 Identified By: NRC Item Type: FIN Finding Safety Buses Relay Sensitivity to External Electrical Distubrances The team identified a finding of very low safety significance for a failure to provide adequate relay setpoint calibration tolerances on safety buses 1-5 and 1-6 loss of voltage relays. The existing relay setting calibration tolerances would have allowed the loss of voltage relays to actuate spuriously during certain offsite electrical system disturbances and un-necessarily separate the safety buses from the offsite power system and result in a plant transient. The licensee implemented corrective actions to revise the appropriate loss of voltage relay surveillance procedures.
The finding was more than minor because the failure to provide adequate relay setting tolerances could result in an unnecessary separation of the safety buses from the electrical grid and an ensuing plant transient. The finding was of very low safety significance because the issue would not preclude the safety buses from being re-energized by the emergency power sources. The finding was a not a violation of regulatory requirements.
Inspection Report# : 2005002(pdf)
Significance:        Feb 18, 2005 Identified By: NRC Item Type: NCV NonCited Violation Operator Actions Following Station Blackout - Lack of Procedure Guidance The team identified a Non-Cited Violation of 10 CFR 50.63, "Loss of All Alternating Current Power," for a failure to maintain procedural steps that minimized the likelihood and duration of a Station Blackout (SBO) event. The deleted procedural steps allowed for the cross-connection of the plant's two redundant safety buses should both the Reserve Auxiliary Transformer and the 1B Emergency Diesel Generator fail. These
 
2Q/2005 Inspection Findings - Kewaunee                                                                                              Page 2 of 10 procedural steps, as originally employed, served to lessen the likelihood of the SBO occurring, and/or reduce the time of the SBO. The licensee implemented corrective actions to revise the appropriate operations procedure.
This finding was more than minor, because it was associated with the likelihood of an initiating event and the reliability of a safety bus that responds to an initiating event. The finding was of very low safety significance, because multiple sources of both onsite and offsite power remained available to supply the two safety buses.
Inspection Report# : 2005002(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Control Of Combustible Matrials A finding of very low safety significance was identified by the inspectors for a violation of a fire protection License Condition. The inspectors identified multiple examples of combustible materials either stored or in use without specific authorization. Specifically, the licensee stored and used lubricating oil in an emergency diesel generator room beyond that authorized by the Fire Protection Program Analysis, the licensee stored unauthorized combustible materials above the shelves in the working materials storage area and on top of cabinets nearby, and the licensee stored compressed flammable gas cylinders in the auxiliary building without authorization. Once these issues were identified, the licensee removed the unauthorized materials. This finding was related to the cross-cutting area of problem identification and resolution in that the NRC had previously identified issues relating to control of transient combustible materials above and near the working materials storage area but adequate corrective actions were not put in place to prevent recurrence of this isse.
The finding was more than minor because the failure to adequately control combustible materials, if left uncorrected, could become a more safety significant concern. The finding was of very low safety significance because the issue was a low degradation of fire prevention and administrative controls. The finding was a Non-Cited Violation of License Condition 2.C(3) which required specific authorization for the storage and use of combustibles in safety-related areas.
Inspection Report# : 2004009(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Corrective Action to Preclude Storage of Oxygen Cylinders Next to Flammable Gas Cylinders A finding of very low safety significance was identified by the inspectors for a violation of a fire protection License Condition. The inspectors identified the storage of compressed oxygen cylinders near compressed flammable gas cylinders. Once this issue was identified, the licensee removed the stored compressed oxygen cylinders from the area.
The finding was more than minor because the inappropriate storage of compressed oxygen cylinders could result in greater severity of a fire affecting equipment important to safety. The finding was of very low safety significance because the issue was a low degradation of fire prevention and administrative controls. The finding was a Non-Cited Violation of License Condition 2.C(3) which required the bulk storage of compressed oxygen cylinders to be separated from compressed flammable gas cylinders and corrective action of conditions significantly adverse to quality to preclude recurrence.
Inspection Report# : 2004009(pdf)
Mitigating Systems Significance:        Jun 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to minimize prior identified and predictable explosive gas concentrations in the WGDTs A finding of very low safety significance was identified by the inspectors for a NCV of Title 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action." The licensee failed to consider the impact on plant fire protection when ineffective resolution of waste gas system issues repeatedly led to explosive mixtures in the Waste Gas Decay Tanks. The licensee entered these issues into their corrective action program. The primary cause of this violation was related to the cross-cutting area of Problem Identification and Resolution. The licensee repeatedly encountered explosive gas levels in the WGDTs and were aware of plant conditions that resulted in these levels but failed to take adequate corrective actions to prevent explosive gas mixtures from developing in the WGDTs. The issue is more than minor because uncontrolled explosive mixtures in the WGDTs could have led to a more significant safety concern. The issue was of very low safety significance because explosive mixtures were only present during plant shutdown conditions; an explosion would not have affected safe shutdown equipment (i.e.
Residual Heat Removal System); the explosive mixture conditions were only present for short periods of time (<12 hours); and the tanks were isolated and vented per procedure when discovered.
Inspection Report# : 2005008(pdf)
 
2Q/2005 Inspection Findings - Kewaunee                                                                                                Page 3 of 10 Significance:        Feb 18, 2005 Identified By: NRC Item Type: FIN Finding Lack of 4160 Vac Bus 1-5 Ovewrcurrent and Loss of Voltage Relay Coordination The team identified a finding of very low safety significance for a failure to provide adequate electrical coordination of protective devices thereby ensuring that postulated electrical faults would be isolated upon detection. Specifically, the team identified that the lack of adequate electrical systems coordination between the undervoltage and overcurrent protection on 4160 Vac safety bus 1-5 would result in the loss of voltage relays actuating before the bus over-current relays. This design deficiency results in the failure to lock out safety bus 1- 5 upon postulated electrical faults and subjects the postulated faulted safety bus 1-5 to be re-energized via an alternate offsite source. This design introduced a challenge to the safety equipment availability and reliability. The licensee planned to develop changes to the affected relays.
The finding was more than minor because the failure to provide adequate electrical coordination of electrical devices provided an unnecessary challenge to safety-related equipment, and if left uncorrected, could become a more safety significant concern. The finding was of very low safety significance because it was a design deficiency that did not result in the loss of system function. The finding was a not a violation of regulatory requirements.
Inspection Report# : 2005002(pdf)
Significance:        Feb 18, 2005 Identified By: NRC Item Type: NCV NonCited Violation Short Circuit Duty of Buses Exceeded - Impact on Safe Shutdown Analysis The team identified a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion II, "Quality Assurance Program," for a failure to identify potentially adverse conditions to the plant's fire protection safe shutdown analysis caused by known overduty conditions on non-safety related buses 1-1, 1-2, 1-3, and 1-4. While the overduty condition was known to have existed at least since 1992, the licensee never entered the issue into the plant's corrective action program, where a proper evaluation should have addressed 10 CFR Part 50, Appendix R, safe shutdown related effects. The licensee planned to continue efforts to identify additional evaluations and corrective actions.
This finding was more than minor, because it was associated with the degradation of a fire protection feature. The finding was of very low safety significance because after extensive evaluation of the deficiency, the licensee was able to determine that the plant could still safely shut down the plant during a postulated fire event.
Inspection Report# : 2005002(pdf)
Significance:        Feb 18, 2005 Identified By: NRC Item Type: NCV NonCited Violation Battery Sizing Deficiencies The team identified a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for failure to implement adequate design controls of documents, inputs, and assumptions in the design of the two safety-related batteries. Specifically, the licensee did not perform and control battery sizing calculations, including consideration of temperature effects, to ensure that the batteries maintained sufficient capacity to perform the intended design function. The team determined that the failure to appropriately evaluate effects of battery room and cell temperatures also affected the cross-cutting area of Problem Identification and Resolution because the subject of battery capacity versus battery temperature had been previously identified in a 1992 NRC inspection. The licensee planned to perform battery sizing calculations as part of an overall electrical systems analysis improvement project.
This finding was more than minor because it affected the mitigating systems cornerstone objective of ensuring the availability and reliability of the 125 Volts direct current battery system to respond to initiating events to prevent undesirable consequences. The finding is of very low safety significance because the battery remained operable. The licensee planned to develop formal battery sizing calculations.
Inspection Report# : 2005002(pdf)
Significance: SL-IV Feb 18, 2005 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Evaluation of Procedure Changes to Address AFW Design Deficiencies The team identified a finding involving a Non-Cited Violation of 10 CFR 50.59, "Changes, Tests, and Experiments." The finding involved a failure to perform an adequate review of operations procedure changes in accordance with 10 CFR 50.59 associated with the operation of motor-operated valves for the auxiliary feedwater suction source from the service water system. The team determined that the licensee's approval of changes to Procedure E-0-05, with the introduction of adverse effects, and a determination that 10 CFR 50.59 was not applicable was a violation of 10 CFR 50.59. The licensee subsequently performed additional evaluations of the procedure changes.
Because the issue affected the NRC's ability to perform its regulatory function, this finding was evaluated with the traditional enforcement process. The finding was determined to be of very low safety significance since the design basis safety-related function of the AFW system, to remove reactor decay heat following a loss of normal feedwater, was not adversely affected. This was determined to be a Severity Level IV NCV of 10 CFR 50.59.
 
2Q/2005 Inspection Findings - Kewaunee                                                                                                Page 4 of 10 Inspection Report# : 2005002(pdf)
Significance:        Feb 18, 2005 Identified By: NRC Item Type: NCV NonCited Violation Lack of Allowance for Manual Actions in Establishing Setpoint to Transfer AFW Pump Suction Source The team identified a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for failure to establish the condensate storage tank (CST) level setpoint to transfer the auxiliary feedwater (AFW) pump suction supply from the CST to service water.
The team determined that the calculation setpoint did not include an allowance for the manual operator actions required by emergency operations procedures. The licensee revised the plant procedure to perform the operator actions earlier in the procedure.
This finding was more than minor because it affected the mitigating systems cornerstone objective of equipment reliability, in that failure to align the AFW pump suctions to service water prior to the CSTs being depleted could have resulted in damage to the AFW pumps. The finding was determined to be of very low safety significance because it was a design deficiency that did not result in a loss of function.
Inspection Report# : 2005002(pdf)
Significance:        Feb 18, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Ensure that Calculation Assumption was Based on Valid Times for Manual Operator Actions The team identified a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control." The finding involved the condensate storage tank (CST) level setpoint to transfer the auxiliary feedwater (AFW) pump suction from the CSTs to service water. A calculation assumption stated that a flow would drain from the CSTs to the condenser for 10 minutes until the operators isolated the flow by closing manual valve MU-2A. The team determined that the actions could not be completed in the time assumed by the calculation. The licensee initiated corrective actions to revise the appropriate operations procedure and calculation.
This finding was greater than minor because it affected the mitigating system cornerstone objective of equipment reliability, in that failure to align the AFW pump suctions to service water prior to the CSTs being depleted could have resulted in damage to the AFW pumps. The finding was determined to be of very low safety significance because it was a design deficiency that was not found to result in a loss of function. The team concluded that it was unlikely that the operators would allow the CST level to reach the EOP setpoint without attempting to refill the tanks from other sources, and that the operators would be aware of the CST levels.
Inspection Report# : 2005002(pdf)
Significance:        Feb 18, 2005 Identified By: NRC Item Type: NCV NonCited Violation TSC DG Target Reliability Methodology Inadequate The team identified a Non-Cited Violation of 10 CFR 50.63, "Loss of All Alternating Current Power." The finding involved the failure to establish a target reliability for the plant's alternate power source consistent with the reliability approved by the NRC staff in the licensee's Station Blackout submittal for 10 CFR 50.63. The non-conservative target reliability employed by the licensee resulted in the failure of the licensee to increase efforts to restore the Technical Support Center (TSC) Diesel Generator (DG) to its approved target reliability at an earlier date. The licensee subsequently initiated a corrective action to change the TSC DG reliability methodology.
This finding was more than minor, because it affected the reliability of a support system required for the mitigation of an Station Blackout event. The finding was of very low safety significance, because the finding did not directly affect the immediate operability of the TSC DG.
Inspection Report# : 2005002(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Scaffolding Erected too Close to Safety-Related Equipment Required to be Operable A finding of very low safety significance was identified by the inspectors for a violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings." The finding was associated with the licensee's failure to adequately implement scaffold control requirements contained in Procedure GMP-127, "Requirements and Guidelines for Scaffold Construction and Inspection," which required that scaffolding be no closer than 2 inches from any safety-related equipment unless otherwise evaluated and approved by Engineering. Specifically, scaffolding was erected within 2 inches of safety-related piping for the Service Water outlet from the jacket water heat exchangers for Diesel Generator B, the piping for the Emergency Borate MOV (CVC-440), and Safety Injection Pump A, without engineering evaluation and approval. Upon discovery of this condition, the licensee took immediate action to bring all noted scaffolding problems into compliance with licensee procedures and initiated a CAP document for the issue.
The finding was more than minor because, if left uncorrected, the issue may have resulted in a more significant safety concern. Specifically, the
 
2Q/2005 Inspection Findings - Kewaunee                                                                                                  Page 5 of 10 failure of scaffolding having adequate spacing in the vicinity of safety-related equipment during a seismic event could result in damage to mitigating equipment. The finding was of very low safety significance because it did not result in the actual loss of the safety function of the train or system. The finding was a Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings."
Inspection Report# : 2004009(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify Inadequate Pre-Fire Strategies A finding of very low safety significance was identified by the inspectors for a violation of a fire protection License Condition. The inspectors identified that the licensee failed to identify pertinent information, such as the presence of compressed flammable gas cylinders, on a fire area strategy for fire brigade personnel. Once this issue was identified, the licensee revised the fire area strategy for the affected area.
The finding was more than minor because the failure to provide adequate warnings and guidance relating to hazards associated with compressed flammable gas cylinders in fire strategies could adversely impact fire fighting strategies used by the fire brigade in fighting a fire.
The finding was of very low safety significance due to extensive training provided to fire brigade members to deal with unexpected contingencies. The finding was a Non-Cited Violation of License Condition 2.C(3) which required that fire area strategies provide pertinent information to help the fire brigade to be better prepared for fire fighting within that area.
Inspection Report# : 2004009(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Non-conforming Condition on the Safety-Related Containment Sump A finding of very low safety significance was identified by the inspectors for a violation of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Actions." The original licensing and design basis of the containment sump screens was to prevent any particles greater than 1/8 inch from entering the sump. The inspectors determined that the screen size was 1/8-inch by 15/32-inch which allowed particles greater than 1/8-inch to enter the sump. The inspectors subsequently determined that this issue had been identified and entered into the licensee's corrective action program in 1997. However, adequate corrective actions were not taken to correct this condition adverse to quality. Once this issue was identified, the licensee conducted an operability determination and concluded that there were no immediate operability issues with the containment sump. The licensee determined that the sump screens were nonconforming in accordance with Generic Letter 91-18, and planned long term corrective actions to be developed in conjunction with the resolution of Generic Safety Issue 191 and NRC Generic Letter 2004-02.
The inspectors concluded that the primary cause of this finding was related to the performance characteristic of corrective actions in the cross-cutting area of problem identification and resolution.
This finding was more than minor because the issue affected the Mitigating System cornerstone attribute of design control for initial design and equipment performance reliability and affected the associated cornerstone objective to ensure the reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was of very low safety significance because it was not a design or qualification deficiency that has been confirmed to result in a loss of function per Generic Letter 91-18. This finding was a Non-Cited Violation of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Actions."
Inspection Report# : 2004009(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Instructions and Procedures for Inspections and Cleaning of the Safety-Related Containment Sump A finding of very low safety significance was identified by the inspectors for a violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, And Drawings," regarding licensee instructions and procedures for containment sump inspections. Specifically, the inspectors identified that current licensee procedures did not require inspection or cleaning when boric acid or small debris may be present in the containment sump. The licensee's procedures for containment coatings did not require inspection of the coating located inside the containment sump which had not been inspected since initial application; and the licensee's procedure for containment sump gap inspections did not specify acceptance criteria to ensure this activity was satisfactorily accomplished. The licensee subsequently initiated several corrective actions to address these issues which included, but are not limited to: immediate inspection and cleaning of the safety-related containment sump; immediate inspection and assessment of the safety-related sump concrete coating; revision of preventive maintenance activities to require inspection and cleaning of the safety-related containment sump every refueling outage; revision of procedures to include inspection of the safety-related containment sump concrete coating every refueling outage; and revision of procedures to include appropriate acceptance criteria for determining that important activities were satisfactorily accomplished.
This finding was more than minor because if left uncorrected the finding could become a more significant safety concern and the issue affected the Mitigating System cornerstone attributes of equipment performance reliability and procedure quality and affected the associated cornerstone objective to ensure the reliability and capability of systems that respond to initiating events to prevent undesirable consequences.
The finding was of very low safety significance because it was not a design or qualification deficiency that has been confirmed to result in a loss of function per Generic Letter 91-18. This finding was a Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, "Instructions,
 
2Q/2005 Inspection Findings - Kewaunee                                                                                              Page 6 of 10 Procedures, and Drawings."
Inspection Report# : 2004009(pdf)
Significance:        Sep 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action." Failure to Promptly Correct Conditions Adverse to Quality, Specifically Associated with Degraded and Nonconforming Conditions A finding of very low safety significance was identified by the inspectors for a violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Actions." During a review of the licensee's list of safety-related equipment designated as degraded or nonconforming, the inspectors identified that the licensee failed to promptly correct three conditions adverse to quality. These conditions adverse to quality included noncompliance of both Residual Heat Removal pump seal coolers with system design requirements, which was previously identified by NRC inspectors in November 2002, but not promptly corrected by the licensee; and two sections of safety-related piping, one associated with the "B" Emergency Diesel Generator fuel oil supply and the other associated with the Component Cooling Water piping from the "B" Residual Heat Removal pump seal cooler and stuffing box, that were identified by the licensee in September and April 2003, respectively, as exceeding Updated Safety Analysis Report stress criteria but not promptly corrected by the licensee. The primary cause of this finding was related to the cross-cutting area of problem identification and resolution. The licensee failed to prioritize and promptly correct these conditions adverse to quality in accordance with the guidelines in the corrective action program. Once these conditions were identified, the licensee restored the following conditions to operable: the A' RHR Pump Seal Cooler; the CCW piping expansion loop from the B' RHR pump seal cooler; and the fuel oil supply piping to the B' EDG. The licensee planned to restore the B' RHR Pump Seal Cooler during the upcoming Fall 2004 Refueling Outage.
This issue was more than minor because it affected the Mitigating System cornerstone attribute of design control for initial design and plant modifications and affected the associated cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was of very low safety significance because it was not a design or qualification deficiency that has been confirmed to result in a loss of function per Generic Letter 91-18. This issue was a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Actions."
Inspection Report# : 2004007(pdf)
Significance:        Sep 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings." Failure to Have Procedures Appropriate to the Circumstances for Preventive Maintenance of the TDAFW Pump Turbine A finding of very low safety significance was self-revealed during the licensee's review of high oil particulate in the Turbine Driven Auxiliary Feedwater Pump Turbine, which resulted in a violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings."
The licensee determined that high oil particulate in the Turbine Driven Auxiliary Feedwater Pump Turbine was caused by damage to the journal bearing. Maintenance procedures did not specify appropriate acceptance criteria for oil sampling, did not specify an appropriate inspection frequency and criteria for the turbine bearings and bearing cavities, and allowed the reuse of bearings in different locations during maintenance of the Turbine, which were not acceptable maintenance practices. The reuse of the upper inboard bearing in a different location contributed to the journal bearing damage. The licensee took immediate remedial corrective actions to replace the bearings, clean the housing and return the pump to service. In addition, the licensee revised its maintenance procedures to include appropriate instructions for turbine and pump maintenance activities.
This self-revealed finding was more than minor because, if left uncorrected, the issue would have become a more significant safety concern. In addition, it affected the Mitigating Systems attributes of equipment performance reliability and the Mitigating Systems cornerstone objective of ensuring the reliability of systems. The finding was of very low safety significance because it was not a design or qualification deficiency that has been confirmed to result in a loss of function per Generic Letter 91-18. This issue was a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, And Drawings."
Inspection Report# : 2004007(pdf)
Significance:        Sep 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings." Failure to Have Acceptance Criteria for Flushing of the 1ARHR Fan Coil Unit A finding of very low safety significance was identified by the inspectors for a violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, And Drawings." This finding was associated with the licensee's failure to implement an appropriate inspection and cleaning procedure containing quantitative or qualitative acceptance criteria for the 1A RHR pump pit Fan Coil Unit to ensure that cleaning was satisfactorily accomplished. Following discovery, the licensee entered the issue into its corrective action program and conducted an immediate operability assessment that determined the involved fan coil units were operable.
This issue was more than minor because it involved the procedure quality attribute of the Mitigating Systems cornerstone and affected the
 
2Q/2005 Inspection Findings - Kewaunee                                                                                                  Page 7 of 10 cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was of very low safety significance because it was not a design or qualification deficiency that has been confirmed to result in a loss of function per Generic Letter 91-18. This issue was a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, And Drawings."
Inspection Report# : 2004007(pdf)
Significance:        Sep 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation 10 CFR 50, Appendix B, Criterion III, "Design Control." Failure to Verify the Acceptability of a Single Failure Vulnerability Introduced During a System Modification 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." This finding was associated with the licensee's failure to perform a design verification to demonstrate that the diesel generator lube oil cooler service water outlet valve actuators, installed under Design Change 3357, would not result in a failure of the valve stems under conditions in which the valve ball froze nor had the licensee provided sufficient justification to show that valve ball freezing was not credible.
Following discovery, the licensee entered the issue into its corrective action program and performed an operability assessment which provided additional justification to demonstrate that the stem failure was considered not credible.
This issue was more than minor because it involved 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. The finding was of very low safety significance because it was not a design or qualification deficiency that has been confirmed to result in a loss of function per Generic Letter 91-18. This issue was a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control."
Inspection Report# : 2004007(pdf)
Significance:        Jul 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Repair the Deluge System Heat Detectors in a Timely Manner The team identified a Non-Cited Violation of a license condition for fire protection. The licensee failed to take timely corrective actions to repair several maintenance storage area deluge system rate-of-rise heat detectors which were inoperable for an extended period of time. At the time of this inspection, the detectors had been repaired and returned to operability.
The finding was greater than minor because it affected the mitigating systems cornerstone attribute of protection against external factors (fire).
Specially, a partially inoperable deluge system can increase the likelihood of a fire which could challenge safe shutdown. The finding was of very low safety significance because this fire area has Pyr-A-Larm ionization detectors located at the ceiling level. These detectors would alarm in the control room and the fire brigade would respond to a fire in this area. In addition, other defense-in-depth fire protection elements remained unaffected and fire in this area would not result in a loss of dedicated safe shutdown systems.
Inspection Report# : 2004005(pdf)
Significance:        Jul 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Acceptable (Quality Related) Pre-Fire Strategies A finding of very low safety significance was identified by the team for a violation of a license condition for fire protection. The licensee failed to include pertinent information in their fire strategies. Specifically, the licensee failed to include information about the potential unavailability of certain fire hose stations and identify hydrogen and propane piping hazards in a fire zone. Once the issues were identified, the licensee entered the issue into their corrective action program and planned to revise their fire strategies to include the pertinent information.
The issue was greater than minor because the failure to include pertinent information relating to the water supply used for manual fire fighting and hazards associated with hydrogen and propane piping in fire strategies could adversely impact fire fighting strategies used by the fire brigade in fighting a fire. The issue was of very low safety significance because of the extensive training provided to fire brigade members to deal with unexpected contingencies. The issue was a Non-Cited Violation of License Condition 2.C(3) which required, in part, that fire area strategies provide the fire brigade pertinent information on a given plan area to help the brigade to be better prepared for fire fighting within that area.
Inspection Report# : 2004005(pdf)
Significance:        Jul 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation
 
2Q/2005 Inspection Findings - Kewaunee                                                                                                Page 8 of 10 Failure to Meet the Fire Protection Program Requirements for Hose Lengths to Maintain an Acceptable Water Pressure and Flow at Hose Stations The team identified a Non-Cited Violation of License Condition 2.C(3), which requires the licensee to implement all provisions of their NRC approved fire protection program. The licensee failed to meet the fire protection program requirements for hose lengths to maintain an acceptable water pressure and flow to hose stations. The licensee's corrective actions included replacing hoses to increase water flow at hose stations The finding was greater than minor because it affected the mitigating systems cornerstone attribute of protection against external factors (fire).
Specifically, the failure to maintain acceptable water pressure and water flow to hose stations can hamper the brigade's ability to fight a fire, thereby, potentially endangering mitigating systems. The finding was of very low safety significance because the problem only impacts the effectiveness of the fire brigade while other fire protection features, such as fire barriers and physical separation remain available.
Inspection Report# : 2004005(pdf)
Significance:        Jul 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Meet the NFPA Code Requirements for Extinguisher Placement The team identified a Non-Cited Violation of License Condition 2.C.(3), which requires the licensee to implement all provisions of their approved fire protection program. Amendment No. 23 to Facility Operating License Safety Evaluation Report dated December 12, 1978, required fire extinguishers in accordance with the National Fire Protection Association Code. The licensee failed to meet the Code requirements for extinguisher placement in Fire Area AX-32. Once identified, the licensee initiated corrective actions to meet the Code requirements.
The finding was greater than minor because it affected the mitigating systems cornerstone attribute of protection against external factors (fire).
Specially, not having an extinguisher to put out a small fire can increase the likelihood of a fire which could challenge safe shutdown. The finding was of very low safety significance because this fire area has fire detectors that would alarm in the control room and the fire brigade would respond to a fire in this area. In addition, other defense-in-depth fire protection elements remained unaffected and fire in this area would not result in a loss of dedicated safe shutdown systems.
Inspection Report# : 2004005(pdf)
Significance:        Jul 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Transient Combustibles Not Adequately Controlled Within Fire Area AX-32 The team identified a Non-Cited Violation of License Condition 2.C(3) for failure to adequately control transient combustibles in fire area AX-
: 32. Specifically, authorization for the storage and use of combustibles in safety-related areas was not obtained. Once uncontrolled transient combustibles were identified, the materials were either included in the transient combustible permit system or removed from the area.
The issue was greater than minor because the failure to adequately control combustible materials could result in a more significant safety issue.
Uncontrolled combustibles could result in the greater likelihood or severity of a fire which affects equipment important to safety. The finding was of very low safety significance because of mitigation capability available in the event of a fire in fire area AX-32.
Inspection Report# : 2004005(pdf)
Significance:        Jul 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation Transient Combustibles Not Adequately Controlled Within Fire Area AX-24 The team identified a Non-Cited Violation of License Condition 2.C(3), in that a hazardous quantity of transient combustibles was present in fire area AX-24. The hazardous quantity of transient combustibles present exceeded the quantity of combustibles allowed with no fire detection systems in this fire area.
The finding was greater than minor because it affected the mitigating systems cornerstone attribute of protection against external factors (fire).
Specifically, the presence of transient combustibles beyond what was approved by the NRC could result in the increased likelihood of a fire which could challenge safe shutdown. The finding was of very low safety significance because a fire from the observed transient combustibles would not result in a loss of the alternate shutdown systems.
Inspection Report# : 2004005(pdf)
Barrier Integrity
 
2Q/2005 Inspection Findings - Kewaunee                                                                                                Page 9 of 10 Significance:        Mar 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Reactor Operation Above LIcensed Power Limit A finding of very low safety significance associated with a Non-Cited Violation of the plant operating license was self-revealed during normal plant operations. The Kewaunee Nuclear Power Plant Facility Operating License, as amended stated, "The Nuclear Management Company (NMC) is authorized to operate the facility at steady-state reactor core power levels not in excess of 1772 megawatts (thermal)." Contrary to this, on January 31, 2005, the 8-hour average thermal power peaked at 1772.07 MWt before being restored to below 1772 MWt. Reactor power was allowed to rise above 1772 MWt because the 8-hr average reactor thermal power indicator on the plant process computer system was not reliable, and the site operating philosophy allowed the 1-minute average and the 15-minute average reactor thermal power indications to exceed 1772 Mwt. Once the 8-hour average was discovered to be in excess of that allowed in the Operating License, operators immediately lowered power to within the licensed limit and entered this issue into the corrective action program. This violation of the plant operating license was considered greater than minor, because it could affect the barrier integrity cornerstone objective of protecting the integrity of the fuel cladding and was associated with the barrier integrity cornerstone attributes of thermal limits and reactivity control. The finding also involved the crosscutting area of human performance. In accordance with Inspection Manual Chaper (IMC) 0609, Appendix A, Phase 1, the finding was of very low safety significance.
Inspection Report# : 2005003(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: VIO Violation Inability to Close Containment Equipment Hatch The inspectors identified an apparent violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, And Drawings," having potential safety significance greater than green. The finding was associated with the licensee's inability to close the containment equipment hatch in an expeditious manner while the plant was in the refueling shutdown mode, fuel was in the reactor vessel, the time to boil was estimated to be less than 30 minutes, and the reactor coolant system was open to the containment atmosphere. The inability to close the containment equipment hatch was caused by a design error in a large steel rail system installed inside the containment which was to be used to bring heavy equipment into the containment. This large steel rail system obstructed closure of the containment equipment hatch. The inability to close the hatch in an expeditious manner violated the licensee's procedure requirements to do so.
This finding was more than minor because it affected the Barrier Integrity Cornerstone objective and was associated with the Barrier Integrity Cornerstone attribute of containment boundary preservation. Since this finding was determined to be potentially greater than Green using the SDP Phase 2 Process, this finding is of a to-be-determined (TBD) safety significance pending review by the NRC Significance Determination Process/Enforcement Review Panel (SERP).
On May 5, 2005, a final significance determination letter was issued for a WHITE finding (IR 2005009).
Inspection Report# : 2004009(pdf)
Inspection Report# : 2005009(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Reactor Building Ventilation Isolation Function Not Available When Required A finding of very low safety significance associated with Technical Specification 3.8 a.1.b., "Refueling Operations - Containment Closure,"
was self-revealed during required daily surveillance testing of reactor building ventilation system isolation. During the surveillance test, plant operators discovered that radiation monitors would not cause a Reactor Building Ventilation System Isolation to occur as designed. The cause of this failure was that other engineered safeguards testing was in progress that disabled the Reactor Building Ventilation System Isolation function, which was required to be operable at the time. Once this issue was identified, the licensee restored the Reactor Building Ventilation System and entered this issue into the corrective action program.
This finding was more than minor, because it represented a degradation of the Barrier Integrity Cornerstone objective and was associated with Barrier Integrity cornerstone attribute of safety system and component performance. The finding was of very low safety significance because it did not result in the actual release of radioactive material. This finding was a Non-Cited Violation of Plant Technical Specification 3.8.a.1.b.,
"Refueling Operations-Containment Closure."
Inspection Report# : 2004009(pdf)
Significance:        Sep 30, 2004 Identified By: NRC Item Type: NCV NonCited Violation 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings." Failure to Implement Procedures for Work on Safety-Related Equipment A finding of very low safety significance was identified by the inspectors for a violation of 10 CFR Part 50, Appendix B, Criterion V,
 
2Q/2005 Inspection Findings - Kewaunee                                                                                            Page 10 of 10 "Instructions, Procedures, And Drawings." The licensee conducted corrective maintenance to fix a deficient condition on the containment personnel hatch seal, a safety-related component, under the toolpouch maintenance' process rather than with the use of a work request or a work order, contrary to procedural requirements. The primary cause of this finding was related to the cross-cutting area of human performance.
Licensee personnel failed to appropriately implement licensee procedures for conducting work on safety-related components. Once this was identified, the licensee performed an extent of condition evaluation on the work control process and identified that, since July 2002, approximately 14 percent of the work performed under toolpouch maintenance' had been performed on safety-related components without a work order. The licensee also implemented a number of corrective actions to ensure work on safety-related equipment is conducted according to procedural requirements.
This issue was more than minor because it affected the Barrier Integrity Cornerstone attribute of reactor containment integrity, and, if left uncorrected, the finding could become a more significant safety concern. The finding was of very low safety significance because it did not represent an actual open pathway in the physical integrity of the reactor containment and none of the work conducted on safety-related equipment without a work order resulted in an operability concern. This issue was a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, And Drawings."
Inspection Report# : 2004007(pdf)
Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Physical Protection Physical Protection information not publicly available.
Miscellaneous Last modified : August 24, 2005
 
3Q/2005 Inspection Findings - Kewaunee                                                                                                Page 1 of 9 Kewaunee 3Q/2005 Plant Inspection Findings Initiating Events Significance:        Jun 30, 2005 Identified By: NRC Item Type: FIN Finding Inadequate controls for loose material in substation A finding of very low safety significance was identified by the inspectors for failure to control loose materials in the protected area and substation. No violation of NRC requirements occurred. Once identified, the licensee initiated a condition report (CAP) to develop a surveillance procedure to remove loose materials before summer months where potential adverse weather was apparent.
The issue was more than minor because, if left uncontrolled, the loose items adjacent to the auxiliary transformers and in the substation would become a more significant safety concern. The issue was of very low safety significance because the finding did not contribute to the likelihood of a primary or secondary system loss of coolant accident initiator; the finding did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available; and the finding did not increase the likelihood of a fire or internal or external flooding. The issue was not considered a violation of regulatory requirements because it did not affect safety-related structures, systems, or components.
Inspection Report# : 2005008(pdf)
Significance:        Jun 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to manage risk duing periods where the grid condition was defined as unstable.
A finding of very low safety significance was identified by the inspectors for a Non-Cited Violation (NCV) of Title 10 CFR Part 50.65(a)(4).
The licensee failed to adequately assess shutdown risk during degraded grid conditions. Once identified, the licensee initiated a CAP to modify shutdown safety assessment and operating procedures to include grid conditions in risk assessments. The finding was more than minor because the licensee's risk assessment had incorrect assumptions that had the potential to change the outcome of the assessment. The inspectors determined that the finding could not be evaluated using the Significance Determination Process because the finding was associated with an inadequate qualitative risk assessment. The inspectors determined that this issue was of very low safety significance which was verified by the regional branch chief.
Inspection Report# : 2005008(pdf)
Significance:        Feb 18, 2005 Identified By: NRC Item Type: FIN Finding Safety Buses Relay Sensitivity to External Electrical Distubrances The team identified a finding of very low safety significance for a failure to provide adequate relay setpoint calibration tolerances on safety buses 1-5 and 1-6 loss of voltage relays. The existing relay setting calibration tolerances would have allowed the loss of voltage relays to actuate spuriously during certain offsite electrical system disturbances and un-necessarily separate the safety buses from the offsite power system and result in a plant transient. The licensee implemented corrective actions to revise the appropriate loss of voltage relay surveillance procedures.
The finding was more than minor because the failure to provide adequate relay setting tolerances could result in an unnecessary separation of the safety buses from the electrical grid and an ensuing plant transient. The finding was of very low safety significance because the issue would not preclude the safety buses from being re-energized by the emergency power sources. The finding was a not a violation of regulatory requirements.
Inspection Report# : 2005002(pdf)
Significance:        Feb 18, 2005 Identified By: NRC Item Type: NCV NonCited Violation Operator Actions Following Station Blackout - Lack of Procedure Guidance The team identified a Non-Cited Violation of 10 CFR 50.63, "Loss of All Alternating Current Power," for a failure to maintain procedural steps that minimized the likelihood and duration of a Station Blackout (SBO) event. The deleted procedural steps allowed for the cross-connection of the plant's two redundant safety buses should both the Reserve Auxiliary Transformer and the 1B Emergency Diesel Generator fail. These
 
3Q/2005 Inspection Findings - Kewaunee                                                                                                Page 2 of 9 procedural steps, as originally employed, served to lessen the likelihood of the SBO occurring, and/or reduce the time of the SBO. The licensee implemented corrective actions to revise the appropriate operations procedure.
This finding was more than minor, because it was associated with the likelihood of an initiating event and the reliability of a safety bus that responds to an initiating event. The finding was of very low safety significance, because multiple sources of both onsite and offsite power remained available to supply the two safety buses.
Inspection Report# : 2005002(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Control Of Combustible Matrials A finding of very low safety significance was identified by the inspectors for a violation of a fire protection License Condition. The inspectors identified multiple examples of combustible materials either stored or in use without specific authorization. Specifically, the licensee stored and used lubricating oil in an emergency diesel generator room beyond that authorized by the Fire Protection Program Analysis, the licensee stored unauthorized combustible materials above the shelves in the working materials storage area and on top of cabinets nearby, and the licensee stored compressed flammable gas cylinders in the auxiliary building without authorization. Once these issues were identified, the licensee removed the unauthorized materials. This finding was related to the cross-cutting area of problem identification and resolution in that the NRC had previously identified issues relating to control of transient combustible materials above and near the working materials storage area but adequate corrective actions were not put in place to prevent recurrence of this isse.
The finding was more than minor because the failure to adequately control combustible materials, if left uncorrected, could become a more safety significant concern. The finding was of very low safety significance because the issue was a low degradation of fire prevention and administrative controls. The finding was a Non-Cited Violation of License Condition 2.C(3) which required specific authorization for the storage and use of combustibles in safety-related areas.
Inspection Report# : 2004009(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Corrective Action to Preclude Storage of Oxygen Cylinders Next to Flammable Gas Cylinders A finding of very low safety significance was identified by the inspectors for a violation of a fire protection License Condition. The inspectors identified the storage of compressed oxygen cylinders near compressed flammable gas cylinders. Once this issue was identified, the licensee removed the stored compressed oxygen cylinders from the area.
The finding was more than minor because the inappropriate storage of compressed oxygen cylinders could result in greater severity of a fire affecting equipment important to safety. The finding was of very low safety significance because the issue was a low degradation of fire prevention and administrative controls. The finding was a Non-Cited Violation of License Condition 2.C(3) which required the bulk storage of compressed oxygen cylinders to be separated from compressed flammable gas cylinders and corrective action of conditions significantly adverse to quality to preclude recurrence.
Inspection Report# : 2004009(pdf)
Mitigating Systems Significance: SL-IV Sep 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Report in a Timely Manner an Unanalyzed Condition Involving a Potential Runout Concern With the CCW Pumps The inspectors identified a Non-Cited Violation (NCV) when the licensee failed to make a written report, within 60 days, to the NRC in accordance with 10 CFR 50.73(a)(2)(ii)(B), when an unanalyzed condition that significantly degraded plant safety was identified. Specifically, the licensee did not recognize the significance of a previously identified condition involving a potential runout issue with the component cooling water (CCW) pumps, and did not report this condition until the inspectors identified the requirement. The concern related to the CCW pump capability to provide required flow under certain conditions. Specifically, during a loss of power, and with specific system configurations, the loss of power could lead to a CCW pump runout condition. The primary cause of this finding was related to the cross-cutting area of human performance.
Because this issue affects the NRC's ability to perform its regulatory function, it was evaluated using the traditional enforcement process. The inspectors determined that this violation is of very low safety significance and because the licensee entered the issue into their corrective action program (CAP026528), this violation is being treated as an NCV consistent with Section VI.A.1 of the NRC Enforcement Policy. The licensee has taken actions to revise plant procedures to address this issue.
Inspection Report# : 2005012(pdf)
 
3Q/2005 Inspection Findings - Kewaunee                                                                                                  Page 3 of 9 Significance:        Jul 29, 2005 Identified By: NRC Item Type: VIO Violation Potential Failure of Auxiliary Feedwater Pumps Due to Air Ingestion or During Runout Conditions The inspectors identified a finding that was preliminarily determined to be of low to moderate safety significance, because Kewaunee failed to provide adequate design control to ensure the AFW pumps would be protected from failure due to air ingestion during tornado or seismic events; as well as from failure during potential runout conditions. The finding is also an apparent violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for not effectively providing controls to check the adequacy of the design for protecting the AFW pumps during design and license basis events.
The finding was determined to be more than minor since it impacted Mitigating System cornerstone attributes of design control (initial design and plant modifications) and the cornerstone objective to ensure availability, reliability, and capability of the AFW system to respond to events to prevent core damage. A Significance Determination Process Phase 3 risk analysis determined that this finding was preliminarily of low to moderate safety significance. The licensee has taken significant corrective actions, including extensive modifications to the system.
On September 16, 2005, a final significance determination letter was issued for a WHITE finding (IR 2005014).
Inspection Report# : 2005010(pdf)
Inspection Report# : 2005014(pdf)
Significance:        Jul 29, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Evaluate the Effect of Modification on Turbine Driven AFW Pump Performance with Reduced Steam Pressure The inspectors identified a finding involving a Green Non-Cited Violation (NCV) of 10 CFR 50, Appendix B, Criterion III, "Design Control".
The finding involved the revision of AFW pump discharge pressure trip setpoints. The licensee had not determined if the turbine driven AFW (TDAFW) pump was capable of providing the required flow under reduced steam pressure conditions prior to approving the modification. This issue could have affected the performance of the AFW system under post accident conditions.
This issue was greater than minor because it potentially affected the Mitigating System cornerstone objective of equipment capability. The issue screened as very low safety significance in Phase 1 of the SDP, because it was a design deficiency that was not found to result in a loss of function and the item was resolved prior to being in the plant conditions where the finding could have impacted the pump's performance. The licensee conducted post modification tests and revised permanent plant procedures to ensure the TDAFW pump was capable of providing the required flow under reduced steam pressure conditions.
Inspection Report# : 2005010(pdf)
Significance:        Jun 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to minimize prior identified and predictable explosive gas concentrations in the WGDTs A finding of very low safety significance was identified by the inspectors for a NCV of Title 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action." The licensee failed to consider the impact on plant fire protection when ineffective resolution of waste gas system issues repeatedly led to explosive mixtures in the Waste Gas Decay Tanks. The licensee entered these issues into their corrective action program. The primary cause of this violation was related to the cross-cutting area of Problem Identification and Resolution. The licensee repeatedly encountered explosive gas levels in the WGDTs and were aware of plant conditions that resulted in these levels but failed to take adequate corrective actions to prevent explosive gas mixtures from developing in the WGDTs. The issue is more than minor because uncontrolled explosive mixtures in the WGDTs could have led to a more significant safety concern. The issue was of very low safety significance because explosive mixtures were only present during plant shutdown conditions; an explosion would not have affected safe shutdown equipment (i.e.
Residual Heat Removal System); the explosive mixture conditions were only present for short periods of time (<12 hours); and the tanks were isolated and vented per procedure when discovered.
Inspection Report# : 2005008(pdf)
Significance:        Feb 18, 2005 Identified By: NRC Item Type: FIN Finding Lack of 4160 Vac Bus 1-5 Ovewrcurrent and Loss of Voltage Relay Coordination The team identified a finding of very low safety significance for a failure to provide adequate electrical coordination of protective devices thereby ensuring that postulated electrical faults would be isolated upon detection. Specifically, the team identified that the lack of adequate electrical systems coordination between the undervoltage and overcurrent protection on 4160 Vac safety bus 1-5 would result in the loss of voltage relays actuating before the bus over-current relays. This design deficiency results in the failure to lock out safety bus 1- 5 upon postulated electrical faults and subjects the postulated faulted safety bus 1-5 to be re-energized via an alternate offsite source. This design introduced a challenge to the safety equipment availability and reliability. The licensee planned to develop changes to the affected relays.
 
3Q/2005 Inspection Findings - Kewaunee                                                                                                Page 4 of 9 The finding was more than minor because the failure to provide adequate electrical coordination of electrical devices provided an unnecessary challenge to safety-related equipment, and if left uncorrected, could become a more safety significant concern. The finding was of very low safety significance because it was a design deficiency that did not result in the loss of system function. The finding was a not a violation of regulatory requirements.
Inspection Report# : 2005002(pdf)
Significance:        Feb 18, 2005 Identified By: NRC Item Type: NCV NonCited Violation Short Circuit Duty of Buses Exceeded - Impact on Safe Shutdown Analysis The team identified a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion II, "Quality Assurance Program," for a failure to identify potentially adverse conditions to the plant's fire protection safe shutdown analysis caused by known overduty conditions on non-safety related buses 1-1, 1-2, 1-3, and 1-4. While the overduty condition was known to have existed at least since 1992, the licensee never entered the issue into the plant's corrective action program, where a proper evaluation should have addressed 10 CFR Part 50, Appendix R, safe shutdown related effects. The licensee planned to continue efforts to identify additional evaluations and corrective actions.
This finding was more than minor, because it was associated with the degradation of a fire protection feature. The finding was of very low safety significance because after extensive evaluation of the deficiency, the licensee was able to determine that the plant could still safely shut down the plant during a postulated fire event.
Inspection Report# : 2005002(pdf)
Significance:        Feb 18, 2005 Identified By: NRC Item Type: NCV NonCited Violation Battery Sizing Deficiencies The team identified a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for failure to implement adequate design controls of documents, inputs, and assumptions in the design of the two safety-related batteries. Specifically, the licensee did not perform and control battery sizing calculations, including consideration of temperature effects, to ensure that the batteries maintained sufficient capacity to perform the intended design function. The team determined that the failure to appropriately evaluate effects of battery room and cell temperatures also affected the cross-cutting area of Problem Identification and Resolution because the subject of battery capacity versus battery temperature had been previously identified in a 1992 NRC inspection. The licensee planned to perform battery sizing calculations as part of an overall electrical systems analysis improvement project.
This finding was more than minor because it affected the mitigating systems cornerstone objective of ensuring the availability and reliability of the 125 Volts direct current battery system to respond to initiating events to prevent undesirable consequences. The finding is of very low safety significance because the battery remained operable. The licensee planned to develop formal battery sizing calculations.
Inspection Report# : 2005002(pdf)
Significance: SL-IV Feb 18, 2005 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Evaluation of Procedure Changes to Address AFW Design Deficiencies The team identified a finding involving a Non-Cited Violation of 10 CFR 50.59, "Changes, Tests, and Experiments." The finding involved a failure to perform an adequate review of operations procedure changes in accordance with 10 CFR 50.59 associated with the operation of motor-operated valves for the auxiliary feedwater suction source from the service water system. The team determined that the licensee's approval of changes to Procedure E-0-05, with the introduction of adverse effects, and a determination that 10 CFR 50.59 was not applicable was a violation of 10 CFR 50.59. The licensee subsequently performed additional evaluations of the procedure changes.
Because the issue affected the NRC's ability to perform its regulatory function, this finding was evaluated with the traditional enforcement process. The finding was determined to be of very low safety significance since the design basis safety-related function of the AFW system, to remove reactor decay heat following a loss of normal feedwater, was not adversely affected. This was determined to be a Severity Level IV NCV of 10 CFR 50.59.
Inspection Report# : 2005002(pdf)
Significance:        Feb 18, 2005 Identified By: NRC Item Type: NCV NonCited Violation Lack of Allowance for Manual Actions in Establishing Setpoint to Transfer AFW Pump Suction Source The team identified a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for failure to establish the condensate storage tank (CST) level setpoint to transfer the auxiliary feedwater (AFW) pump suction supply from the CST to service water.
The team determined that the calculation setpoint did not include an allowance for the manual operator actions required by emergency operations procedures. The licensee revised the plant procedure to perform the operator actions earlier in the procedure.
 
3Q/2005 Inspection Findings - Kewaunee                                                                                                  Page 5 of 9 This finding was more than minor because it affected the mitigating systems cornerstone objective of equipment reliability, in that failure to align the AFW pump suctions to service water prior to the CSTs being depleted could have resulted in damage to the AFW pumps. The finding was determined to be of very low safety significance because it was a design deficiency that did not result in a loss of function.
Inspection Report# : 2005002(pdf)
Significance:        Feb 18, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Ensure that Calculation Assumption was Based on Valid Times for Manual Operator Actions The team identified a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control." The finding involved the condensate storage tank (CST) level setpoint to transfer the auxiliary feedwater (AFW) pump suction from the CSTs to service water. A calculation assumption stated that a flow would drain from the CSTs to the condenser for 10 minutes until the operators isolated the flow by closing manual valve MU-2A. The team determined that the actions could not be completed in the time assumed by the calculation. The licensee initiated corrective actions to revise the appropriate operations procedure and calculation.
This finding was greater than minor because it affected the mitigating system cornerstone objective of equipment reliability, in that failure to align the AFW pump suctions to service water prior to the CSTs being depleted could have resulted in damage to the AFW pumps. The finding was determined to be of very low safety significance because it was a design deficiency that was not found to result in a loss of function. The team concluded that it was unlikely that the operators would allow the CST level to reach the EOP setpoint without attempting to refill the tanks from other sources, and that the operators would be aware of the CST levels.
Inspection Report# : 2005002(pdf)
Significance:        Feb 18, 2005 Identified By: NRC Item Type: NCV NonCited Violation TSC DG Target Reliability Methodology Inadequate The team identified a Non-Cited Violation of 10 CFR 50.63, "Loss of All Alternating Current Power." The finding involved the failure to establish a target reliability for the plant's alternate power source consistent with the reliability approved by the NRC staff in the licensee's Station Blackout submittal for 10 CFR 50.63. The non-conservative target reliability employed by the licensee resulted in the failure of the licensee to increase efforts to restore the Technical Support Center (TSC) Diesel Generator (DG) to its approved target reliability at an earlier date. The licensee subsequently initiated a corrective action to change the TSC DG reliability methodology.
This finding was more than minor, because it affected the reliability of a support system required for the mitigation of an Station Blackout event. The finding was of very low safety significance, because the finding did not directly affect the immediate operability of the TSC DG.
Inspection Report# : 2005002(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Scaffolding Erected too Close to Safety-Related Equipment Required to be Operable A finding of very low safety significance was identified by the inspectors for a violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings." The finding was associated with the licensee's failure to adequately implement scaffold control requirements contained in Procedure GMP-127, "Requirements and Guidelines for Scaffold Construction and Inspection," which required that scaffolding be no closer than 2 inches from any safety-related equipment unless otherwise evaluated and approved by Engineering. Specifically, scaffolding was erected within 2 inches of safety-related piping for the Service Water outlet from the jacket water heat exchangers for Diesel Generator B, the piping for the Emergency Borate MOV (CVC-440), and Safety Injection Pump A, without engineering evaluation and approval. Upon discovery of this condition, the licensee took immediate action to bring all noted scaffolding problems into compliance with licensee procedures and initiated a CAP document for the issue.
The finding was more than minor because, if left uncorrected, the issue may have resulted in a more significant safety concern. Specifically, the failure of scaffolding having adequate spacing in the vicinity of safety-related equipment during a seismic event could result in damage to mitigating equipment. The finding was of very low safety significance because it did not result in the actual loss of the safety function of the train or system. The finding was a Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings."
Inspection Report# : 2004009(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify Inadequate Pre-Fire Strategies A finding of very low safety significance was identified by the inspectors for a violation of a fire protection License Condition. The inspectors identified that the licensee failed to identify pertinent information, such as the presence of compressed flammable gas cylinders, on a fire area
 
3Q/2005 Inspection Findings - Kewaunee                                                                                                  Page 6 of 9 strategy for fire brigade personnel. Once this issue was identified, the licensee revised the fire area strategy for the affected area.
The finding was more than minor because the failure to provide adequate warnings and guidance relating to hazards associated with compressed flammable gas cylinders in fire strategies could adversely impact fire fighting strategies used by the fire brigade in fighting a fire.
The finding was of very low safety significance due to extensive training provided to fire brigade members to deal with unexpected contingencies. The finding was a Non-Cited Violation of License Condition 2.C(3) which required that fire area strategies provide pertinent information to help the fire brigade to be better prepared for fire fighting within that area.
Inspection Report# : 2004009(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Non-conforming Condition on the Safety-Related Containment Sump A finding of very low safety significance was identified by the inspectors for a violation of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Actions." The original licensing and design basis of the containment sump screens was to prevent any particles greater than 1/8 inch from entering the sump. The inspectors determined that the screen size was 1/8-inch by 15/32-inch which allowed particles greater than 1/8-inch to enter the sump. The inspectors subsequently determined that this issue had been identified and entered into the licensee's corrective action program in 1997. However, adequate corrective actions were not taken to correct this condition adverse to quality. Once this issue was identified, the licensee conducted an operability determination and concluded that there were no immediate operability issues with the containment sump. The licensee determined that the sump screens were nonconforming in accordance with Generic Letter 91-18, and planned long term corrective actions to be developed in conjunction with the resolution of Generic Safety Issue 191 and NRC Generic Letter 2004-02.
The inspectors concluded that the primary cause of this finding was related to the performance characteristic of corrective actions in the cross-cutting area of problem identification and resolution.
This finding was more than minor because the issue affected the Mitigating System cornerstone attribute of design control for initial design and equipment performance reliability and affected the associated cornerstone objective to ensure the reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was of very low safety significance because it was not a design or qualification deficiency that has been confirmed to result in a loss of function per Generic Letter 91-18. This finding was a Non-Cited Violation of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Actions."
Inspection Report# : 2004009(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Instructions and Procedures for Inspections and Cleaning of the Safety-Related Containment Sump A finding of very low safety significance was identified by the inspectors for a violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, And Drawings," regarding licensee instructions and procedures for containment sump inspections. Specifically, the inspectors identified that current licensee procedures did not require inspection or cleaning when boric acid or small debris may be present in the containment sump. The licensee's procedures for containment coatings did not require inspection of the coating located inside the containment sump which had not been inspected since initial application; and the licensee's procedure for containment sump gap inspections did not specify acceptance criteria to ensure this activity was satisfactorily accomplished. The licensee subsequently initiated several corrective actions to address these issues which included, but are not limited to: immediate inspection and cleaning of the safety-related containment sump; immediate inspection and assessment of the safety-related sump concrete coating; revision of preventive maintenance activities to require inspection and cleaning of the safety-related containment sump every refueling outage; revision of procedures to include inspection of the safety-related containment sump concrete coating every refueling outage; and revision of procedures to include appropriate acceptance criteria for determining that important activities were satisfactorily accomplished.
This finding was more than minor because if left uncorrected the finding could become a more significant safety concern and the issue affected the Mitigating System cornerstone attributes of equipment performance reliability and procedure quality and affected the associated cornerstone objective to ensure the reliability and capability of systems that respond to initiating events to prevent undesirable consequences.
The finding was of very low safety significance because it was not a design or qualification deficiency that has been confirmed to result in a loss of function per Generic Letter 91-18. This finding was a Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings."
Inspection Report# : 2004009(pdf)
Barrier Integrity Significance:        Mar 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation
 
3Q/2005 Inspection Findings - Kewaunee                                                                                                Page 7 of 9 Reactor Operation Above LIcensed Power Limit A finding of very low safety significance associated with a Non-Cited Violation of the plant operating license was self-revealed during normal plant operations. The Kewaunee Nuclear Power Plant Facility Operating License, as amended stated, "The Nuclear Management Company (NMC) is authorized to operate the facility at steady-state reactor core power levels not in excess of 1772 megawatts (thermal)." Contrary to this, on January 31, 2005, the 8-hour average thermal power peaked at 1772.07 MWt before being restored to below 1772 MWt. Reactor power was allowed to rise above 1772 MWt because the 8-hr average reactor thermal power indicator on the plant process computer system was not reliable, and the site operating philosophy allowed the 1-minute average and the 15-minute average reactor thermal power indications to exceed 1772 Mwt. Once the 8-hour average was discovered to be in excess of that allowed in the Operating License, operators immediately lowered power to within the licensed limit and entered this issue into the corrective action program. This violation of the plant operating license was considered greater than minor, because it could affect the barrier integrity cornerstone objective of protecting the integrity of the fuel cladding and was associated with the barrier integrity cornerstone attributes of thermal limits and reactivity control. The finding also involved the crosscutting area of human performance. In accordance with Inspection Manual Chaper (IMC) 0609, Appendix A, Phase 1, the finding was of very low safety significance.
Inspection Report# : 2005003(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: VIO Violation Inability to Close Containment Equipment Hatch The inspectors identified an apparent violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, And Drawings," having potential safety significance greater than green. The finding was associated with the licensee's inability to close the containment equipment hatch in an expeditious manner while the plant was in the refueling shutdown mode, fuel was in the reactor vessel, the time to boil was estimated to be less than 30 minutes, and the reactor coolant system was open to the containment atmosphere. The inability to close the containment equipment hatch was caused by a design error in a large steel rail system installed inside the containment which was to be used to bring heavy equipment into the containment. This large steel rail system obstructed closure of the containment equipment hatch. The inability to close the hatch in an expeditious manner violated the licensee's procedure requirements to do so.
This finding was more than minor because it affected the Barrier Integrity Cornerstone objective and was associated with the Barrier Integrity Cornerstone attribute of containment boundary preservation. Since this finding was determined to be potentially greater than Green using the SDP Phase 2 Process, this finding is of a to-be-determined (TBD) safety significance pending review by the NRC Significance Determination Process/Enforcement Review Panel (SERP).
On May 5, 2005, a final significance determination letter was issued for a WHITE finding (IR 2005009).
On September 20, 2005, supplemental inspection 95001 was completed. The licensee conducted and adequate root cause analysis and this issue is closed.
Inspection Report# : 2004009(pdf)
Inspection Report# : 2005009(pdf)
Inspection Report# : 2005015(pdf)
Significance:        Dec 31, 2004 Identified By: NRC Item Type: NCV NonCited Violation Reactor Building Ventilation Isolation Function Not Available When Required A finding of very low safety significance associated with Technical Specification 3.8 a.1.b., "Refueling Operations - Containment Closure,"
was self-revealed during required daily surveillance testing of reactor building ventilation system isolation. During the surveillance test, plant operators discovered that radiation monitors would not cause a Reactor Building Ventilation System Isolation to occur as designed. The cause of this failure was that other engineered safeguards testing was in progress that disabled the Reactor Building Ventilation System Isolation function, which was required to be operable at the time. Once this issue was identified, the licensee restored the Reactor Building Ventilation System and entered this issue into the corrective action program.
This finding was more than minor, because it represented a degradation of the Barrier Integrity Cornerstone objective and was associated with Barrier Integrity cornerstone attribute of safety system and component performance. The finding was of very low safety significance because it did not result in the actual release of radioactive material. This finding was a Non-Cited Violation of Plant Technical Specification 3.8.a.1.b.,
"Refueling Operations-Containment Closure."
Inspection Report# : 2004009(pdf)
Emergency Preparedness
 
3Q/2005 Inspection Findings - Kewaunee                                                                                              Page 8 of 9 Occupational Radiation Safety Significance:        Sep 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Post and Control Access Into a Locked High Radiation Area During Radiographic Activities A self-revealed finding of very low safety significance and an associated Non-Cited Violation of NRC requirements were identified when a high radiation area boundary was breached by two workers during radiography. An unnecessary radiation exposure could have been received by the workers had they not been stopped by radiography personnel as they moved toward the exposed radiographic source.
The issue 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 adequate protection of worker health and safety from exposure to radiation. The issue represents a finding of very low safety significance because there was no overexposure or substantial potential for an overexposure given the actual radiological conditions in the area coupled with the duration of the radiographic operation and the presence of radiography personnel who provided surveillance of the area, nor was the licensee's ability to assess worker dose compromised. A Non-Cited Violation of Technical Specification 6.13(a) and 10 CFR 20.1601(b) was identified for the failure to comply with the RP requirements that govern the control of access into high radiation areas. Corrective actions taken by the licensee included enhanced administrative measures by revising the radiography procedure and counseling of involved staff. Since the cause of the problem included corrective action deficiencies from previous similar radiography boundary control events, the finding also relates to the cross-cutting area of problem identification and resolution.
Inspection Report# : 2005012(pdf)
Significance:        Sep 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Control Access Into a High Radiation Area During Radiographic Activities A self-revealed finding of very low safety significance and an associated Non-Cited Violation of NRC requirements were identified for an unposted/uncontrolled locked high radiation area in the turbine building during radiography activities. A radiography source created radiation levels such that a major portion of the whole body could have received in one hour a dose in excess of 1000 mrem in accessible areas of the turbine building, which were not posted or controlled in accordance with regulatory requirements. The areas with elevated dose rates were not positively controlled by locked door/gate, use of a barrier and flashing light, or maintained under continuous visual or electronic surveillance.
The issue 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 adequate protection of worker health and safety from exposure to radiation. The issue represents a finding of very low safety significance because there was no overexposure or substantial potential for an overexposure given the actual radiological conditions in the uncontrolled areas coupled with the duration of the radiographic shot. A Non-Cited Violation of Technical Specification 6.13(b) and 10 CFR 20.1601(b) was identified for the failure to comply with the RP requirements that govern the control of access into locked high radiation areas. Corrective actions taken by the licensee included enhanced administrative measures by revising the radiography procedure, and counseling and training of RP staff.
Inspection Report# : 2005012(pdf)
Significance:        Sep 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Control Access Into a High Radiation Area While Moving a Radioactive Filter A self-revealed finding of very low safety significance and two associated Non-Cited Violations of regulatory requirements were identified for an unposted and uncontrolled high radiation area in an auxiliary building elevator during the transfer of a radioactive seal water injection filter.
As a result of this failure, workers could have unknowingly entered a high radiation area in the elevator without knowledge of the radiological conditions.
The issue 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 adequate protection of worker health and safety from exposure to radiation. The issue represents a finding of very low safety significance because there was no overexposure or substantial potential for an overexposure given the circumstances and the actual radiological conditions in the area, nor was the licensee's ability to assess worker dose compromised. A Non-Cited Violation of Technical Specification 6.13(a) and 10 CFR 20.1601(b) was identified for the failure to comply with the RP requirements that govern the control of access into high radiation areas. This issue also represents a Non-Cited Violation of 10 CFR 20.1902(b)/20.1903 for failure to post a high radiation area. Corrective actions taken by the licensee included enhanced administrative measures (RP Job Guide) for change-out and transport of all radioactive filters.
Inspection Report# : 2005012(pdf)
Significance:        Sep 30, 2005 Identified By: NRC
 
3Q/2005 Inspection Findings - Kewaunee                                                                                              Page 9 of 9 Item Type: NCV NonCited Violation Failure to Post a High Radiation Area While Moving a Radioactive Filter A self-revealed finding of very low safety significance and two associated Non-Cited Violations of regulatory requirements were identified for an unposted and uncontrolled high radiation area in an auxiliary building elevator during the transfer of a radioactive seal water injection filter.
As a result of this failure, workers could have unknowingly entered a high radiation area in the elevator without knowledge of the radiological conditions.
The issue 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 adequate protection of worker health and safety from exposure to radiation. The issue represents a finding of very low safety significance because there was no overexposure or substantial potential for an overexposure given the circumstances and the actual radiological conditions in the area, nor was the licensee's ability to assess worker dose compromised. A Non-Cited Violation of Technical Specification 6.13(a) and 10 CFR 20.1601(b) was identified for the failure to comply with the RP requirements that govern the control of access into high radiation areas. This issue also represents a Non-Cited Violation of 10 CFR 20.1902(b)/20.1903 for failure to post a high radiation area. Corrective actions taken by the licensee included enhanced administrative measures (RP Job Guide) for change-out and transport of all radioactive filters.
Inspection Report# : 2005012(pdf)
Public Radiation Safety Physical Protection Physical Protection information not publicly available.
Miscellaneous Last modified : November 30, 2005
 
4Q/2005 Inspection Findings - Kewaunee                                                                                                Page 1 of 8 Kewaunee 4Q/2005 Plant Inspection Findings Initiating Events Significance:        Jun 30, 2005 Identified By: NRC Item Type: FIN Finding Inadequate controls for loose material in substation A finding of very low safety significance was identified by the inspectors for failure to control loose materials in the protected area and substation. No violation of NRC requirements occurred. Once identified, the licensee initiated a condition report (CAP) to develop a surveillance procedure to remove loose materials before summer months where potential adverse weather was apparent.
The issue was more than minor because, if left uncontrolled, the loose items adjacent to the auxiliary transformers and in the substation would become a more significant safety concern. The issue was of very low safety significance because the finding did not contribute to the likelihood of a primary or secondary system loss of coolant accident initiator; the finding did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available; and the finding did not increase the likelihood of a fire or internal or external flooding. The issue was not considered a violation of regulatory requirements because it did not affect safety-related structures, systems, or components.
Inspection Report# : 2005008(pdf)
Significance:        Jun 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to manage risk duing periods where the grid condition was defined as unstable.
A finding of very low safety significance was identified by the inspectors for a Non-Cited Violation (NCV) of Title 10 CFR Part 50.65(a)(4).
The licensee failed to adequately assess shutdown risk during degraded grid conditions. Once identified, the licensee initiated a CAP to modify shutdown safety assessment and operating procedures to include grid conditions in risk assessments. The finding was more than minor because the licensee's risk assessment had incorrect assumptions that had the potential to change the outcome of the assessment. The inspectors determined that the finding could not be evaluated using the Significance Determination Process because the finding was associated with an inadequate qualitative risk assessment. The inspectors determined that this issue was of very low safety significance which was verified by the regional branch chief.
Inspection Report# : 2005008(pdf)
Significance:        Feb 18, 2005 Identified By: NRC Item Type: FIN Finding Safety Buses Relay Sensitivity to External Electrical Distubrances The team identified a finding of very low safety significance for a failure to provide adequate relay setpoint calibration tolerances on safety buses 1-5 and 1-6 loss of voltage relays. The existing relay setting calibration tolerances would have allowed the loss of voltage relays to actuate spuriously during certain offsite electrical system disturbances and un-necessarily separate the safety buses from the offsite power system and result in a plant transient. The licensee implemented corrective actions to revise the appropriate loss of voltage relay surveillance procedures.
The finding was more than minor because the failure to provide adequate relay setting tolerances could result in an unnecessary separation of the safety buses from the electrical grid and an ensuing plant transient. The finding was of very low safety significance because the issue would not preclude the safety buses from being re-energized by the emergency power sources. The finding was a not a violation of regulatory requirements.
Inspection Report# : 2005002(pdf)
Significance:        Feb 18, 2005 Identified By: NRC Item Type: NCV NonCited Violation Operator Actions Following Station Blackout - Lack of Procedure Guidance The team identified a Non-Cited Violation of 10 CFR 50.63, "Loss of All Alternating Current Power," for a failure to maintain procedural steps that minimized the likelihood and duration of a Station Blackout (SBO) event. The deleted procedural steps allowed for the cross-connection of the plant's two redundant safety buses should both the Reserve Auxiliary Transformer and the 1B Emergency Diesel Generator fail. These
 
4Q/2005 Inspection Findings - Kewaunee                                                                                                Page 2 of 8 procedural steps, as originally employed, served to lessen the likelihood of the SBO occurring, and/or reduce the time of the SBO. The licensee implemented corrective actions to revise the appropriate operations procedure.
This finding was more than minor, because it was associated with the likelihood of an initiating event and the reliability of a safety bus that responds to an initiating event. The finding was of very low safety significance, because multiple sources of both onsite and offsite power remained available to supply the two safety buses.
Inspection Report# : 2005002(pdf)
Mitigating Systems Significance:        Dec 16, 2005 Identified By: NRC Item Type: FIN Finding No Trending of Adverse Conditions Identified During Outages The inspectors identified a finding of very low safety significance for the licensee not reviewing corrective action program documents (CAPs) during outages for potential trends of conditions adverse to quality. As part of the screening process of CAPs, the licensee assigned, as possible, CAPs to various "hot buttons." Hot buttons were searchable categories in the corrective action program computer system that had been established for various problems, such as equipment tagging errors, security door control, and reactivity management. For non-outage times, the licensee assigned a monthly number of hits for each hot button that, if exceeded for 3 months in succession, would result in the generation of a CAP to investigate a possible trend. However, as of December 16, 2005, the licensee did not use hot button action levels during outages when the number of CAPs written was much higher than during non-outage times.
This finding is greater than minor because if left uncorrected would become a more significant safety concern. This finding is not suitable for Significance Determination Process evaluation, but has been reviewed by NRC management and is determined to be a finding of very low safety significance. No violation of regulatory requirements occurred. The cause of the finding is related to the cross-cutting element of problem identification and resolution, because of not identifying potential conditions adverse to quality through trending of CAPs during outages.
Inspection Report# : 2005005(pdf)
Significance:        Dec 16, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct Procedure Non-Adherence The inspectors identified a finding of very low safety significance and a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," for the failure to take corrective action for procedure non-compliance identified during the licensee's 2004 self-assessment of the corrective action program. As a result of the assessment, CAP025194, "Corrective Action Program Procedure and Guidance Document Use," was written and documented that plant workers were not following corrective action program procedures for apparent cause evaluations and root cause evaluations, effectiveness review content, priority and due date assignments, initiator feedback, and documentation of corrective action completion. To correct this problem, corrective action CA018094, "Corrective Action Program Procedure and Guidance Document Use," was written and specified one or 2 weeks of requiring "in-hand" use by the plant staff of the corrective action program administrative procedure. However, completion of this action was delayed several times and on July 25, 2005, CAP025194 and CA018094 were closed with the only documented action taken being a July 18, 2005, meeting of the station human performance steering committee at which the licensee decided not to take action because of the pending transition to the corrective action program documents of the plant's new owner.
This finding is greater than minor because if left uncorrected would become a more significant safety concern. This finding is not suitable for Significance Determination Process evaluation, but has been reviewed by NRC management and is determined to be a finding of very low safety significance. The cause of the finding is related to the cross-cutting element of problem identification and resolution, because of the failure to take corrective action for non-adherence to station procedures.
Inspection Report# : 2005005(pdf)
Significance:        Dec 16, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Correct Residual Heat Removal Pump Seal Leakage A finding of very low safety significance that was a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action,"
was identified for the licensee's ineffective corrective action to repair a leak on the seal of the "B" residual heat removal (RHR) pump. The leak was identified on November 2, 2005, when the pump was stopped following the performance of a required surveillance. The leak rate exceeded leakage control program limits. A similar leak was identified on June 16, 2004, for which the licensee replaced the seal in November 2004.
This finding is greater than minor because it was associated with the "RCS (reactor coolant system) equipment and barrier performance"
 
4Q/2005 Inspection Findings - Kewaunee                                                                                                Page 3 of 8 attribute of the barrier integrity cornerstone and does affect the cornerstone objective of providing reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events.
Although the RCS barrier was affected, the finding did not affect the mitigation capability of the RHR system and did not contribute to the likelihood of a primary or secondary system loss of coolant accident initiator or affect the containment integrity. Therefore, the finding is of very low safety significance.
Inspection Report# : 2005005(pdf)
Significance:        Oct 06, 2005 Identified By: NRC Item Type: VIO Violation Potential Flooding in the Turbine Building Basement A review of design drawings by the inspectors revealed a direct piping connection from the turbine building sump to the trench in safeguards alley. The inspectors determined that there were no check valves located in the piping to prevent water spills in the turbine building basement from backing up into the safeguards alley. The inspectors also noted that no flood barriers specifically designed to protect equipment in the safeguards alley from flooding in the turbine building basement were installed. The inspectors requested additional information from the licensee regarding potential flooding events occurring in the safeguards alley. The licensee documented its response to the inspectors information request in Condition Evaluation (CE) 014653. This CE stated that it would take approximately 3 hours for flooding caused by AFW pump discharge to affect safety-related equipment, and such flooding could be mitigated by opening doors between the safeguards alley and the turbine building basement. The CE also stated that other sources of flooding in the turbine building basement need not be considered since such flooding events are outside the design basis of the plant.
The inspectors identified a finding that was preliminarily determined to be of substantial to high safety significance because the licensee failed to provide adequate design control to ensure that Class I equipment was protected against damage from the rupture of a pipe or tank resulting in serious flooding or excessive steam release to the extent that the Class I equipment's function is impaired. Specifically, the design of Kewaunee Power Station (KPS) did not ensure that the auxiliary feedwater (AFW) pumps, the 480-volt (V) safeguards buses, the safe shutdown panel, emergency diesel generators (EDGs) 1A and 1B, and 4160-V safeguards buses 1-5 and 1-6 would be protected from random or seismically induced failures of non-Class I systems in the turbine building. The finding is also an apparent violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for not ensuring that the design of KPS prevented turbine building flooding from impacting multiple safety related equipment trains needed for safe shutdown of the plant. The inspectors determined that a primary cause of this finding was related to the cross-cutting area of Problem Identification and Resolution, because there was an earlier opportunity to discover and correct this issue based on the licensee's 2003 experience when minor flooding from the turbine building had challenged safety equipment located adjacent to the turbine building basement.
The finding was more than minor because it impacted Mitigating Systems cornerstone attributes of design control (initial design and plant modifications) and protection against external factors (internal flood hazards and seismic events) and it impacted the Mitigating Systems cornerstone objective to ensure availability, reliability and capability of multiple trains of safety related equipment to respond to events to prevent core damage. A Significance Determination Process Phase 3 risk analysis determined that this finding was preliminarily of substantial to high safety significance. The licensee has taken significant corrective actions, including extensive system and structural modifications to address this issue.
After considering the information developed during the inspection, and the additional information you provided prior to, during, and in response to our questions at the Regulatory Conference, the NRC has concluded the inspection finding is appropriately characterized as Yellow (i.e., an issue with substantial importance to safety, that will result in additional NRC inspection and potentially other NRC action).
Inspection Report# : 2004009(pdf)
Inspection Report# : 2005002(pdf)
Inspection Report# : 2005011(pdf)
Inspection Report# : 2005018(pdf)
Significance: SL-IV Sep 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Report in a Timely Manner an Unanalyzed Condition Involving a Potential Runout Concern With the CCW Pumps The inspectors identified a Non-Cited Violation (NCV) when the licensee failed to make a written report, within 60 days, to the NRC in accordance with 10 CFR 50.73(a)(2)(ii)(B), when an unanalyzed condition that significantly degraded plant safety was identified. Specifically, the licensee did not recognize the significance of a previously identified condition involving a potential runout issue with the component cooling water (CCW) pumps, and did not report this condition until the inspectors identified the requirement. The concern related to the CCW pump capability to provide required flow under certain conditions. Specifically, during a loss of power, and with specific system configurations, the loss of power could lead to a CCW pump runout condition. The primary cause of this finding was related to the cross-cutting area of human performance.
Because this issue affects the NRC's ability to perform its regulatory function, it was evaluated using the traditional enforcement process. The inspectors determined that this violation is of very low safety significance and because the licensee entered the issue into their corrective action program (CAP026528), this violation is being treated as an NCV consistent with Section VI.A.1 of the NRC Enforcement Policy. The licensee has taken actions to revise plant procedures to address this issue.
Inspection Report# : 2005012(pdf)
 
4Q/2005 Inspection Findings - Kewaunee                                                                                                Page 4 of 8 Significance:        Aug 16, 2005 Identified By: NRC Item Type: VIO Violation Potential Common Mode Failure of Auxiliary Feedwater URI 05000305/2005002-05 is associated with the design of the AFW pump's discharge pressure switches. The inspectors identified the potential for air intrusion into operating AFW pumps, potentially resulting in a common mode failure of the AFW system. This could occur during certain events where the suction source is lost prior to being able to manually swap the source of water from the CST to the SW system.
The inspectors identified a finding that was preliminarily determined to be of low to moderate safety significance, because Kewaunee failed to provide adequate design control to ensure the AFW pumps would be protected from failure due to air ingestion during tornado or seismic events; as well as from failure during potential runout conditions. The finding is also an apparent violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for not effectively providing controls to check the adequacy of the design for protecting the AFW pumps during design and license basis events.
The finding was determined to be more than minor since it impacted Mitigating System cornerstone attributes of design control (initial design and plant modifications) and the cornerstone objective to ensure availability, reliability, and capability of the AFW system to respond to events to prevent core damage. A Significance Determination Process Phase 3 risk analysis determined that this finding was preliminarily of low to moderate safety significance. The licensee has taken significant corrective actions, including extensive modifications to the system.
After considering the information developed during the inspection, the NRC has concluded the inspection finding is appropriately characterized as White (i.e., an issue with low to moderate increased importance to safety, which may require additional NRC inspections).
Inspection Report# : 2005002(pdf)
Inspection Report# : 2005010(pdf)
Inspection Report# : 2005014(pdf)
Significance:        Jul 29, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Evaluate the Effect of Modification on Turbine Driven AFW Pump Performance with Reduced Steam Pressure The inspectors identified a finding involving a Green Non-Cited Violation (NCV) of 10 CFR 50, Appendix B, Criterion III, "Design Control".
The finding involved the revision of AFW pump discharge pressure trip setpoints. The licensee had not determined if the turbine driven AFW (TDAFW) pump was capable of providing the required flow under reduced steam pressure conditions prior to approving the modification. This issue could have affected the performance of the AFW system under post accident conditions.
This issue was greater than minor because it potentially affected the Mitigating System cornerstone objective of equipment capability. The issue screened as very low safety significance in Phase 1 of the SDP, because it was a design deficiency that was not found to result in a loss of function and the item was resolved prior to being in the plant conditions where the finding could have impacted the pump's performance. The licensee conducted post modification tests and revised permanent plant procedures to ensure the TDAFW pump was capable of providing the required flow under reduced steam pressure conditions.
Inspection Report# : 2005010(pdf)
Significance:        Jun 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to minimize prior identified and predictable explosive gas concentrations in the WGDTs A finding of very low safety significance was identified by the inspectors for a NCV of Title 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action." The licensee failed to consider the impact on plant fire protection when ineffective resolution of waste gas system issues repeatedly led to explosive mixtures in the Waste Gas Decay Tanks. The licensee entered these issues into their corrective action program. The primary cause of this violation was related to the cross-cutting area of Problem Identification and Resolution. The licensee repeatedly encountered explosive gas levels in the WGDTs and were aware of plant conditions that resulted in these levels but failed to take adequate corrective actions to prevent explosive gas mixtures from developing in the WGDTs. The issue is more than minor because uncontrolled explosive mixtures in the WGDTs could have led to a more significant safety concern. The issue was of very low safety significance because explosive mixtures were only present during plant shutdown conditions; an explosion would not have affected safe shutdown equipment (i.e.
Residual Heat Removal System); the explosive mixture conditions were only present for short periods of time (<12 hours); and the tanks were isolated and vented per procedure when discovered.
Inspection Report# : 2005008(pdf)
Significance:        Feb 18, 2005 Identified By: NRC Item Type: FIN Finding Lack of 4160 Vac Bus 1-5 Ovewrcurrent and Loss of Voltage Relay Coordination The team identified a finding of very low safety significance for a failure to provide adequate electrical coordination of protective devices
 
4Q/2005 Inspection Findings - Kewaunee                                                                                                  Page 5 of 8 thereby ensuring that postulated electrical faults would be isolated upon detection. Specifically, the team identified that the lack of adequate electrical systems coordination between the undervoltage and overcurrent protection on 4160 Vac safety bus 1-5 would result in the loss of voltage relays actuating before the bus over-current relays. This design deficiency results in the failure to lock out safety bus 1- 5 upon postulated electrical faults and subjects the postulated faulted safety bus 1-5 to be re-energized via an alternate offsite source. This design introduced a challenge to the safety equipment availability and reliability. The licensee planned to develop changes to the affected relays.
The finding was more than minor because the failure to provide adequate electrical coordination of electrical devices provided an unnecessary challenge to safety-related equipment, and if left uncorrected, could become a more safety significant concern. The finding was of very low safety significance because it was a design deficiency that did not result in the loss of system function. The finding was a not a violation of regulatory requirements.
Inspection Report# : 2005002(pdf)
Significance:        Feb 18, 2005 Identified By: NRC Item Type: NCV NonCited Violation Short Circuit Duty of Buses Exceeded - Impact on Safe Shutdown Analysis The team identified a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion II, "Quality Assurance Program," for a failure to identify potentially adverse conditions to the plant's fire protection safe shutdown analysis caused by known overduty conditions on non-safety related buses 1-1, 1-2, 1-3, and 1-4. While the overduty condition was known to have existed at least since 1992, the licensee never entered the issue into the plant's corrective action program, where a proper evaluation should have addressed 10 CFR Part 50, Appendix R, safe shutdown related effects. The licensee planned to continue efforts to identify additional evaluations and corrective actions.
This finding was more than minor, because it was associated with the degradation of a fire protection feature. The finding was of very low safety significance because after extensive evaluation of the deficiency, the licensee was able to determine that the plant could still safely shut down the plant during a postulated fire event.
Inspection Report# : 2005002(pdf)
Significance:        Feb 18, 2005 Identified By: NRC Item Type: NCV NonCited Violation Battery Sizing Deficiencies The team identified a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for failure to implement adequate design controls of documents, inputs, and assumptions in the design of the two safety-related batteries. Specifically, the licensee did not perform and control battery sizing calculations, including consideration of temperature effects, to ensure that the batteries maintained sufficient capacity to perform the intended design function. The team determined that the failure to appropriately evaluate effects of battery room and cell temperatures also affected the cross-cutting area of Problem Identification and Resolution because the subject of battery capacity versus battery temperature had been previously identified in a 1992 NRC inspection. The licensee planned to perform battery sizing calculations as part of an overall electrical systems analysis improvement project.
This finding was more than minor because it affected the mitigating systems cornerstone objective of ensuring the availability and reliability of the 125 Volts direct current battery system to respond to initiating events to prevent undesirable consequences. The finding is of very low safety significance because the battery remained operable. The licensee planned to develop formal battery sizing calculations.
Inspection Report# : 2005002(pdf)
Significance: SL-IV Feb 18, 2005 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Evaluation of Procedure Changes to Address AFW Design Deficiencies The team identified a finding involving a Non-Cited Violation of 10 CFR 50.59, "Changes, Tests, and Experiments." The finding involved a failure to perform an adequate review of operations procedure changes in accordance with 10 CFR 50.59 associated with the operation of motor-operated valves for the auxiliary feedwater suction source from the service water system. The team determined that the licensee's approval of changes to Procedure E-0-05, with the introduction of adverse effects, and a determination that 10 CFR 50.59 was not applicable was a violation of 10 CFR 50.59. The licensee subsequently performed additional evaluations of the procedure changes.
Because the issue affected the NRC's ability to perform its regulatory function, this finding was evaluated with the traditional enforcement process. The finding was determined to be of very low safety significance since the design basis safety-related function of the AFW system, to remove reactor decay heat following a loss of normal feedwater, was not adversely affected. This was determined to be a Severity Level IV NCV of 10 CFR 50.59.
Inspection Report# : 2005002(pdf)
Significance:        Feb 18, 2005 Identified By: NRC Item Type: NCV NonCited Violation
 
4Q/2005 Inspection Findings - Kewaunee                                                                                                  Page 6 of 8 Lack of Allowance for Manual Actions in Establishing Setpoint to Transfer AFW Pump Suction Source The team identified a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for failure to establish the condensate storage tank (CST) level setpoint to transfer the auxiliary feedwater (AFW) pump suction supply from the CST to service water.
The team determined that the calculation setpoint did not include an allowance for the manual operator actions required by emergency operations procedures. The licensee revised the plant procedure to perform the operator actions earlier in the procedure.
This finding was more than minor because it affected the mitigating systems cornerstone objective of equipment reliability, in that failure to align the AFW pump suctions to service water prior to the CSTs being depleted could have resulted in damage to the AFW pumps. The finding was determined to be of very low safety significance because it was a design deficiency that did not result in a loss of function.
Inspection Report# : 2005002(pdf)
Significance:        Feb 18, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Ensure that Calculation Assumption was Based on Valid Times for Manual Operator Actions The team identified a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control." The finding involved the condensate storage tank (CST) level setpoint to transfer the auxiliary feedwater (AFW) pump suction from the CSTs to service water. A calculation assumption stated that a flow would drain from the CSTs to the condenser for 10 minutes until the operators isolated the flow by closing manual valve MU-2A. The team determined that the actions could not be completed in the time assumed by the calculation. The licensee initiated corrective actions to revise the appropriate operations procedure and calculation.
This finding was greater than minor because it affected the mitigating system cornerstone objective of equipment reliability, in that failure to align the AFW pump suctions to service water prior to the CSTs being depleted could have resulted in damage to the AFW pumps. The finding was determined to be of very low safety significance because it was a design deficiency that was not found to result in a loss of function. The team concluded that it was unlikely that the operators would allow the CST level to reach the EOP setpoint without attempting to refill the tanks from other sources, and that the operators would be aware of the CST levels.
Inspection Report# : 2005002(pdf)
Significance:        Feb 18, 2005 Identified By: NRC Item Type: NCV NonCited Violation TSC DG Target Reliability Methodology Inadequate The team identified a Non-Cited Violation of 10 CFR 50.63, "Loss of All Alternating Current Power." The finding involved the failure to establish a target reliability for the plant's alternate power source consistent with the reliability approved by the NRC staff in the licensee's Station Blackout submittal for 10 CFR 50.63. The non-conservative target reliability employed by the licensee resulted in the failure of the licensee to increase efforts to restore the Technical Support Center (TSC) Diesel Generator (DG) to its approved target reliability at an earlier date. The licensee subsequently initiated a corrective action to change the TSC DG reliability methodology.
This finding was more than minor, because it affected the reliability of a support system required for the mitigation of an Station Blackout event. The finding was of very low safety significance, because the finding did not directly affect the immediate operability of the TSC DG.
Inspection Report# : 2005002(pdf)
Barrier Integrity Significance:        Mar 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Reactor Operation Above LIcensed Power Limit A finding of very low safety significance associated with a Non-Cited Violation of the plant operating license was self-revealed during normal plant operations. The Kewaunee Nuclear Power Plant Facility Operating License, as amended stated, "The Nuclear Management Company (NMC) is authorized to operate the facility at steady-state reactor core power levels not in excess of 1772 megawatts (thermal)." Contrary to this, on January 31, 2005, the 8-hour average thermal power peaked at 1772.07 MWt before being restored to below 1772 MWt. Reactor power was allowed to rise above 1772 MWt because the 8-hr average reactor thermal power indicator on the plant process computer system was not reliable, and the site operating philosophy allowed the 1-minute average and the 15-minute average reactor thermal power indications to exceed 1772 Mwt. Once the 8-hour average was discovered to be in excess of that allowed in the Operating License, operators immediately lowered power to within the licensed limit and entered this issue into the corrective action program. This violation of the plant operating license was considered greater than minor, because it could affect the barrier integrity cornerstone objective of protecting the integrity of the fuel cladding and was associated with the barrier integrity cornerstone attributes of thermal limits and reactivity control. The finding also involved the crosscutting area of human performance. In accordance with Inspection Manual Chaper (IMC) 0609, Appendix A, Phase 1, the finding was of very low safety significance.
 
4Q/2005 Inspection Findings - Kewaunee                                                                                              Page 7 of 8 Inspection Report# : 2005003(pdf)
Emergency Preparedness Occupational Radiation Safety Significance:        Sep 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Post and Control Access Into a Locked High Radiation Area During Radiographic Activities A self-revealed finding of very low safety significance and an associated Non-Cited Violation of NRC requirements were identified when a high radiation area boundary was breached by two workers during radiography. An unnecessary radiation exposure could have been received by the workers had they not been stopped by radiography personnel as they moved toward the exposed radiographic source.
The issue 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 adequate protection of worker health and safety from exposure to radiation. The issue represents a finding of very low safety significance because there was no overexposure or substantial potential for an overexposure given the actual radiological conditions in the area coupled with the duration of the radiographic operation and the presence of radiography personnel who provided surveillance of the area, nor was the licensee's ability to assess worker dose compromised. A Non-Cited Violation of Technical Specification 6.13(a) and 10 CFR 20.1601(b) was identified for the failure to comply with the RP requirements that govern the control of access into high radiation areas. Corrective actions taken by the licensee included enhanced administrative measures by revising the radiography procedure and counseling of involved staff. Since the cause of the problem included corrective action deficiencies from previous similar radiography boundary control events, the finding also relates to the cross-cutting area of problem identification and resolution.
Inspection Report# : 2005012(pdf)
Significance:        Sep 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Control Access Into a High Radiation Area During Radiographic Activities A self-revealed finding of very low safety significance and an associated Non-Cited Violation of NRC requirements were identified for an unposted/uncontrolled locked high radiation area in the turbine building during radiography activities. A radiography source created radiation levels such that a major portion of the whole body could have received in one hour a dose in excess of 1000 mrem in accessible areas of the turbine building, which were not posted or controlled in accordance with regulatory requirements. The areas with elevated dose rates were not positively controlled by locked door/gate, use of a barrier and flashing light, or maintained under continuous visual or electronic surveillance.
The issue 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 adequate protection of worker health and safety from exposure to radiation. The issue represents a finding of very low safety significance because there was no overexposure or substantial potential for an overexposure given the actual radiological conditions in the uncontrolled areas coupled with the duration of the radiographic shot. A Non-Cited Violation of Technical Specification 6.13(b) and 10 CFR 20.1601(b) was identified for the failure to comply with the RP requirements that govern the control of access into locked high radiation areas. Corrective actions taken by the licensee included enhanced administrative measures by revising the radiography procedure, and counseling and training of RP staff.
Inspection Report# : 2005012(pdf)
Significance:        Sep 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Control Access Into a High Radiation Area While Moving a Radioactive Filter A self-revealed finding of very low safety significance and two associated Non-Cited Violations of regulatory requirements were identified for an unposted and uncontrolled high radiation area in an auxiliary building elevator during the transfer of a radioactive seal water injection filter.
As a result of this failure, workers could have unknowingly entered a high radiation area in the elevator without knowledge of the radiological conditions.
The issue 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 adequate protection of worker health and safety from exposure to radiation. The issue represents a finding of very low safety significance because there was no overexposure or substantial potential for an overexposure given the
 
4Q/2005 Inspection Findings - Kewaunee                                                                                              Page 8 of 8 circumstances and the actual radiological conditions in the area, nor was the licensee's ability to assess worker dose compromised. A Non-Cited Violation of Technical Specification 6.13(a) and 10 CFR 20.1601(b) was identified for the failure to comply with the RP requirements that govern the control of access into high radiation areas. This issue also represents a Non-Cited Violation of 10 CFR 20.1902(b)/20.1903 for failure to post a high radiation area. Corrective actions taken by the licensee included enhanced administrative measures (RP Job Guide) for change-out and transport of all radioactive filters.
Inspection Report# : 2005012(pdf)
Significance:        Sep 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Post a High Radiation Area While Moving a Radioactive Filter A self-revealed finding of very low safety significance and two associated Non-Cited Violations of regulatory requirements were identified for an unposted and uncontrolled high radiation area in an auxiliary building elevator during the transfer of a radioactive seal water injection filter.
As a result of this failure, workers could have unknowingly entered a high radiation area in the elevator without knowledge of the radiological conditions.
The issue 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 adequate protection of worker health and safety from exposure to radiation. The issue represents a finding of very low safety significance because there was no overexposure or substantial potential for an overexposure given the circumstances and the actual radiological conditions in the area, nor was the licensee's ability to assess worker dose compromised. A Non-Cited Violation of Technical Specification 6.13(a) and 10 CFR 20.1601(b) was identified for the failure to comply with the RP requirements that govern the control of access into high radiation areas. This issue also represents a Non-Cited Violation of 10 CFR 20.1902(b)/20.1903 for failure to post a high radiation area. Corrective actions taken by the licensee included enhanced administrative measures (RP Job Guide) for change-out and transport of all radioactive filters.
Inspection Report# : 2005012(pdf)
Public Radiation Safety Physical Protection Physical Protection information not publicly available.
Miscellaneous Last modified : March 03, 2006
 
1Q/2006 Inspection Findings - Kewaunee                                                                                                Page 1 of 8 Kewaunee 1Q/2006 Plant Inspection Findings Initiating Events Significance:        Mar 31, 2006 Identified By: NRC Item Type: FIN Finding Failure to Control Loose Materials Within the Protected Area in Response to Adverse Weather Conditions A finding of very low safety significance was identified by the inspectors for the licensees failure to control loose materials within the protected area south of the transformer bays in response to adverse weather conditions. The material could have been blown into the transformers and initiate a transient. The primary cause of this finding was related to the cross-cutting area of problem identification and resolution for the failure to implement effective corrective actions in response to a similar, previous inspection finding (Inspection Report 05000305/2005008). No violation of regulatory requirements occurred.
The licensee entered this issue into its corrective action program and removed the loose material from the transformer bays.
The finding is more than minor because, if left uncorrected, the loose items would become a more significant safety concern by becoming missile hazards; thereby, increasing the likelihood of an initiating event. Additionally, the inspectors determined that this issue was associated with the procedure quality 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 because the station procedure used to control potential airborne material was too narrow in scope. The finding was of very low safety significance because the inspectors answered no to all the screening questions in the Significance Determination Process Phase 1 Screening Worksheet under the Initiating Events column.
Inspection Report# : 2006002(pdf)
Significance:        Mar 30, 2006 Identified By: NRC Item Type: FIN Finding Failure to Adequately Evaluate an Inoperative Indicating Lamp for a Turbine control Valve A finding of very low safety significance was identified by the inspectors for the failure to adequately evaluate an inoperative indicating lamp associated with the turbine control valves. The primary cause of this finding was attributed to the cross-cutting area of human performance because procedures were available, but not followed, that would have facilitated proper performance of the task.
The licensee entered this item into its corrective action program and reviewed open work orders, provided a status update to management, and increased communications of related expectations.
The finding is greater than minor because the failure to adequately evaluate deficient conditions, if left uncorrected, would become a more significant safety concern. The finding was of very low safety significance because the inspectors answered no to all the questions in the Significance Determination Process Phase 1 Screening Worksheet under the Initiating Events column.
Inspection Report# : 2006002(pdf)
Significance:        Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Inadequate startup procedure resulted in an inadvertent carbon-dioxide fire suppression discharge and declaration of a Notice of Unusual Event A finding of very low safety significance was self-revealed during two events when use of an inadequate plant prestartup procedure resulted in actuation of the CARDOX Carbon Dioxide system. A Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was identified for the failure to include adequate acceptance criteria in Procedure N-0-02-CLA, "Plant Prestartup Checklist".
The primary cause of this finding was related to the resource attribute in the cross-cutting area of Human Performance. The licensee failed to provide the operators with quality procedures containing criteria to know when the secondary plant was appropriately aligned.
The inspectors determined that the finding was greater than minor because it involved the configuration control, human performance, and procedure quality attributes of the Initiating Events Cornerstone. Additionally the finding affected the cornerstone objective of limiting the likelihood of those events that upset plant stability during power operations. Specifically, an incorrect lineup could exist in the secondary system resulting in an initiating event, or an unanalyzed secondary system response after a trip. The issue was of very low safety significance because the finding did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would
 
1Q/2006 Inspection Findings - Kewaunee                                                                                                  Page 2 of 8 not be available. Corrective actions taken by the licensee include procedural enhancements to ensure that systems are lined up properly before continuing with plant startup.
Inspection Report# : 2005017(pdf)
Significance:        Jun 30, 2005 Identified By: NRC Item Type: FIN Finding Inadequate controls for loose material in substation A finding of very low safety significance was identified by the inspectors for failure to control loose materials in the protected area and substation. No violation of NRC requirements occurred. Once identified, the licensee initiated a condition report (CAP) to develop a surveillance procedure to remove loose materials before summer months where potential adverse weather was apparent.
The issue was more than minor because, if left uncontrolled, the loose items adjacent to the auxiliary transformers and in the substation would become a more significant safety concern. The issue was of very low safety significance because the finding did not contribute to the likelihood of a primary or secondary system loss of coolant accident initiator; the finding did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available; and the finding did not increase the likelihood of a fire or internal or external flooding. The issue was not considered a violation of regulatory requirements because it did not affect safety-related structures, systems, or components.
Inspection Report# : 2005008(pdf)
Significance:        Jun 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to manage risk duing periods where the grid condition was defined as unstable.
A finding of very low safety significance was identified by the inspectors for a Non-Cited Violation (NCV) of Title 10 CFR Part 50.65(a)(4).
The licensee failed to adequately assess shutdown risk during degraded grid conditions. Once identified, the licensee initiated a CAP to modify shutdown safety assessment and operating procedures to include grid conditions in risk assessments. The finding was more than minor because the licensee's risk assessment had incorrect assumptions that had the potential to change the outcome of the assessment. The inspectors determined that the finding could not be evaluated using the Significance Determination Process because the finding was associated with an inadequate qualitative risk assessment. The inspectors determined that this issue was of very low safety significance which was verified by the regional branch chief.
Inspection Report# : 2005008(pdf)
Mitigating Systems Significance:        Mar 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation Ineffective Corrective Actions to Resolve Boric Acid Leakage from the 1A RHR Pump Flange Studs and Nuts A finding of very low safety significance and an associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified by the inspectors for ineffective identification and the initiation of corrective actions to resolve boric acid leakage from the 1A residual heat removal (RHR) pump flange studs and nuts. The primary cause of this finding was attributed to the cross-cutting area of problem identification and resolution. During a review of corrective actions associated with the licensees identification of a moderate amount of boric acid around various pump flange studs and nuts, the inspectors found that numerous prior occasions existed where the licensee had identified similar conditions yet failed to adequately identify and initiate actions to evaluate or correct this condition adverse to quality.
The licensee entered this item into its corrective action program and wrote a work order to replace the pump casing flange gasket.
The finding is greater than minor because it is associated with the equipment performance attribute of the Mitigating System 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). Additionally, failure to correct a condition adverse to quality in a safety-related system, if left uncorrected, would become a more significant safety concern. The finding was of very low safety significance because the inspectors answered no to all the screening questions in the Significance Determination Process Phase 1 Screening Worksheet under the Mitigating Systems column.
Inspection Report# : 2006002(pdf)
Significance:        Mar 31, 2006 Identified By: NRC
 
1Q/2006 Inspection Findings - Kewaunee                                                                                                  Page 3 of 8 Item Type: NCV NonCited Violation Failure to Apply Appropriate Quality Classification to TSC Diesel Generator Modifications as Required by Procedures A finding of very low safety significance and an associated non-cited violation of the Kewaunee Technical Specifications, Section 6.8, Procedures, was identified by the inspectors during a review of plant modification Design Change Request 3490, which replaced the existing Technical Support Center diesel generator fuel oil day tank level switches with new level switches of a different design. The inspectors determined that, in accordance with procedure GNP-01.01.01, Determination of Nuclear Safety Designed Classifications, QA [Quality Assurance] Type and EQ [Environmental Qualification] Type, the new level switches should have been designated as Augmented Quality.
Contrary to this, the new switches were not designated as augmented quality. The primary cause of this finding was attributed to the cross-cutting area of problem identification and resolution because of the licensees failure to take effective corrective actions for previously identified problems with its quality assurance program.
The licensee entered this item into its corrective action program and conducted supplemental audits of quality-designated equipment, added additional related elements to an upcoming quality assurance group audit of the quality assurance program, and the conduct of a cause evaluation of related issues.
The finding is greater than minor because it is 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). Additionally, failure to comply with the provisions of nuclear safety-related procedures, if left uncorrected, would become a more significant safety concern. The finding is of very low safety significance because the inspectors answered no to all the screening questions in the Significance Determination Process Phase 1 Screening Worksheet under the Mitigating Systems column.
Inspection Report# : 2006002(pdf)
Significance:        Mar 30, 2006 Identified By: NRC Item Type: FIN Finding Failure to Adequately Evaluate the Extent-of-Condition of Degraded Fuses in Installed Equipment A finding of very low safety significance was identified by the inspectors for the failure to adequately evaluate the extent-of-condition relative to installed equipment for a 10 CFR Part 21 notification for degraded Bussmann fuses. The primary cause of the finding was attributed to the cross-cutting area of human performance because procedures were available, but not followed, that would have facilitated proper performance of the task.
The licensee entered this item into its corrective action program and planned to review other installed fuses and to conduct an evaluation of original problem.
The finding was greater than minor because the failure to adequately evaluate the impact of potentially degraded safety-related fuses on installed equipment, if left uncorrected, would become a significant safety concern. Specifically, the condition could cause premature circuit interruptions of safety-related or risk significant mitigating components, when called upon to perform the related functions, and this is an undesirable condition. The finding was of very low safety significance because the inspectors answered no to all the screening questions in the Significance Determination Process Phase 1 Screening Worksheet under the Mitigating Systems column.
Inspection Report# : 2006002(pdf)
Significance:        Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Adjustments performed on safety-related service water Valve 4B without procedure resulted in valve being declared inoperable On October 5, 2005, a finding of very low safety significance was self-revealed when SW-4B failed to meet its In-Service Testing stroke time requirements during the performance of Surveillance Procedure SP-02-138B and an associated unplanned entry into a Technical Specification Limiting Condition for Operation occurred. The condition occurred because the licensee made adjustments to safety-related Valve SW-4B, "Turbine Building Service Water Train "B" Header Isolation," without procedural guidance to perform such adjustments. The primary cause of this finding was related to the personal attribute of the cross-cutting area of human performance because maintenance was performed without required procedures.
The finding was more than minor because performing adjustment of safety-related equipment without procedural guidance, if left uncorrected, would become a more significant safety concern. Additionally, the finding is associated with the Reactor Safety/Mitigating Systems Cornerstone attribute of Procedure Quality and effects the associated Cornerstone objective of insuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using IMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," the inspectors answered "no" to all five screening questions in the Phase 1 Screening Worksheet under the Mitigating Systems column. Therefore, this finding was of very low safety significance. A Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was identified for the failure to provide procedural guidance for adjusting SW-4B; a safety-related valve which could affect the ability of safety-related mitigating system components to perform their intended function. Corrective actions taken by the licensee include procedural revisions to strengthen guidance on adjustment of safety-related components.
Inspection Report# : 2005017(pdf)
 
1Q/2006 Inspection Findings - Kewaunee                                                                                                Page 4 of 8 Significance:        Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Operator Licensing Exam Results Were Less Than Minimum Acceptable Percentage For Passing A finding of very low safety significance was identified. The finding was associated with unsatisfactory operating crew performance on the simulator during facility-administered licensed annual operator requalification examinations. Of the 7 crews evaluated, 2 did not pass their annual operating tests. The finding is of very low safety significance because the failures occurred during testing of the operators on the simulator, because there were no actual consequences to the failures, and because the crews were removed from watch-standing duties, retrained, and re-evaluated before they were authorized to return to control room watches.
Inspection Report# : 2005017(pdf)
Significance:        Dec 16, 2005 Identified By: NRC Item Type: FIN Finding No Trending of Adverse Conditions Identified During Outages The inspectors identified a finding of very low safety significance for the licensee not reviewing corrective action program documents (CAPs) during outages for potential trends of conditions adverse to quality. As part of the screening process of CAPs, the licensee assigned, as possible, CAPs to various "hot buttons." Hot buttons were searchable categories in the corrective action program computer system that had been established for various problems, such as equipment tagging errors, security door control, and reactivity management. For non-outage times, the licensee assigned a monthly number of hits for each hot button that, if exceeded for 3 months in succession, would result in the generation of a CAP to investigate a possible trend. However, as of December 16, 2005, the licensee did not use hot button action levels during outages when the number of CAPs written was much higher than during non-outage times.
This finding is greater than minor because if left uncorrected would become a more significant safety concern. This finding is not suitable for Significance Determination Process evaluation, but has been reviewed by NRC management and is determined to be a finding of very low safety significance. No violation of regulatory requirements occurred. The cause of the finding is related to the cross-cutting element of problem identification and resolution, because of not identifying potential conditions adverse to quality through trending of CAPs during outages.
Inspection Report# : 2005005(pdf)
Significance:        Dec 16, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct Procedure Non-Adherence The inspectors identified a finding of very low safety significance and a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," for the failure to take corrective action for procedure non-compliance identified during the licensee's 2004 self-assessment of the corrective action program. As a result of the assessment, CAP025194, "Corrective Action Program Procedure and Guidance Document Use," was written and documented that plant workers were not following corrective action program procedures for apparent cause evaluations and root cause evaluations, effectiveness review content, priority and due date assignments, initiator feedback, and documentation of corrective action completion. To correct this problem, corrective action CA018094, "Corrective Action Program Procedure and Guidance Document Use," was written and specified one or 2 weeks of requiring "in-hand" use by the plant staff of the corrective action program administrative procedure. However, completion of this action was delayed several times and on July 25, 2005, CAP025194 and CA018094 were closed with the only documented action taken being a July 18, 2005, meeting of the station human performance steering committee at which the licensee decided not to take action because of the pending transition to the corrective action program documents of the plant's new owner.
This finding is greater than minor because if left uncorrected would become a more significant safety concern. This finding is not suitable for Significance Determination Process evaluation, but has been reviewed by NRC management and is determined to be a finding of very low safety significance. The cause of the finding is related to the cross-cutting element of problem identification and resolution, because of the failure to take corrective action for non-adherence to station procedures.
Inspection Report# : 2005005(pdf)
Significance:        Dec 16, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Correct Residual Heat Removal Pump Seal Leakage A finding of very low safety significance that was a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action,"
was identified for the licensee's ineffective corrective action to repair a leak on the seal of the "B" residual heat removal (RHR) pump. The leak was identified on November 2, 2005, when the pump was stopped following the performance of a required surveillance. The leak rate exceeded leakage control program limits. A similar leak was identified on June 16, 2004, for which the licensee replaced the seal in November 2004.
This finding is greater than minor because it was associated with the "RCS (reactor coolant system) equipment and barrier performance" attribute of the barrier integrity cornerstone and does affect the cornerstone objective of providing reasonable assurance that physical design
 
1Q/2006 Inspection Findings - Kewaunee                                                                                                Page 5 of 8 barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events.
Although the RCS barrier was affected, the finding did not affect the mitigation capability of the RHR system and did not contribute to the likelihood of a primary or secondary system loss of coolant accident initiator or affect the containment integrity. Therefore, the finding is of very low safety significance.
Inspection Report# : 2005005(pdf)
Significance:        Oct 06, 2005 Identified By: NRC Item Type: VIO Violation Potential Flooding in the Turbine Building Basement A review of design drawings by the inspectors revealed a direct piping connection from the turbine building sump to the trench in safeguards alley. The inspectors determined that there were no check valves located in the piping to prevent water spills in the turbine building basement from backing up into the safeguards alley. The inspectors also noted that no flood barriers specifically designed to protect equipment in the safeguards alley from flooding in the turbine building basement were installed. The inspectors requested additional information from the licensee regarding potential flooding events occurring in the safeguards alley. The licensee documented its response to the inspectors information request in Condition Evaluation (CE) 014653. This CE stated that it would take approximately 3 hours for flooding caused by AFW pump discharge to affect safety-related equipment, and such flooding could be mitigated by opening doors between the safeguards alley and the turbine building basement. The CE also stated that other sources of flooding in the turbine building basement need not be considered since such flooding events are outside the design basis of the plant.
The inspectors identified a finding that was preliminarily determined to be of substantial to high safety significance because the licensee failed to provide adequate design control to ensure that Class I equipment was protected against damage from the rupture of a pipe or tank resulting in serious flooding or excessive steam release to the extent that the Class I equipment's function is impaired. Specifically, the design of Kewaunee Power Station (KPS) did not ensure that the auxiliary feedwater (AFW) pumps, the 480-volt (V) safeguards buses, the safe shutdown panel, emergency diesel generators (EDGs) 1A and 1B, and 4160-V safeguards buses 1-5 and 1-6 would be protected from random or seismically induced failures of non-Class I systems in the turbine building. The finding is also an apparent violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for not ensuring that the design of KPS prevented turbine building flooding from impacting multiple safety related equipment trains needed for safe shutdown of the plant. The inspectors determined that a primary cause of this finding was related to the cross-cutting area of Problem Identification and Resolution, because there was an earlier opportunity to discover and correct this issue based on the licensee's 2003 experience when minor flooding from the turbine building had challenged safety equipment located adjacent to the turbine building basement.
The finding was more than minor because it impacted Mitigating Systems cornerstone attributes of design control (initial design and plant modifications) and protection against external factors (internal flood hazards and seismic events) and it impacted the Mitigating Systems cornerstone objective to ensure availability, reliability and capability of multiple trains of safety related equipment to respond to events to prevent core damage. A Significance Determination Process Phase 3 risk analysis determined that this finding was preliminarily of substantial to high safety significance. The licensee has taken significant corrective actions, including extensive system and structural modifications to address this issue.
After considering the information developed during the inspection, and the additional information you provided prior to, during, and in response to our questions at the Regulatory Conference, the NRC has concluded the inspection finding is appropriately characterized as Yellow (i.e., an issue with substantial importance to safety, that will result in additional NRC inspection and potentially other NRC action).
Inspection Report# : 2004009(pdf)
Inspection Report# : 2005002(pdf)
Inspection Report# : 2005011(pdf)
Inspection Report# : 2005018(pdf)
Significance: SL-IV Sep 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Report in a Timely Manner an Unanalyzed Condition Involving a Potential Runout Concern With the CCW Pumps The inspectors identified a Non-Cited Violation (NCV) when the licensee failed to make a written report, within 60 days, to the NRC in accordance with 10 CFR 50.73(a)(2)(ii)(B), when an unanalyzed condition that significantly degraded plant safety was identified. Specifically, the licensee did not recognize the significance of a previously identified condition involving a potential runout issue with the component cooling water (CCW) pumps, and did not report this condition until the inspectors identified the requirement. The concern related to the CCW pump capability to provide required flow under certain conditions. Specifically, during a loss of power, and with specific system configurations, the loss of power could lead to a CCW pump runout condition. The primary cause of this finding was related to the cross-cutting area of human performance.
Because this issue affects the NRC's ability to perform its regulatory function, it was evaluated using the traditional enforcement process. The inspectors determined that this violation is of very low safety significance and because the licensee entered the issue into their corrective action program (CAP026528), this violation is being treated as an NCV consistent with Section VI.A.1 of the NRC Enforcement Policy. The licensee has taken actions to revise plant procedures to address this issue.
Inspection Report# : 2005012(pdf)
 
1Q/2006 Inspection Findings - Kewaunee                                                                                              Page 6 of 8 Significance:        Aug 16, 2005 Identified By: NRC Item Type: VIO Violation Potential Common Mode Failure of Auxiliary Feedwater URI 05000305/2005002-05 is associated with the design of the AFW pump's discharge pressure switches. The inspectors identified the potential for air intrusion into operating AFW pumps, potentially resulting in a common mode failure of the AFW system. This could occur during certain events where the suction source is lost prior to being able to manually swap the source of water from the CST to the SW system.
The inspectors identified a finding that was preliminarily determined to be of low to moderate safety significance, because Kewaunee failed to provide adequate design control to ensure the AFW pumps would be protected from failure due to air ingestion during tornado or seismic events; as well as from failure during potential runout conditions. The finding is also an apparent violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for not effectively providing controls to check the adequacy of the design for protecting the AFW pumps during design and license basis events.
The finding was determined to be more than minor since it impacted Mitigating System cornerstone attributes of design control (initial design and plant modifications) and the cornerstone objective to ensure availability, reliability, and capability of the AFW system to respond to events to prevent core damage. A Significance Determination Process Phase 3 risk analysis determined that this finding was preliminarily of low to moderate safety significance. The licensee has taken significant corrective actions, including extensive modifications to the system.
After considering the information developed during the inspection, the NRC has concluded the inspection finding is appropriately characterized as White (i.e., an issue with low to moderate increased importance to safety, which may require additional NRC inspections).
Inspection Report# : 2005002(pdf)
Inspection Report# : 2005010(pdf)
Inspection Report# : 2005014(pdf)
Significance:        Jul 29, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Evaluate the Effect of Modification on Turbine Driven AFW Pump Performance with Reduced Steam Pressure The inspectors identified a finding involving a Green Non-Cited Violation (NCV) of 10 CFR 50, Appendix B, Criterion III, "Design Control".
The finding involved the revision of AFW pump discharge pressure trip setpoints. The licensee had not determined if the turbine driven AFW (TDAFW) pump was capable of providing the required flow under reduced steam pressure conditions prior to approving the modification. This issue could have affected the performance of the AFW system under post accident conditions.
This issue was greater than minor because it potentially affected the Mitigating System cornerstone objective of equipment capability. The issue screened as very low safety significance in Phase 1 of the SDP, because it was a design deficiency that was not found to result in a loss of function and the item was resolved prior to being in the plant conditions where the finding could have impacted the pump's performance. The licensee conducted post modification tests and revised permanent plant procedures to ensure the TDAFW pump was capable of providing the required flow under reduced steam pressure conditions.
Inspection Report# : 2005010(pdf)
Significance:        Jun 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to minimize prior identified and predictable explosive gas concentrations in the WGDTs A finding of very low safety significance was identified by the inspectors for a NCV of Title 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action." The licensee failed to consider the impact on plant fire protection when ineffective resolution of waste gas system issues repeatedly led to explosive mixtures in the Waste Gas Decay Tanks. The licensee entered these issues into their corrective action program. The primary cause of this violation was related to the cross-cutting area of Problem Identification and Resolution. The licensee repeatedly encountered explosive gas levels in the WGDTs and were aware of plant conditions that resulted in these levels but failed to take adequate corrective actions to prevent explosive gas mixtures from developing in the WGDTs. The issue is more than minor because uncontrolled explosive mixtures in the WGDTs could have led to a more significant safety concern. The issue was of very low safety significance because explosive mixtures were only present during plant shutdown conditions; an explosion would not have affected safe shutdown equipment (i.e.
Residual Heat Removal System); the explosive mixture conditions were only present for short periods of time (<12 hours); and the tanks were isolated and vented per procedure when discovered.
Inspection Report# : 2005008(pdf)
Barrier Integrity
 
1Q/2006 Inspection Findings - Kewaunee                                                                                              Page 7 of 8 Emergency Preparedness Occupational Radiation Safety Significance:        Sep 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Post and Control Access Into a Locked High Radiation Area During Radiographic Activities A self-revealed finding of very low safety significance and an associated Non-Cited Violation of NRC requirements were identified when a high radiation area boundary was breached by two workers during radiography. An unnecessary radiation exposure could have been received by the workers had they not been stopped by radiography personnel as they moved toward the exposed radiographic source.
The issue 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 adequate protection of worker health and safety from exposure to radiation. The issue represents a finding of very low safety significance because there was no overexposure or substantial potential for an overexposure given the actual radiological conditions in the area coupled with the duration of the radiographic operation and the presence of radiography personnel who provided surveillance of the area, nor was the licensee's ability to assess worker dose compromised. A Non-Cited Violation of Technical Specification 6.13(a) and 10 CFR 20.1601(b) was identified for the failure to comply with the RP requirements that govern the control of access into high radiation areas. Corrective actions taken by the licensee included enhanced administrative measures by revising the radiography procedure and counseling of involved staff. Since the cause of the problem included corrective action deficiencies from previous similar radiography boundary control events, the finding also relates to the cross-cutting area of problem identification and resolution.
Inspection Report# : 2005012(pdf)
Significance:        Sep 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Control Access Into a High Radiation Area During Radiographic Activities A self-revealed finding of very low safety significance and an associated Non-Cited Violation of NRC requirements were identified for an unposted/uncontrolled locked high radiation area in the turbine building during radiography activities. A radiography source created radiation levels such that a major portion of the whole body could have received in one hour a dose in excess of 1000 mrem in accessible areas of the turbine building, which were not posted or controlled in accordance with regulatory requirements. The areas with elevated dose rates were not positively controlled by locked door/gate, use of a barrier and flashing light, or maintained under continuous visual or electronic surveillance.
The issue 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 adequate protection of worker health and safety from exposure to radiation. The issue represents a finding of very low safety significance because there was no overexposure or substantial potential for an overexposure given the actual radiological conditions in the uncontrolled areas coupled with the duration of the radiographic shot. A Non-Cited Violation of Technical Specification 6.13(b) and 10 CFR 20.1601(b) was identified for the failure to comply with the RP requirements that govern the control of access into locked high radiation areas. Corrective actions taken by the licensee included enhanced administrative measures by revising the radiography procedure, and counseling and training of RP staff.
Inspection Report# : 2005012(pdf)
Significance:        Sep 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Control Access Into a High Radiation Area While Moving a Radioactive Filter A self-revealed finding of very low safety significance and two associated Non-Cited Violations of regulatory requirements were identified for an unposted and uncontrolled high radiation area in an auxiliary building elevator during the transfer of a radioactive seal water injection filter.
As a result of this failure, workers could have unknowingly entered a high radiation area in the elevator without knowledge of the radiological conditions.
The issue 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 adequate protection of worker health and safety from exposure to radiation. The issue represents a finding of very low safety significance because there was no overexposure or substantial potential for an overexposure given the circumstances and the actual radiological conditions in the area, nor was the licensee's ability to assess worker dose compromised. A Non-Cited Violation of Technical Specification 6.13(a) and 10 CFR 20.1601(b) was identified for the failure to comply with the RP requirements that govern the control of access into high radiation areas. This issue also represents a Non-Cited Violation of 10 CFR 20.1902(b)/20.1903 for failure to post a high radiation area. Corrective actions taken by the licensee included enhanced administrative measures (RP Job Guide) for change-out and transport of all radioactive filters.
 
1Q/2006 Inspection Findings - Kewaunee                                                                                                Page 8 of 8 Inspection Report# : 2005012(pdf)
Significance:        Sep 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Post a High Radiation Area While Moving a Radioactive Filter A self-revealed finding of very low safety significance and two associated Non-Cited Violations of regulatory requirements were identified for an unposted and uncontrolled high radiation area in an auxiliary building elevator during the transfer of a radioactive seal water injection filter.
As a result of this failure, workers could have unknowingly entered a high radiation area in the elevator without knowledge of the radiological conditions.
The issue 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 adequate protection of worker health and safety from exposure to radiation. The issue represents a finding of very low safety significance because there was no overexposure or substantial potential for an overexposure given the circumstances and the actual radiological conditions in the area, nor was the licensee's ability to assess worker dose compromised. A Non-Cited Violation of Technical Specification 6.13(a) and 10 CFR 20.1601(b) was identified for the failure to comply with the RP requirements that govern the control of access into high radiation areas. This issue also represents a Non-Cited Violation of 10 CFR 20.1902(b)/20.1903 for failure to post a high radiation area. Corrective actions taken by the licensee included enhanced administrative measures (RP Job Guide) for change-out and transport of all radioactive filters.
Inspection Report# : 2005012(pdf)
Public Radiation Safety Significance:        Mar 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Evaluate Degraded Flow Conditions on a SW System Radiation Monitor A finding of very low safety significance and an associated non-cited violation of the Kewaunee Technical Specifications, Section 6.8, Procedures, was identified by the inspectors for the failure to adequately evaluate degraded flow in a service water system radiation monitor.
The primary cause of this finding was attributed to the cross-cutting area of human performance because procedures were available, but not followed, that would have facilitated proper performance of the task.
The licensee entered this item into its corrective action program and planned to conduct inspections of other radiation monitor sample chambers, assess the need for an in-line filter, and assess the need for a modification to correct the recurring problem with the service water radiation monitor.
The finding was greater than minor because the finding involved conditions contrary to those required by the offsite dose calculation manual.
Specifically, sampling requirements that were required to be initiated when the related radiation monitoring instrumentation should have been declared inoperable were not accomplished. The finding was of very low safety significance because no radiological releases were possible from the indicated pathways when the condition existed.
Inspection Report# : 2006002(pdf)
Physical Protection Physical Protection information not publicly available.
Miscellaneous Last modified : May 25, 2006
 
2Q/2006 Inspection Findings - Kewaunee                                                                                                    Page 1 of 11 Kewaunee 2Q/2006 Plant Inspection Findings Initiating Events Significance:        Jun 23, 2006 Identified By: NRC Item Type: NCV NonCited Violation Procedure for Reactor Startup Not Followed The inspectors identified a finding associated with a non-cited violation of Technical Specification 6.8.a (written procedures and administrative policies). The finding was for the licensees failure to follow approved procedures during a plant startup. The finding was of very low safety significance and there were three examples of the finding. The first example of a failure to follow approved procedures occurred when operators incorrectly marked a procedure step as not applicable and failed to execute the step. The second example of the failure to follow approved procedures occurred when operators executed procedure steps out of sequence. The third example occurred during the previous reactor startup conducted in November 2005 when operators performed procedure steps out of sequence in the same manner as executed during this plant startup.
Corrective actions included placing Procedure N-0-01 on administrative hold until appropriate procedure changes could be made and training operating crews on procedure adherence.
This finding was of more than minor safety significance. Failure to comply with reactivity management requirements can lead to an uncontrolled reactivity event. In this particular event, the failure to follow the procedural sequence could have resulted in shutdown margin being less than that required by Technical Specifications. However, this finding is of very low significance because the actual shutdown margin did not go below the minimum required by Technical Specifications. This finding affected the cross-cutting issue of human performance.
Inspection Report# : 2006011(pdf)
Significance:        Jun 23, 2006 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedure for Reactor Startup The inspectors identified a finding associated with an non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, of very low safety significance associated with an event. The inspectors identified that Procedure N-0-01, Plant Startup from Cold Shutdown Condition to Hot Shutdown Condition, Revision BI, Step 4.45 was inadequate to start up the reactor for the conditions that existed on May 17, 2006. The procedure, as written, would have required the operators to dilute the reactor to a lower boron concentration than the Estimated Critical Position boron concentration prior to withdrawing the Shutdown Bank rods. Corrective actions to address this finding included placing Procedure N-0-01 on administrative hold until appropriate procedure changes could be implemented.
This finding was more than minor in safety significance because this issue, if left uncorrected, would have resulted in the core reactivity shutdown margin being less than that required by Technical Specifications. However, this finding is of very low significance because the procedure step was not executed and shutdown was never below that required by Technical Specifications. This finding affected the cross-cutting issue of human performance.
Inspection Report# : 2006011(pdf)
Significance:        May 19, 2006 Identified By: NRC Item Type: NCV NonCited Violation Criterion XVI: Failed to Identify Causes and Corrective Actions to Preclude Repetition for Significant Conditions Adverse to Quality The NRC inspectors identified a finding of very low safety significance that involved a violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions. Specifically, for the turbine building flooding and auxiliary feedwater air entrainment performance deficiencies, which were significant conditions adverse to quality, the licensee failed to identify the causes, and to determine corrective actions to preclude repetition.
The finding was greater than minor because the failure to identify the cause and corrective actions to preclude repetition of significant conditions adverse to quality, which led to a degraded cornerstone could result in the NRC needing to take more significant action. The finding was determined to be of very low safety significance based on management review, and the determination that no additional instances of significant conditions adverse to quality have actually occurred due to the failure to identify the causes and corrective actions for the previous performance deficiencies.
The cause of the finding was related to the evaluation aspect of the cross-cutting element of problem identification and resolution.
Inspection Report# : 2006007(pdf)
Significance:        May 05, 2006 Identified By: NRC
 
2Q/2006 Inspection Findings - Kewaunee                                                                                                    Page 2 of 11 Item Type: NCV NonCited Violation Failure to Incorporate Operating Experience Into Preventive Maintenance Procedures The inspectors identified a finding associated with a non-cited violation (NCV) of 10 CFR 50.65 (the Maintenance Rule), having very low safety significance for the licensees failure to incorporate into station procedures available internal and external operating experience pertaining to 4.16-kilovolt (kV) switchgear mechanically operated contact (MOC) switch linkage assemblies. As a result, preventive maintenance procedures for 4.16-kV safety- and nonsafety-related switchgear breaker cubicles were inadequate and had not been upgraded to incorporate important MOC switch linkage measurements and adjustments to be used during periodic breaker/cubicle maintenance. The licensee entered the problem with the procedures into its corrective action program for resolution. Corrective action included the revision of the procedures to incorporate the need to inspect the linkage and adjust it to within specified values.
The finding is greater than minor because it is associated with the procedure adequacy 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 operation. The finding was determined to be of very low safety significance because the transient initiator contributor is a reactor trip that did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available. The cause of the finding is related to the cross-cutting element of problem identification and resolution.
Inspection Report# : 2006010(pdf)
Significance:        Mar 31, 2006 Identified By: NRC Item Type: FIN Finding Failure to Control Loose Materials Within the Protected Area in Response to Adverse Weather Conditions A finding of very low safety significance was identified by the inspectors for the licensees failure to control loose materials within the protected area south of the transformer bays in response to adverse weather conditions. The material could have been blown into the transformers and initiate a transient. The primary cause of this finding was related to the cross-cutting area of problem identification and resolution for the failure to implement effective corrective actions in response to a similar, previous inspection finding (Inspection Report 05000305/2005008). No violation of regulatory requirements occurred.
The licensee entered this issue into its corrective action program and removed the loose material from the transformer bays.
The finding is more than minor because, if left uncorrected, the loose items would become a more significant safety concern by becoming missile hazards; thereby, increasing the likelihood of an initiating event. Additionally, the inspectors determined that this issue was associated with the procedure quality 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 because the station procedure used to control potential airborne material was too narrow in scope. The finding was of very low safety significance because the inspectors answered no to all the screening questions in the Significance Determination Process Phase 1 Screening Worksheet under the Initiating Events column.
Inspection Report# : 2006002(pdf)
Significance:        Mar 30, 2006 Identified By: NRC Item Type: FIN Finding Failure to Adequately Evaluate an Inoperative Indicating Lamp for a Turbine control Valve A finding of very low safety significance was identified by the inspectors for the failure to adequately evaluate an inoperative indicating lamp associated with the turbine control valves. The primary cause of this finding was attributed to the cross-cutting area of human performance because procedures were available, but not followed, that would have facilitated proper performance of the task.
The licensee entered this item into its corrective action program and reviewed open work orders, provided a status update to management, and increased communications of related expectations.
The finding is greater than minor because the failure to adequately evaluate deficient conditions, if left uncorrected, would become a more significant safety concern. The finding was of very low safety significance because the inspectors answered no to all the questions in the Significance Determination Process Phase 1 Screening Worksheet under the Initiating Events column.
Inspection Report# : 2006002(pdf)
Significance:        Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Startup Procedure Resulted in an Inadvertent Carbon-Dioxide Fire Suppression Discharge and Declaration of a Notice of Unusual Event A finding of very low safety significance was self-revealed during two events when use of an inadequate plant prestartup procedure resulted in actuation of the CARDOX Carbon Dioxide system. A Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was identified for the failure to include adequate acceptance criteria in Procedure N-0-02-CLA, "Plant Prestartup Checklist". The primary cause of this finding was related to the resource attribute in the cross-cutting area of Human Performance. The licensee failed to provide the operators with quality procedures containing criteria to know when the secondary plant was appropriately aligned.
 
2Q/2006 Inspection Findings - Kewaunee                                                                                                      Page 3 of 11 The inspectors determined that the finding was greater than minor because it involved the configuration control, human performance, and procedure quality attributes of the Initiating Events Cornerstone. Additionally the finding affected the cornerstone objective of limiting the likelihood of those events that upset plant stability during power operations. Specifically, an incorrect lineup could exist in the secondary system resulting in an initiating event, or an unanalyzed secondary system response after a trip. The issue was of very low safety significance because the finding did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available. Corrective actions taken by the licensee include procedural enhancements to ensure that systems are lined up properly before continuing with plant startup.
Inspection Report# : 2005017(pdf)
Mitigating Systems Significance:        Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Reactor Protection System Surveillance Procedure Revised Without Proper Review The inspectors identified a finding of very low safety significance and an associated non-cited violation of Technical Specification 6.8, Procedures, during a review of a procedure. The licensee had changed the procedure to allow the turbine-driven auxiliary feedwater (TDAFW) pump to be considered available for risk management purposes while the pump control switch was in pull-to-lock during the performance of the surveillance procedure; however, the required Plant Operating Review Committee review and approval for the change was not obtained. Corrective actions, to date, included review of the surveillance procedure by the Plant Operating Review Committee and inclusion into the procedure of additional provisions to ensure availability of the TDAFW pump while the control switch is in pull-to-lock during performance of the procedure.
The cause of this finding is related to the cross-cutting area of human performance because of the licensees failure to follow a plant procedure regarding the review and approval of safety-related procedures.
The finding is greater than minor because if left uncorrected the finding would become a more significant safety concern. Specifically, improper application of the temporary procedure change process could lead to a more significant unreviewed, improper procedure change. Additionally, this issue is associated with the procedure quality 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).
Specifically, the failure to provide adequate review and approval of a safety-related surviellance procedure prior to issuance for use and the failure to include adequate provisions to ensure availability of a safety-related component in the surveillance procedure potentially impacted equipment availability. The finding is of very low safety significance because the answer to all the screening questions in the significance determination process Phase 1 screening worksheet in the Mitigating Systems column was no.
Inspection Report# : 2006003(pdf)
Significance:        Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Leak Developed in Service Water Pipe after Wall Thinning Evaluation was Cancelled A self-revealed finding of very low safety significance and an associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, were identified on April 25, 2006, when a leak due to pipe-wall thinning was identified in a 90&deg; elbow in a service water (SW) line to the B emergency diesel generator. This wall-thinning and leak, a condition adverse to quality, resulted in the need to declare the emergency diesel generator inoperable and a shut down of the reactor to allow repair of the leak. In April 2004, a work order to inspect the elbow for wall-thinning was cancelled after wall thickness in a nearby elbow was evaluated by the licensee and deemed acceptable. The extrapolation of inspection results from one elbow to the other elbow was inappropriate. Corrective actions taken by the licensee included replacement of the failed section of SW piping, performance of additional inspections on SW piping, and replacement of other safety-related sections of SW piping. The cause of this finding is related to the cross-cutting area of problem identification and resolution because the licensee failed to promptly identify an issue potentially impacting safety-related piping.
The finding is greater than minor because it is associated with the equipment performance attribute of the Mitigating System 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). Specifically, the failure to conduct a wall-thinning evaluation in April 2004 resulted in the need to take the emergency diesel generator out-of-service and shut down the reactor to allow repair of the pipe. Additionally, the failure to inspect and correct, as necessary, wall-thinning in a safety-related system, if left uncorrected, would become a more significant safety concern through the possible development of a large system leak or the complication of the operations of a safety-related system. The finding is of very low safety significance because the answer to all the screening questions in the significance determination process Phase 1 screening worksheet in the Mitigating Systems column was no.
Inspection Report# : 2006003(pdf)
Significance:        May 19, 2006 Identified By: NRC Item Type: NCV NonCited Violation Criterion V: Failed to Incorporate Appropriate Acceptance Criteria for Assessing Operability of the AFW Pump
 
2Q/2006 Inspection Findings - Kewaunee                                                                                                      Page 4 of 11 The NRC inspectors identified a finding of very low safety significance that involved a violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings. Specifically, the licensee failed to incorporate appropriate acceptance criteria for assessing operability of the auxiliary feedwater pump following identification of a piping obstruction.
The finding was greater than minor because the finding was associated with the Mitigating Systems cornerstone attribute of procedure quality which affected the cornerstone objective. Specifically, the relevant procedure was not adequate to ensure the availability, reliability, and capability of the auxiliary feedwater system to respond to initiating events. The finding was determined to be of very low safety significance because subsequent evaluation of the pipe occlusions, using appropriate acceptance criteria, supported past operability of the pump. The cause of the finding was related to the evaluation aspect of the cross-cutting element of problem identification and resolution.
Inspection Report# : 2006007(pdf)
Significance:        May 19, 2006 Identified By: NRC Item Type: NCV NonCited Violation Criterion III: Failed to Correctly Translate Containment Sump Volume into Design The NRC inspectors identified a finding of very low safety significance that involved a violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to ensure that design basis calculations correctly translated the containment sump volume at the time of the switch over from the refueling water storage tank to the containment sump to ensure adequate available net positive suction head and vortex suppression for the residual heat removal pumps.
The finding was greater than minor because the finding was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective because the inadequate calculation impacted the design requirements for the new containment strainers being installed to resolve Generic Safety Issue 191. The finding was determined to be of very low safety significance because (1) the licensee normally kept the refueling water storage tank at a level above the Technical Specification minimum; (2) new strainers were not yet installed; and (3) inspector-independent calculations indicated that the pumps had adequate net positive suction head and vortex suppression, with the additional non-conservatisms incorporated. The cause of the finding was related to the corrective action aspect of the cross-cutting element of problem identification and resolution.
Inspection Report# : 2006007(pdf)
Significance:        May 19, 2006 Identified By: NRC Item Type: NCV NonCited Violation Criterion III: Failed to Verify or Check the Adequacy of the Design Canceling Design Change Request 2548 The NRC inspectors identified a finding of very low safety significance that involved a violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to properly evaluate the minimum flow requirements of the high head safety injection pumps.
The finding was greater than minor because the finding was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective as providing inadequate minimum flow to the SI pumps could result in the pumps failing under certain accident scenarios. The finding was determined to be of very low safety significance because both the licensee and the inspectors determined that the safety injection pumps remained operable with the 47 gpm minimum flow rate. The cause of the finding was related to the corrective action of the cross-cutting element of problem identification and resolution.
Inspection Report# : 2006007(pdf)
Significance:        May 05, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain cable separation for cables 1N15010 and IN15012 associated with train 'B' of ICCMS The inspectors identified a finding associated with a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, that pertained to a modification that failed to incorporate applicable design requirements for cable separation. Nonsafety-related cables associated with train B reactor coolant pump (RCP) safety-related cable trays and cables were bundled inside the RCP breaker cubicles with train A RCP safety-related cables feeding the reactor protection system (RPS). Consequently, a fault in the train B cable/cable tray could propagate to train A. The licensee entered the problem into its corrective action program for resolution. Corrective actions included encasing the nonsafety-related cables in flexible metal conduit and confirming that other safety-related cables were not affected.
The finding is greater than minor 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). The finding was determined to be of very low safety significance because of the redundancy and coincident logic in the RPS design; and it did not represent a loss of system safety function, an actual loss of safety function of a single train, an actual loss of safety function of one or more non-technical specification trains of equipment, designated as risk significant per 10 CFR 50.65, for greater than 24 hours, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event.
Inspection Report# : 2006010(pdf)
 
2Q/2006 Inspection Findings - Kewaunee                                                                                                      Page 5 of 11 Significance:        Mar 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation Ineffective Corrective Actions to Resolve Boric Acid Leakage from the 1A RHR Pump Flange Studs and Nuts A finding of very low safety significance and an associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified by the inspectors for ineffective identification and the initiation of corrective actions to resolve boric acid leakage from the 1A residual heat removal (RHR) pump flange studs and nuts. The primary cause of this finding was attributed to the cross-cutting area of problem identification and resolution. During a review of corrective actions associated with the licensees identification of a moderate amount of boric acid around various pump flange studs and nuts, the inspectors found that numerous prior occasions existed where the licensee had identified similar conditions yet failed to adequately identify and initiate actions to evaluate or correct this condition adverse to quality.
The licensee entered this item into its corrective action program and wrote a work order to replace the pump casing flange gasket.
The finding is greater than minor because it is associated with the equipment performance attribute of the Mitigating System 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). Additionally, failure to correct a condition adverse to quality in a safety-related system, if left uncorrected, would become a more significant safety concern. The finding was of very low safety significance because the inspectors answered no to all the screening questions in the Significance Determination Process Phase 1 Screening Worksheet under the Mitigating Systems column.
Inspection Report# : 2006002(pdf)
Significance:        Mar 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Apply Appropriate Quality Classification to TSC Diesel Generator Modifications as Required by Procedures A finding of very low safety significance and an associated non-cited violation of the Kewaunee Technical Specifications, Section 6.8, Procedures, was identified by the inspectors during a review of plant modification Design Change Request 3490, which replaced the existing Technical Support Center diesel generator fuel oil day tank level switches with new level switches of a different design. The inspectors determined that, in accordance with procedure GNP-01.01.01, Determination of Nuclear Safety Designed Classifications, QA [Quality Assurance] Type and EQ [Environmental Qualification] Type, the new level switches should have been designated as Augmented Quality. Contrary to this, the new switches were not designated as augmented quality. The primary cause of this finding was attributed to the cross-cutting area of problem identification and resolution because of the licensees failure to take effective corrective actions for previously identified problems with its quality assurance program.
The licensee entered this item into its corrective action program and conducted supplemental audits of quality-designated equipment, added additional related elements to an upcoming quality assurance group audit of the quality assurance program, and the conduct of a cause evaluation of related issues.
The finding is greater than minor because it is 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). Additionally, failure to comply with the provisions of nuclear safety-related procedures, if left uncorrected, would become a more significant safety concern. The finding is of very low safety significance because the inspectors answered no to all the screening questions in the Significance Determination Process Phase 1 Screening Worksheet under the Mitigating Systems column.
Inspection Report# : 2006002(pdf)
Significance:        Mar 30, 2006 Identified By: NRC Item Type: FIN Finding Failure to Adequately Evaluate the Extent-of-Condition of Degraded Fuses in Installed Equipment A finding of very low safety significance was identified by the inspectors for the failure to adequately evaluate the extent-of-condition relative to installed equipment for a 10 CFR Part 21 notification for degraded Bussmann fuses. The primary cause of the finding was attributed to the cross-cutting area of human performance because procedures were available, but not followed, that would have facilitated proper performance of the task.
The licensee entered this item into its corrective action program and planned to review other installed fuses and to conduct an evaluation of original problem.
The finding was greater than minor because the failure to adequately evaluate the impact of potentially degraded safety-related fuses on installed equipment, if left uncorrected, would become a significant safety concern. Specifically, the condition could cause premature circuit interruptions of safety-related or risk significant mitigating components, when called upon to perform the related functions, and this is an undesirable condition. The finding was of very low safety significance because the inspectors answered no to all the screening questions in the Significance Determination Process Phase 1 Screening Worksheet under the Mitigating Systems column.
Inspection Report# : 2006002(pdf)
 
2Q/2006 Inspection Findings - Kewaunee                                                                                                    Page 6 of 11 Significance:        Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Adjustments Performed on Safety-Related Service Water Valve 4B Without Procedure Resulted in Valve Being Declared Inoperable On October 5, 2005, a finding of very low safety significance was self-revealed when SW-4B failed to meet its In-Service Testing stroke time requirements during the performance of Surveillance Procedure SP-02-138B and an associated unplanned entry into a Technical Specification Limiting Condition for Operation occurred. The condition occurred because the licensee made adjustments to safety-related Valve SW-4B, "Turbine Building Service Water Train "B" Header Isolation," without procedural guidance to perform such adjustments. The primary cause of this finding was related to the personal attribute of the cross-cutting area of human performance because maintenance was performed without required procedures.
The finding was more than minor because performing adjustment of safety-related equipment without procedural guidance, if left uncorrected, would become a more significant safety concern. Additionally, the finding is associated with the Reactor Safety/Mitigating Systems Cornerstone attribute of Procedure Quality and effects the associated Cornerstone objective of insuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using IMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," the inspectors answered "no" to all five screening questions in the Phase 1 Screening Worksheet under the Mitigating Systems column. Therefore, this finding was of very low safety significance. A Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was identified for the failure to provide procedural guidance for adjusting SW-4B; a safety-related valve which could affect the ability of safety-related mitigating system components to perform their intended function. Corrective actions taken by the licensee include procedural revisions to strengthen guidance on adjustment of safety-related components.
Inspection Report# : 2005017(pdf)
Significance:        Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Operator Licensing Exam Results Were Less Than Minimum Acceptable Percentage For Passing A finding of very low safety significance was identified. The finding was associated with unsatisfactory operating crew performance on the simulator during facility-administered licensed annual operator requalification examinations. Of the 7 crews evaluated, 2 did not pass their annual operating tests. The finding is of very low safety significance because the failures occurred during testing of the operators on the simulator, because there were no actual consequences to the failures, and because the crews were removed from watch-standing duties, retrained, and re-evaluated before they were authorized to return to control room watches.
Inspection Report# : 2005017(pdf)
Significance:        Dec 16, 2005 Identified By: NRC Item Type: FIN Finding No Trending of Adverse Conditions Identified During Outages The inspectors identified a finding of very low safety significance for the licensee not reviewing corrective action program documents (CAPs) during outages for potential trends of conditions adverse to quality. As part of the screening process of CAPs, the licensee assigned, as possible, CAPs to various "hot buttons." Hot buttons were searchable categories in the corrective action program computer system that had been established for various problems, such as equipment tagging errors, security door control, and reactivity management. For non-outage times, the licensee assigned a monthly number of hits for each hot button that, if exceeded for 3 months in succession, would result in the generation of a CAP to investigate a possible trend. However, as of December 16, 2005, the licensee did not use hot button action levels during outages when the number of CAPs written was much higher than during non-outage times.
This finding is greater than minor because if left uncorrected would become a more significant safety concern. This finding is not suitable for Significance Determination Process evaluation, but has been reviewed by NRC management and is determined to be a finding of very low safety significance. No violation of regulatory requirements occurred. The cause of the finding is related to the cross-cutting element of problem identification and resolution, because of not identifying potential conditions adverse to quality through trending of CAPs during outages.
Inspection Report# : 2005005(pdf)
Significance:        Dec 16, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct Procedure Non-Adherence The inspectors identified a finding of very low safety significance and a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," for the failure to take corrective action for procedure non-compliance identified during the licensee's 2004 self-assessment of the corrective action program. As a result of the assessment, CAP025194, "Corrective Action Program Procedure and Guidance Document Use,"
was written and documented that plant workers were not following corrective action program procedures for apparent cause evaluations and root cause evaluations, effectiveness review content, priority and due date assignments, initiator feedback, and documentation of corrective action completion. To correct this problem, corrective action CA018094, "Corrective Action Program Procedure and Guidance Document Use," was written and specified one or 2 weeks of requiring "in-hand" use by the plant staff of the corrective action program administrative procedure.
However, completion of this action was delayed several times and on July 25, 2005, CAP025194 and CA018094 were closed with the only
 
2Q/2006 Inspection Findings - Kewaunee                                                                                                    Page 7 of 11 documented action taken being a July 18, 2005, meeting of the station human performance steering committee at which the licensee decided not to take action because of the pending transition to the corrective action program documents of the plant's new owner.
This finding is greater than minor because if left uncorrected would become a more significant safety concern. This finding is not suitable for Significance Determination Process evaluation, but has been reviewed by NRC management and is determined to be a finding of very low safety significance. The cause of the finding is related to the cross-cutting element of problem identification and resolution, because of the failure to take corrective action for non-adherence to station procedures.
Inspection Report# : 2005005(pdf)
Significance:        Dec 16, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Correct Residual Heat Removal Pump Seal Leakage A finding of very low safety significance that was a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was identified for the licensee's ineffective corrective action to repair a leak on the seal of the "B" residual heat removal (RHR) pump. The leak was identified on November 2, 2005, when the pump was stopped following the performance of a required surveillance. The leak rate exceeded leakage control program limits. A similar leak was identified on June 16, 2004, for which the licensee replaced the seal in November 2004.
This finding is greater than minor because it was associated with the "RCS (reactor coolant system) equipment and barrier performance" attribute of the barrier integrity cornerstone and does affect the cornerstone objective of providing reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Although the RCS barrier was affected, the finding did not affect the mitigation capability of the RHR system and did not contribute to the likelihood of a primary or secondary system loss of coolant accident initiator or affect the containment integrity. Therefore, the finding is of very low safety significance.
Inspection Report# : 2005005(pdf)
Significance:        Oct 06, 2005 Identified By: NRC Item Type: VIO Violation Potential Flooding in the Turbine Building Basement A review of design drawings by the inspectors revealed a direct piping connection from the turbine building sump to the trench in safeguards alley.
The inspectors determined that there were no check valves located in the piping to prevent water spills in the turbine building basement from backing up into the safeguards alley. The inspectors also noted that no flood barriers specifically designed to protect equipment in the safeguards alley from flooding in the turbine building basement were installed. The inspectors requested additional information from the licensee regarding potential flooding events occurring in the safeguards alley. The licensee documented its response to the inspectors information request in Condition Evaluation (CE) 014653. This CE stated that it would take approximately 3 hours for flooding caused by AFW pump discharge to affect safety-related equipment, and such flooding could be mitigated by opening doors between the safeguards alley and the turbine building basement. The CE also stated that other sources of flooding in the turbine building basement need not be considered since such flooding events are outside the design basis of the plant.
The inspectors identified a finding that was preliminarily determined to be of substantial to high safety significance because the licensee failed to provide adequate design control to ensure that Class I equipment was protected against damage from the rupture of a pipe or tank resulting in serious flooding or excessive steam release to the extent that the Class I equipment's function is impaired. Specifically, the design of Kewaunee Power Station (KPS) did not ensure that the auxiliary feedwater (AFW) pumps, the 480-volt (V) safeguards buses, the safe shutdown panel, emergency diesel generators (EDGs) 1A and 1B, and 4160-V safeguards buses 1-5 and 1-6 would be protected from random or seismically induced failures of non-Class I systems in the turbine building. The finding is also an apparent violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for not ensuring that the design of KPS prevented turbine building flooding from impacting multiple safety related equipment trains needed for safe shutdown of the plant. The inspectors determined that a primary cause of this finding was related to the cross-cutting area of Problem Identification and Resolution, because there was an earlier opportunity to discover and correct this issue based on the licensee's 2003 experience when minor flooding from the turbine building had challenged safety equipment located adjacent to the turbine building basement.
The finding was more than minor because it impacted Mitigating Systems cornerstone attributes of design control (initial design and plant modifications) and protection against external factors (internal flood hazards and seismic events) and it impacted the Mitigating Systems cornerstone objective to ensure availability, reliability and capability of multiple trains of safety related equipment to respond to events to prevent core damage. A Significance Determination Process Phase 3 risk analysis determined that this finding was preliminarily of substantial to high safety significance. The licensee has taken significant corrective actions, including extensive system and structural modifications to address this issue.
After considering the information developed during the inspection, and the additional information you provided prior to, during, and in response to our questions at the Regulatory Conference, the NRC has concluded the inspection finding is appropriately characterized as Yellow (i.e., an issue with substantial importance to safety, that will result in additional NRC inspection and potentially other NRC action).
Inspection Report# : 2004009(pdf)
Inspection Report# : 2005002(pdf)
Inspection Report# : 2005011(pdf)
Inspection Report# : 2005018(pdf)
Significance: SL-IV Sep 30, 2005
 
2Q/2006 Inspection Findings - Kewaunee                                                                                                    Page 8 of 11 Identified By: NRC Item Type: NCV NonCited Violation Failure to Report in a Timely Manner an Unanalyzed Condition Involving a Potential Runout Concern With the CCW Pumps The inspectors identified a Non-Cited Violation (NCV) when the licensee failed to make a written report, within 60 days, to the NRC in accordance with 10 CFR 50.73(a)(2)(ii)(B), when an unanalyzed condition that significantly degraded plant safety was identified. Specifically, the licensee did not recognize the significance of a previously identified condition involving a potential runout issue with the component cooling water (CCW) pumps, and did not report this condition until the inspectors identified the requirement. The concern related to the CCW pump capability to provide required flow under certain conditions. Specifically, during a loss of power, and with specific system configurations, the loss of power could lead to a CCW pump runout condition. The primary cause of this finding was related to the cross-cutting area of human performance.
Because this issue affects the NRC's ability to perform its regulatory function, it was evaluated using the traditional enforcement process. The inspectors determined that this violation is of very low safety significance and because the licensee entered the issue into their corrective action program (CAP026528), this violation is being treated as an NCV consistent with Section VI.A.1 of the NRC Enforcement Policy. The licensee has taken actions to revise plant procedures to address this issue.
Inspection Report# : 2005012(pdf)
Significance:        Aug 16, 2005 Identified By: NRC Item Type: VIO Violation Potential Common Mode Failure of Auxiliary Feedwater URI 05000305/2005002-05 is associated with the design of the AFW pump's discharge pressure switches. The inspectors identified the potential for air intrusion into operating AFW pumps, potentially resulting in a common mode failure of the AFW system. This could occur during certain events where the suction source is lost prior to being able to manually swap the source of water from the CST to the SW system.
The inspectors identified a finding that was preliminarily determined to be of low to moderate safety significance, because Kewaunee failed to provide adequate design control to ensure the AFW pumps would be protected from failure due to air ingestion during tornado or seismic events; as well as from failure during potential runout conditions. The finding is also an apparent violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for not effectively providing controls to check the adequacy of the design for protecting the AFW pumps during design and license basis events.
The finding was determined to be more than minor since it impacted Mitigating System cornerstone attributes of design control (initial design and plant modifications) and the cornerstone objective to ensure availability, reliability, and capability of the AFW system to respond to events to prevent core damage. A Significance Determination Process Phase 3 risk analysis determined that this finding was preliminarily of low to moderate safety significance. The licensee has taken significant corrective actions, including extensive modifications to the system.
After considering the information developed during the inspection, the NRC has concluded the inspection finding is appropriately characterized as White (i.e., an issue with low to moderate increased importance to safety, which may require additional NRC inspections).
Inspection Report# : 2005002(pdf)
Inspection Report# : 2005010(pdf)
Inspection Report# : 2005014(pdf)
Significance:        Jul 29, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Evaluate the Effect of Modification on Turbine Driven AFW Pump Performance with Reduced Steam Pressure The inspectors identified a finding involving a Green Non-Cited Violation (NCV) of 10 CFR 50, Appendix B, Criterion III, "Design Control". The finding involved the revision of AFW pump discharge pressure trip setpoints. The licensee had not determined if the turbine driven AFW (TDAFW) pump was capable of providing the required flow under reduced steam pressure conditions prior to approving the modification. This issue could have affected the performance of the AFW system under post accident conditions.
This issue was greater than minor because it potentially affected the Mitigating System cornerstone objective of equipment capability. The issue screened as very low safety significance in Phase 1 of the SDP, because it was a design deficiency that was not found to result in a loss of function and the item was resolved prior to being in the plant conditions where the finding could have impacted the pump's performance. The licensee conducted post modification tests and revised permanent plant procedures to ensure the TDAFW pump was capable of providing the required flow under reduced steam pressure conditions.
Inspection Report# : 2005010(pdf)
Barrier Integrity Significance:        May 19, 2006 Identified By: NRC
 
2Q/2006 Inspection Findings - Kewaunee                                                                                                  Page 9 of 11 Item Type: NCV NonCited Violation Criterion III: Failed to Properly Translate the ICS Design Basis into the Technical Specifications The NRC inspectors identified a finding of very low safety significance that involved a violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to ensure that design basis calculations correctly translated the internal containment spray flow requirements into the Technical Specification allowed number of blocked internal containment spray nozzles.
The finding was greater than minor because the containment spray system could have been inoperable with the allowable pump degradation and allowable number of blocked containment spray nozzles. The finding was determined to be of very low safety significance because the internal containment spray system was determined to be operable. The cause of the finding was related to the evaluation aspect of the cross-cutting element of problem identification and resolution.
Inspection Report# : 2006007(pdf)
Emergency Preparedness Occupational Radiation Safety Significance:        Sep 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Post and Control Access Into a Locked High Radiation Area During Radiographic Activities A self-revealed finding of very low safety significance and an associated Non-Cited Violation of NRC requirements were identified when a high radiation area boundary was breached by two workers during radiography. An unnecessary radiation exposure could have been received by the workers had they not been stopped by radiography personnel as they moved toward the exposed radiographic source.
The issue 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 adequate protection of worker health and safety from exposure to radiation. The issue represents a finding of very low safety significance because there was no overexposure or substantial potential for an overexposure given the actual radiological conditions in the area coupled with the duration of the radiographic operation and the presence of radiography personnel who provided surveillance of the area, nor was the licensee's ability to assess worker dose compromised. A Non-Cited Violation of Technical Specification 6.13(a) and 10 CFR 20.1601(b) was identified for the failure to comply with the RP requirements that govern the control of access into high radiation areas. Corrective actions taken by the licensee included enhanced administrative measures by revising the radiography procedure and counseling of involved staff.
Since the cause of the problem included corrective action deficiencies from previous similar radiography boundary control events, the finding also relates to the cross-cutting area of problem identification and resolution.
Inspection Report# : 2005012(pdf)
Significance:        Sep 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Control Access Into a High Radiation Area During Radiographic Activities A self-revealed finding of very low safety significance and an associated Non-Cited Violation of NRC requirements were identified for an unposted/uncontrolled locked high radiation area in the turbine building during radiography activities. A radiography source created radiation levels such that a major portion of the whole body could have received in one hour a dose in excess of 1000 mrem in accessible areas of the turbine building, which were not posted or controlled in accordance with regulatory requirements. The areas with elevated dose rates were not positively controlled by locked door/gate, use of a barrier and flashing light, or maintained under continuous visual or electronic surveillance.
The issue 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 adequate protection of worker health and safety from exposure to radiation. The issue represents a finding of very low safety significance because there was no overexposure or substantial potential for an overexposure given the actual radiological conditions in the uncontrolled areas coupled with the duration of the radiographic shot. A Non-Cited Violation of Technical Specification 6.13(b) and 10 CFR 20.1601(b) was identified for the failure to comply with the RP requirements that govern the control of access into locked high radiation areas. Corrective actions taken by the licensee included enhanced administrative measures by revising the radiography procedure, and counseling and training of RP staff.
Inspection Report# : 2005012(pdf)
Significance:        Sep 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation
 
2Q/2006 Inspection Findings - Kewaunee                                                                                                  Page 10 of 11 Failure to Control Access Into a High Radiation Area While Moving a Radioactive Filter A self-revealed finding of very low safety significance and two associated Non-Cited Violations of regulatory requirements were identified for an unposted and uncontrolled high radiation area in an auxiliary building elevator during the transfer of a radioactive seal water injection filter. As a result of this failure, workers could have unknowingly entered a high radiation area in the elevator without knowledge of the radiological conditions.
The issue 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 adequate protection of worker health and safety from exposure to radiation. The issue represents a finding of very low safety significance because there was no overexposure or substantial potential for an overexposure given the circumstances and the actual radiological conditions in the area, nor was the licensee's ability to assess worker dose compromised. A Non-Cited Violation of Technical Specification 6.13(a) and 10 CFR 20.1601(b) was identified for the failure to comply with the RP requirements that govern the control of access into high radiation areas. This issue also represents a Non-Cited Violation of 10 CFR 20.1902(b)/20.1903 for failure to post a high radiation area.
Corrective actions taken by the licensee included enhanced administrative measures (RP Job Guide) for change-out and transport of all radioactive filters.
Inspection Report# : 2005012(pdf)
Significance:        Sep 30, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Post a High Radiation Area While Moving a Radioactive Filter A self-revealed finding of very low safety significance and two associated Non-Cited Violations of regulatory requirements were identified for an unposted and uncontrolled high radiation area in an auxiliary building elevator during the transfer of a radioactive seal water injection filter. As a result of this failure, workers could have unknowingly entered a high radiation area in the elevator without knowledge of the radiological conditions.
The issue 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 adequate protection of worker health and safety from exposure to radiation. The issue represents a finding of very low safety significance because there was no overexposure or substantial potential for an overexposure given the circumstances and the actual radiological conditions in the area, nor was the licensee's ability to assess worker dose compromised. A Non-Cited Violation of Technical Specification 6.13(a) and 10 CFR 20.1601(b) was identified for the failure to comply with the RP requirements that govern the control of access into high radiation areas. This issue also represents a Non-Cited Violation of 10 CFR 20.1902(b)/20.1903 for failure to post a high radiation area.
Corrective actions taken by the licensee included enhanced administrative measures (RP Job Guide) for change-out and transport of all radioactive filters.
Inspection Report# : 2005012(pdf)
Public Radiation Safety Significance:        Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Properly Calibrate the Waste Discharge Liquid and the Steam Generator Blowdown Radiation Monitors The inspectors identified a finding of very low safety significance and an associated violation of NRC requirements for the failure to comply with technical specification and Offsite Dose Calculation Manual (ODCM) requirements in the calibration of two liquid discharge radiation monitors listed in the ODCM. Specifically, the radiation monitor high alarm trip functions were not verified with radiation sources during instrument calibration.
The finding is greater than minor because it is associated with the plant facilities/equipment and instrumentation attribute 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 released into the public domain. Specifically, not verifying the proper operation of a radiation monitor at its high alarm trip setpoint could result in the use of a monitor that does not properly operate at the high alarm setpoint and the consequent unintended release of radioactive material to the environment in excess of regulatory limits. The finding is of very low safety significance because actual effluent discharges were adequately analyzed for radioactive content by the licensee prior to release, and the licensees ability to assess dose from radioactive waste (radwaste) liquid discharges was not impaired, nor were regulatory dose limits or As-Low-As-Is-Reasonably-Achievable dose constraints exceeded due to liquid effluent discharges.
Inspection Report# : 2006003(pdf)
Significance:        Mar 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Evaluate Degraded Flow Conditions on a SW System Radiation Monitor A finding of very low safety significance and an associated non-cited violation of the Kewaunee Technical Specifications, Section 6.8,
 
2Q/2006 Inspection Findings - Kewaunee                                                                                                  Page 11 of 11 Procedures, was identified by the inspectors for the failure to adequately evaluate degraded flow in a service water system radiation monitor. The primary cause of this finding was attributed to the cross-cutting area of human performance because procedures were available, but not followed, that would have facilitated proper performance of the task.
The licensee entered this item into its corrective action program and planned to conduct inspections of other radiation monitor sample chambers, assess the need for an in-line filter, and assess the need for a modification to correct the recurring problem with the service water radiation monitor.
The finding was greater than minor because the finding involved conditions contrary to those required by the offsite dose calculation manual.
Specifically, sampling requirements that were required to be initiated when the related radiation monitoring instrumentation should have been declared inoperable were not accomplished. The finding was of very low safety significance because no radiological releases were possible from the indicated pathways when the condition existed.
Inspection Report# : 2006002(pdf)
Physical Protection Physical Protection information not publicly available.
Miscellaneous Last modified : August 25, 2006
 
3Q/2006 Inspection Findings - Kewaunee                                                                              Page 1 of 12 Kewaunee 3Q/2006 Plant Inspection Findings Initiating Events Significance:      Jun 23, 2006 Identified By: NRC Item Type: NCV NonCited Violation Procedure for Reactor Startup Not Followed The inspectors identified a finding associated with a non-cited violation of Technical Specification 6.8.a (written procedures and administrative policies). The finding was for the licensees failure to follow approved procedures during a plant startup. The finding was of very low safety significance and there were three examples of the finding. The first example of a failure to follow approved procedures occurred when operators incorrectly marked a procedure step as not applicable and failed to execute the step. The second example of the failure to follow approved procedures occurred when operators executed procedure steps out of sequence. The third example occurred during the previous reactor startup conducted in November 2005 when operators performed procedure steps out of sequence in the same manner as executed during this plant startup. Corrective actions included placing Procedure N-0-01 on administrative hold until appropriate procedure changes could be made and training operating crews on procedure adherence.
This finding was of more than minor safety significance. Failure to comply with reactivity management requirements can lead to an uncontrolled reactivity event. In this particular event, the failure to follow the procedural sequence could have resulted in shutdown margin being less than that required by Technical Specifications. However, this finding is of very low significance because the actual shutdown margin did not go below the minimum required by Technical Specifications. This finding affected the cross-cutting issue of human performance.
Inspection Report# : 2006011(pdf)
Significance:      Jun 23, 2006 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedure for Reactor Startup The inspectors identified a finding associated with an non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, of very low safety significance associated with an event. The inspectors identified that Procedure N-0-01, Plant Startup from Cold Shutdown Condition to Hot Shutdown Condition, Revision BI, Step 4.45 was inadequate to start up the reactor for the conditions that existed on May 17, 2006. The procedure, as written, would have required the operators to dilute the reactor to a lower boron concentration than the Estimated Critical Position boron concentration prior to withdrawing the Shutdown Bank rods. Corrective actions to address this finding included placing Procedure N-0-01 on administrative hold until appropriate procedure changes could be implemented.
This finding was more than minor in safety significance because this issue, if left uncorrected, would have resulted in the core reactivity shutdown margin being less than that required by Technical Specifications. However, this finding is of very low significance because the procedure step was not executed and shutdown was never below that required by Technical Specifications. This finding affected the cross-cutting issue of human performance.
Inspection Report# : 2006011(pdf)
Significance:      May 19, 2006 Identified By: NRC Item Type: NCV NonCited Violation Criterion XVI: Failed to Identify Causes and Corrective Actions to Preclude Repetition for Significant Conditions Adverse to Quality The NRC inspectors identified a finding of very low safety significance that involved a violation of 10 CFR Part 50,
 
3Q/2006 Inspection Findings - Kewaunee                                                                                Page 2 of 12 Appendix B, Criterion XVI, Corrective Actions. Specifically, for the turbine building flooding and auxiliary feedwater air entrainment performance deficiencies, which were significant conditions adverse to quality, the licensee failed to identify the causes, and to determine corrective actions to preclude repetition.
The finding was greater than minor because the failure to identify the cause and corrective actions to preclude repetition of significant conditions adverse to quality, which led to a degraded cornerstone could result in the NRC needing to take more significant action. The finding was determined to be of very low safety significance based on management review, and the determination that no additional instances of significant conditions adverse to quality have actually occurred due to the failure to identify the causes and corrective actions for the previous performance deficiencies. The cause of the finding was related to the evaluation aspect of the cross-cutting element of problem identification and resolution.
Inspection Report# : 2006007(pdf)
Significance:        May 05, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Incorporate Operating Experience Into Preventive Maintenance Procedures The inspectors identified a finding associated with a non-cited violation (NCV) of 10 CFR 50.65 (the Maintenance Rule),
having very low safety significance for the licensees failure to incorporate into station procedures available internal and external operating experience pertaining to 4.16-kilovolt (kV) switchgear mechanically operated contact (MOC) switch linkage assemblies. As a result, preventive maintenance procedures for 4.16-kV safety- and nonsafety-related switchgear breaker cubicles were inadequate and had not been upgraded to incorporate important MOC switch linkage measurements and adjustments to be used during periodic breaker/cubicle maintenance. The licensee entered the problem with the procedures into its corrective action program for resolution. Corrective action included the revision of the procedures to incorporate the need to inspect the linkage and adjust it to within specified values.
The finding is greater than minor because it is associated with the procedure adequacy 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 operation. The finding was determined to be of very low safety significance because the transient initiator contributor is a reactor trip that did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available. The cause of the finding is related to the cross-cutting element of problem identification and resolution.
Inspection Report# : 2006010(pdf)
Significance:        Mar 31, 2006 Identified By: NRC Item Type: FIN Finding Failure to Control Loose Materials Within the Protected Area in Response to Adverse Weather Conditions A finding of very low safety significance was identified by the inspectors for the licensees failure to control loose materials within the protected area south of the transformer bays in response to adverse weather conditions. The material could have been blown into the transformers and initiate a transient. The primary cause of this finding was related to the cross-cutting area of problem identification and resolution for the failure to implement effective corrective actions in response to a similar, previous inspection finding (Inspection Report 05000305/2005008). No violation of regulatory requirements occurred.
The licensee entered this issue into its corrective action program and removed the loose material from the transformer bays.
The finding is more than minor because, if left uncorrected, the loose items would become a more significant safety concern by becoming missile hazards; thereby, increasing the likelihood of an initiating event. Additionally, the inspectors determined that this issue was associated with the procedure quality 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 because the station procedure used to control potential airborne material was too narrow in scope. The finding was of very low safety significance because the inspectors answered no to all the screening questions in the Significance Determination Process Phase 1 Screening Worksheet under the Initiating Events column.
 
3Q/2006 Inspection Findings - Kewaunee                                                                            Page 3 of 12 Inspection Report# : 2006002(pdf)
Significance:      Mar 30, 2006 Identified By: NRC Item Type: FIN Finding Failure to Adequately Evaluate an Inoperative Indicating Lamp for a Turbine control Valve A finding of very low safety significance was identified by the inspectors for the failure to adequately evaluate an inoperative indicating lamp associated with the turbine control valves. The primary cause of this finding was attributed to the cross-cutting area of human performance because procedures were available, but not followed, that would have facilitated proper performance of the task.
The licensee entered this item into its corrective action program and reviewed open work orders, provided a status update to management, and increased communications of related expectations.
The finding is greater than minor because the failure to adequately evaluate deficient conditions, if left uncorrected, would become a more significant safety concern. The finding was of very low safety significance because the inspectors answered no to all the questions in the Significance Determination Process Phase 1 Screening Worksheet under the Initiating Events column.
Inspection Report# : 2006002(pdf)
Significance:      Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Startup Procedure Resulted in an Inadvertent Carbon-Dioxide Fire Suppression Discharge and Declaration of a Notice of Unusual Event A finding of very low safety significance was self-revealed during two events when use of an inadequate plant prestartup procedure resulted in actuation of the CARDOX Carbon Dioxide system. A Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was identified for the failure to include adequate acceptance criteria in Procedure N-0-02-CLA, "Plant Prestartup Checklist". The primary cause of this finding was related to the resource attribute in the cross-cutting area of Human Performance. The licensee failed to provide the operators with quality procedures containing criteria to know when the secondary plant was appropriately aligned.
The inspectors determined that the finding was greater than minor because it involved the configuration control, human performance, and procedure quality attributes of the Initiating Events Cornerstone. Additionally the finding affected the cornerstone objective of limiting the likelihood of those events that upset plant stability during power operations.
Specifically, an incorrect lineup could exist in the secondary system resulting in an initiating event, or an unanalyzed secondary system response after a trip. The issue was of very low safety significance because the finding did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available.
Corrective actions taken by the licensee include procedural enhancements to ensure that systems are lined up properly before continuing with plant startup.
Inspection Report# : 2005017(pdf)
Mitigating Systems Significance:      Sep 30, 2006 Identified By: NRC Item Type: FIN Finding Technical Specification LCO not Entered for diesel Generators Inoperable while in Refueling Shutdown Inspection Report# : 2006004(pdf)
 
3Q/2006 Inspection Findings - Kewaunee                                                                              Page 4 of 12 Significance:      Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Reactor Protection System Surveillance Procedure Revised Without Proper Review The inspectors identified a finding of very low safety significance and an associated non-cited violation of Technical Specification 6.8, Procedures, during a review of a procedure. The licensee had changed the procedure to allow the turbine-driven auxiliary feedwater (TDAFW) pump to be considered available for risk management purposes while the pump control switch was in pull-to-lock during the performance of the surveillance procedure; however, the required Plant Operating Review Committee review and approval for the change was not obtained. Corrective actions, to date, included review of the surveillance procedure by the Plant Operating Review Committee and inclusion into the procedure of additional provisions to ensure availability of the TDAFW pump while the control switch is in pull-to-lock during performance of the procedure. The cause of this finding is related to the cross-cutting area of human performance because of the licensees failure to follow a plant procedure regarding the review and approval of safety-related procedures.
The finding is greater than minor because if left uncorrected the finding would become a more significant safety concern.
Specifically, improper application of the temporary procedure change process could lead to a more significant unreviewed, improper procedure change. Additionally, this issue is associated with the procedure quality 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). Specifically, the failure to provide adequate review and approval of a safety-related surviellance procedure prior to issuance for use and the failure to include adequate provisions to ensure availability of a safety-related component in the surveillance procedure potentially impacted equipment availability. The finding is of very low safety significance because the answer to all the screening questions in the significance determination process Phase 1 screening worksheet in the Mitigating Systems column was no.
Inspection Report# : 2006003(pdf)
Significance:      Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Leak Developed in Service Water Pipe after Wall Thinning Evaluation was Cancelled A self-revealed finding of very low safety significance and an associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, were identified on April 25, 2006, when a leak due to pipe-wall thinning was identified in a 90&deg; elbow in a service water (SW) line to the B emergency diesel generator. This wall-thinning and leak, a condition adverse to quality, resulted in the need to declare the emergency diesel generator inoperable and a shut down of the reactor to allow repair of the leak. In April 2004, a work order to inspect the elbow for wall-thinning was cancelled after wall thickness in a nearby elbow was evaluated by the licensee and deemed acceptable. The extrapolation of inspection results from one elbow to the other elbow was inappropriate. Corrective actions taken by the licensee included replacement of the failed section of SW piping, performance of additional inspections on SW piping, and replacement of other safety-related sections of SW piping. The cause of this finding is related to the cross-cutting area of problem identification and resolution because the licensee failed to promptly identify an issue potentially impacting safety-related piping.
The finding is greater than minor because it is associated with the equipment performance attribute of the Mitigating System 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). Specifically, the failure to conduct a wall-thinning evaluation in April 2004 resulted in the need to take the emergency diesel generator out-of-service and shut down the reactor to allow repair of the pipe. Additionally, the failure to inspect and correct, as necessary, wall-thinning in a safety-related system, if left uncorrected, would become a more significant safety concern through the possible development of a large system leak or the complication of the operations of a safety-related system. The finding is of very low safety significance because the answer to all the screening questions in the significance determination process Phase 1 screening worksheet in the Mitigating Systems column was no.
Inspection Report# : 2006003(pdf)
Significance:      May 19, 2006 Identified By: NRC Item Type: NCV NonCited Violation
 
3Q/2006 Inspection Findings - Kewaunee                                                                              Page 5 of 12 Criterion V: Failed to Incorporate Appropriate Acceptance Criteria for Assessing Operability of the AFW Pump The NRC inspectors identified a finding of very low safety significance that involved a violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings. Specifically, the licensee failed to incorporate appropriate acceptance criteria for assessing operability of the auxiliary feedwater pump following identification of a piping obstruction.
The finding was greater than minor because the finding was associated with the Mitigating Systems cornerstone attribute of procedure quality which affected the cornerstone objective. Specifically, the relevant procedure was not adequate to ensure the availability, reliability, and capability of the auxiliary feedwater system to respond to initiating events. The finding was determined to be of very low safety significance because subsequent evaluation of the pipe occlusions, using appropriate acceptance criteria, supported past operability of the pump. The cause of the finding was related to the evaluation aspect of the cross-cutting element of problem identification and resolution.
Inspection Report# : 2006007(pdf)
Significance:        May 19, 2006 Identified By: NRC Item Type: NCV NonCited Violation Criterion III: Failed to Correctly Translate Containment Sump Volume into Design The NRC inspectors identified a finding of very low safety significance that involved a violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to ensure that design basis calculations correctly translated the containment sump volume at the time of the switch over from the refueling water storage tank to the containment sump to ensure adequate available net positive suction head and vortex suppression for the residual heat removal pumps.
The finding was greater than minor because the finding was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective because the inadequate calculation impacted the design requirements for the new containment strainers being installed to resolve Generic Safety Issue 191. The finding was determined to be of very low safety significance because (1) the licensee normally kept the refueling water storage tank at a level above the Technical Specification minimum; (2) new strainers were not yet installed; and (3) inspector-independent calculations indicated that the pumps had adequate net positive suction head and vortex suppression, with the additional non-conservatisms incorporated. The cause of the finding was related to the corrective action aspect of the cross-cutting element of problem identification and resolution.
Inspection Report# : 2006007(pdf)
Significance:        May 19, 2006 Identified By: NRC Item Type: NCV NonCited Violation Criterion III: Failed to Verify or Check the Adequacy of the Design Canceling Design Change Request 2548 The NRC inspectors identified a finding of very low safety significance that involved a violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to properly evaluate the minimum flow requirements of the high head safety injection pumps.
The finding was greater than minor because the finding was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective as providing inadequate minimum flow to the SI pumps could result in the pumps failing under certain accident scenarios. The finding was determined to be of very low safety significance because both the licensee and the inspectors determined that the safety injection pumps remained operable with the 47 gpm minimum flow rate. The cause of the finding was related to the corrective action of the cross-cutting element of problem identification and resolution.
Inspection Report# : 2006007(pdf)
Significance:        May 05, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to maintain cable separation for cables 1N15010 and IN15012 associated with train 'B' of ICCMS
 
3Q/2006 Inspection Findings - Kewaunee                                                                                Page 6 of 12 The inspectors identified a finding associated with a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, that pertained to a modification that failed to incorporate applicable design requirements for cable separation. Nonsafety-related cables associated with train B reactor coolant pump (RCP) safety-related cable trays and cables were bundled inside the RCP breaker cubicles with train A RCP safety-related cables feeding the reactor protection system (RPS). Consequently, a fault in the train B cable/cable tray could propagate to train A. The licensee entered the problem into its corrective action program for resolution. Corrective actions included encasing the nonsafety-related cables in flexible metal conduit and confirming that other safety-related cables were not affected.
The finding is greater than minor 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). The finding was determined to be of very low safety significance because of the redundancy and coincident logic in the RPS design; and it did not represent a loss of system safety function, an actual loss of safety function of a single train, an actual loss of safety function of one or more non-technical specification trains of equipment, designated as risk significant per 10 CFR 50.65, for greater than 24 hours, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event.
Inspection Report# : 2006010(pdf)
Significance:      Mar 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation Ineffective Corrective Actions to Resolve Boric Acid Leakage from the 1A RHR Pump Flange Studs and Nuts A finding of very low safety significance and an associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified by the inspectors for ineffective identification and the initiation of corrective actions to resolve boric acid leakage from the 1A residual heat removal (RHR) pump flange studs and nuts. The primary cause of this finding was attributed to the cross-cutting area of problem identification and resolution. During a review of corrective actions associated with the licensees identification of a moderate amount of boric acid around various pump flange studs and nuts, the inspectors found that numerous prior occasions existed where the licensee had identified similar conditions yet failed to adequately identify and initiate actions to evaluate or correct this condition adverse to quality.
The licensee entered this item into its corrective action program and wrote a work order to replace the pump casing flange gasket.
The finding is greater than minor because it is associated with the equipment performance attribute of the Mitigating System 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). Additionally, failure to correct a condition adverse to quality in a safety-related system, if left uncorrected, would become a more significant safety concern.
The finding was of very low safety significance because the inspectors answered no to all the screening questions in the Significance Determination Process Phase 1 Screening Worksheet under the Mitigating Systems column.
Inspection Report# : 2006002(pdf)
Significance:      Mar 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Apply Appropriate Quality Classification to TSC Diesel Generator Modifications as Required by Procedures A finding of very low safety significance and an associated non-cited violation of the Kewaunee Technical Specifications, Section 6.8, Procedures, was identified by the inspectors during a review of plant modification Design Change Request 3490, which replaced the existing Technical Support Center diesel generator fuel oil day tank level switches with new level switches of a different design. The inspectors determined that, in accordance with procedure GNP-01.01.01, Determination of Nuclear Safety Designed Classifications, QA [Quality Assurance] Type and EQ [Environmental Qualification] Type, the new level switches should have been designated as Augmented Quality. Contrary to this, the new switches were not designated as augmented quality. The primary cause of this finding was attributed to the cross-cutting area of problem identification and resolution because of the licensees failure to take effective corrective actions for previously identified problems with its quality assurance program.
 
3Q/2006 Inspection Findings - Kewaunee                                                                              Page 7 of 12 The licensee entered this item into its corrective action program and conducted supplemental audits of quality-designated equipment, added additional related elements to an upcoming quality assurance group audit of the quality assurance program, and the conduct of a cause evaluation of related issues.
The finding is greater than minor because it is 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). Additionally, failure to comply with the provisions of nuclear safety-related procedures, if left uncorrected, would become a more significant safety concern.
The finding is of very low safety significance because the inspectors answered no to all the screening questions in the Significance Determination Process Phase 1 Screening Worksheet under the Mitigating Systems column.
Inspection Report# : 2006002(pdf)
Significance:        Mar 30, 2006 Identified By: NRC Item Type: FIN Finding Failure to Adequately Evaluate the Extent-of-Condition of Degraded Fuses in Installed Equipment A finding of very low safety significance was identified by the inspectors for the failure to adequately evaluate the extent-of-condition relative to installed equipment for a 10 CFR Part 21 notification for degraded Bussmann fuses. The primary cause of the finding was attributed to the cross-cutting area of human performance because procedures were available, but not followed, that would have facilitated proper performance of the task.
The licensee entered this item into its corrective action program and planned to review other installed fuses and to conduct an evaluation of original problem.
The finding was greater than minor because the failure to adequately evaluate the impact of potentially degraded safety-related fuses on installed equipment, if left uncorrected, would become a significant safety concern. Specifically, the condition could cause premature circuit interruptions of safety-related or risk significant mitigating components, when called upon to perform the related functions, and this is an undesirable condition. The finding was of very low safety significance because the inspectors answered no to all the screening questions in the Significance Determination Process Phase 1 Screening Worksheet under the Mitigating Systems column.
Inspection Report# : 2006002(pdf)
Significance:        Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Adjustments Performed on Safety-Related Service Water Valve 4B Without Procedure Resulted in Valve Being Declared Inoperable On October 5, 2005, a finding of very low safety significance was self-revealed when SW-4B failed to meet its In-Service Testing stroke time requirements during the performance of Surveillance Procedure SP-02-138B and an associated unplanned entry into a Technical Specification Limiting Condition for Operation occurred. The condition occurred because the licensee made adjustments to safety-related Valve SW-4B, "Turbine Building Service Water Train "B" Header Isolation," without procedural guidance to perform such adjustments. The primary cause of this finding was related to the personal attribute of the cross-cutting area of human performance because maintenance was performed without required procedures.
The finding was more than minor because performing adjustment of safety-related equipment without procedural guidance, if left uncorrected, would become a more significant safety concern. Additionally, the finding is associated with the Reactor Safety/Mitigating Systems Cornerstone attribute of Procedure Quality and effects the associated Cornerstone objective of insuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using IMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," the inspectors answered "no" to all five screening questions in the Phase 1 Screening Worksheet under the Mitigating Systems column. Therefore, this finding was of very low safety significance. A Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was identified for the failure to provide procedural guidance for adjusting SW-4B; a safety-related valve which could affect the ability of safety-related mitigating system components to perform their intended function. Corrective actions taken by the licensee include procedural revisions to
 
3Q/2006 Inspection Findings - Kewaunee                                                                            Page 8 of 12 strengthen guidance on adjustment of safety-related components.
Inspection Report# : 2005017(pdf)
Significance:        Dec 31, 2005 Identified By: NRC Item Type: NCV NonCited Violation Operator Licensing Exam Results Were Less Than Minimum Acceptable Percentage For Passing A finding of very low safety significance was identified. The finding was associated with unsatisfactory operating crew performance on the simulator during facility-administered licensed annual operator requalification examinations. Of the 7 crews evaluated, 2 did not pass their annual operating tests. The finding is of very low safety significance because the failures occurred during testing of the operators on the simulator, because there were no actual consequences to the failures, and because the crews were removed from watch-standing duties, retrained, and re-evaluated before they were authorized to return to control room watches.
Inspection Report# : 2005017(pdf)
Significance:        Dec 16, 2005 Identified By: NRC Item Type: FIN Finding No Trending of Adverse Conditions Identified During Outages The inspectors identified a finding of very low safety significance for the licensee not reviewing corrective action program documents (CAPs) during outages for potential trends of conditions adverse to quality. As part of the screening process of CAPs, the licensee assigned, as possible, CAPs to various "hot buttons." Hot buttons were searchable categories in the corrective action program computer system that had been established for various problems, such as equipment tagging errors, security door control, and reactivity management. For non-outage times, the licensee assigned a monthly number of hits for each hot button that, if exceeded for 3 months in succession, would result in the generation of a CAP to investigate a possible trend. However, as of December 16, 2005, the licensee did not use hot button action levels during outages when the number of CAPs written was much higher than during non-outage times.
This finding is greater than minor because if left uncorrected would become a more significant safety concern. This finding is not suitable for Significance Determination Process evaluation, but has been reviewed by NRC management and is determined to be a finding of very low safety significance. No violation of regulatory requirements occurred. The cause of the finding is related to the cross-cutting element of problem identification and resolution, because of not identifying potential conditions adverse to quality through trending of CAPs during outages.
Inspection Report# : 2005005(pdf)
Significance:        Dec 16, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct Procedure Non-Adherence The inspectors identified a finding of very low safety significance and a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," for the failure to take corrective action for procedure non-compliance identified during the licensee's 2004 self-assessment of the corrective action program. As a result of the assessment, CAP025194, "Corrective Action Program Procedure and Guidance Document Use," was written and documented that plant workers were not following corrective action program procedures for apparent cause evaluations and root cause evaluations, effectiveness review content, priority and due date assignments, initiator feedback, and documentation of corrective action completion. To correct this problem, corrective action CA018094, "Corrective Action Program Procedure and Guidance Document Use," was written and specified one or 2 weeks of requiring "in-hand" use by the plant staff of the corrective action program administrative procedure. However, completion of this action was delayed several times and on July 25, 2005, CAP025194 and CA018094 were closed with the only documented action taken being a July 18, 2005, meeting of the station human performance steering committee at which the licensee decided not to take action because of the pending transition to the corrective action program documents of the plant's new owner.
This finding is greater than minor because if left uncorrected would become a more significant safety concern. This finding is not suitable for Significance Determination Process evaluation, but has been reviewed by NRC management and is
 
3Q/2006 Inspection Findings - Kewaunee                                                                              Page 9 of 12 determined to be a finding of very low safety significance. The cause of the finding is related to the cross-cutting element of problem identification and resolution, because of the failure to take corrective action for non-adherence to station procedures.
Inspection Report# : 2005005(pdf)
Significance:        Dec 16, 2005 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Correct Residual Heat Removal Pump Seal Leakage A finding of very low safety significance that was a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was identified for the licensee's ineffective corrective action to repair a leak on the seal of the "B" residual heat removal (RHR) pump. The leak was identified on November 2, 2005, when the pump was stopped following the performance of a required surveillance. The leak rate exceeded leakage control program limits. A similar leak was identified on June 16, 2004, for which the licensee replaced the seal in November 2004.
This finding is greater than minor because it was associated with the "RCS (reactor coolant system) equipment and barrier performance" attribute of the barrier integrity cornerstone and does affect the cornerstone objective of providing reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Although the RCS barrier was affected, the finding did not affect the mitigation capability of the RHR system and did not contribute to the likelihood of a primary or secondary system loss of coolant accident initiator or affect the containment integrity. Therefore, the finding is of very low safety significance.
Inspection Report# : 2005005(pdf)
Significance:        Oct 06, 2005 Identified By: NRC Item Type: VIO Violation Potential Flooding in the Turbine Building Basement A review of design drawings by the inspectors revealed a direct piping connection from the turbine building sump to the trench in safeguards alley. The inspectors determined that there were no check valves located in the piping to prevent water spills in the turbine building basement from backing up into the safeguards alley. The inspectors also noted that no flood barriers specifically designed to protect equipment in the safeguards alley from flooding in the turbine building basement were installed. The inspectors requested additional information from the licensee regarding potential flooding events occurring in the safeguards alley. The licensee documented its response to the inspectors information request in Condition Evaluation (CE) 014653. This CE stated that it would take approximately 3 hours for flooding caused by AFW pump discharge to affect safety-related equipment, and such flooding could be mitigated by opening doors between the safeguards alley and the turbine building basement. The CE also stated that other sources of flooding in the turbine building basement need not be considered since such flooding events are outside the design basis of the plant.
The inspectors identified a finding that was preliminarily determined to be of substantial to high safety significance because the licensee failed to provide adequate design control to ensure that Class I equipment was protected against damage from the rupture of a pipe or tank resulting in serious flooding or excessive steam release to the extent that the Class I equipment's function is impaired. Specifically, the design of Kewaunee Power Station (KPS) did not ensure that the auxiliary feedwater (AFW) pumps, the 480-volt (V) safeguards buses, the safe shutdown panel, emergency diesel generators (EDGs) 1A and 1B, and 4160-V safeguards buses 1-5 and 1-6 would be protected from random or seismically induced failures of non-Class I systems in the turbine building. The finding is also an apparent violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for not ensuring that the design of KPS prevented turbine building flooding from impacting multiple safety related equipment trains needed for safe shutdown of the plant. The inspectors determined that a primary cause of this finding was related to the cross-cutting area of Problem Identification and Resolution, because there was an earlier opportunity to discover and correct this issue based on the licensee's 2003 experience when minor flooding from the turbine building had challenged safety equipment located adjacent to the turbine building basement.
The finding was more than minor because it impacted Mitigating Systems cornerstone attributes of design control (initial design and plant modifications) and protection against external factors (internal flood hazards and seismic events) and it impacted the Mitigating Systems cornerstone objective to ensure availability, reliability and capability of multiple trains of safety related equipment to respond to events to prevent core damage. A Significance Determination Process Phase 3 risk
 
3Q/2006 Inspection Findings - Kewaunee                                                                              Page 10 of 12 analysis determined that this finding was preliminarily of substantial to high safety significance. The licensee has taken significant corrective actions, including extensive system and structural modifications to address this issue.
After considering the information developed during the inspection, and the additional information you provided prior to, during, and in response to our questions at the Regulatory Conference, the NRC has concluded the inspection finding is appropriately characterized as Yellow (i.e., an issue with substantial importance to safety, that will result in additional NRC inspection and potentially other NRC action).
Inspection Report# : 2004009(pdf)
Inspection Report# : 2005002(pdf)
Inspection Report# : 2005011(pdf)
Inspection Report# : 2005018(pdf)
Inspection Report# : 2006015(pdf)
Significance:        Aug 16, 2005 Identified By: NRC Item Type: VIO Violation Potential Common Mode Failure of Auxiliary Feedwater URI 05000305/2005002-05 is associated with the design of the AFW pump's discharge pressure switches. The inspectors identified the potential for air intrusion into operating AFW pumps, potentially resulting in a common mode failure of the AFW system. This could occur during certain events where the suction source is lost prior to being able to manually swap the source of water from the CST to the SW system.
The inspectors identified a finding that was preliminarily determined to be of low to moderate safety significance, because Kewaunee failed to provide adequate design control to ensure the AFW pumps would be protected from failure due to air ingestion during tornado or seismic events; as well as from failure during potential runout conditions. The finding is also an apparent violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for not effectively providing controls to check the adequacy of the design for protecting the AFW pumps during design and license basis events.
The finding was determined to be more than minor since it impacted Mitigating System cornerstone attributes of design control (initial design and plant modifications) and the cornerstone objective to ensure availability, reliability, and capability of the AFW system to respond to events to prevent core damage. A Significance Determination Process Phase 3 risk analysis determined that this finding was preliminarily of low to moderate safety significance. The licensee has taken significant corrective actions, including extensive modifications to the system.
After considering the information developed during the inspection, the NRC has concluded the inspection finding is appropriately characterized as White (i.e., an issue with low to moderate increased importance to safety, which may require additional NRC inspections).
Inspection Report# : 2006015(pdf)
Inspection Report# : 2005002(pdf)
Inspection Report# : 2005010(pdf)
Inspection Report# : 2005014(pdf)
Barrier Integrity Significance:        May 19, 2006 Identified By: NRC Item Type: NCV NonCited Violation Criterion III: Failed to Properly Translate the ICS Design Basis into the Technical Specifications The NRC inspectors identified a finding of very low safety significance that involved a violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to ensure that design basis calculations correctly translated the internal containment spray flow requirements into the Technical Specification allowed number of blocked internal containment spray nozzles.
 
3Q/2006 Inspection Findings - Kewaunee                                                                            Page 11 of 12 The finding was greater than minor because the containment spray system could have been inoperable with the allowable pump degradation and allowable number of blocked containment spray nozzles. The finding was determined to be of very low safety significance because the internal containment spray system was determined to be operable. The cause of the finding was related to the evaluation aspect of the cross-cutting element of problem identification and resolution.
Inspection Report# : 2006007(pdf)
Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Significance:      Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Properly Calibrate the Waste Discharge Liquid and the Steam Generator Blowdown Radiation Monitors The inspectors identified a finding of very low safety significance and an associated violation of NRC requirements for the failure to comply with technical specification and Offsite Dose Calculation Manual (ODCM) requirements in the calibration of two liquid discharge radiation monitors listed in the ODCM. Specifically, the radiation monitor high alarm trip functions were not verified with radiation sources during instrument calibration.
The finding is greater than minor because it is associated with the plant facilities/equipment and instrumentation attribute 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 released into the public domain. Specifically, not verifying the proper operation of a radiation monitor at its high alarm trip setpoint could result in the use of a monitor that does not properly operate at the high alarm setpoint and the consequent unintended release of radioactive material to the environment in excess of regulatory limits. The finding is of very low safety significance because actual effluent discharges were adequately analyzed for radioactive content by the licensee prior to release, and the licensees ability to assess dose from radioactive waste (radwaste) liquid discharges was not impaired, nor were regulatory dose limits or As-Low-As-Is-Reasonably-Achievable dose constraints exceeded due to liquid effluent discharges.
Inspection Report# : 2006003(pdf)
Significance:      Mar 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Evaluate Degraded Flow Conditions on a SW System Radiation Monitor A finding of very low safety significance and an associated non-cited violation of the Kewaunee Technical Specifications, Section 6.8, Procedures, was identified by the inspectors for the failure to adequately evaluate degraded flow in a service water system radiation monitor. The primary cause of this finding was attributed to the cross-cutting area of human performance because procedures were available, but not followed, that would have facilitated proper performance of the task.
The licensee entered this item into its corrective action program and planned to conduct inspections of other radiation monitor sample chambers, assess the need for an in-line filter, and assess the need for a modification to correct the recurring problem with the service water radiation monitor.
 
3Q/2006 Inspection Findings - Kewaunee                                                                        Page 12 of 12 The finding was greater than minor because the finding involved conditions contrary to those required by the offsite dose calculation manual. Specifically, sampling requirements that were required to be initiated when the related radiation monitoring instrumentation should have been declared inoperable were not accomplished. The finding was of very low safety significance because no radiological releases were possible from the indicated pathways when the condition existed.
Inspection Report# : 2006002(pdf)
Physical Protection Physical Protection information not publicly available.
Miscellaneous Last modified : December 21, 2006
 
4Q/2006 Inspection Findings - Kewaunee                                                                              Page 1 of 11 Kewaunee 4Q/2006 Plant Inspection Findings Initiating Events Significance:      Dec 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Pre-Fire Strategy Identified in Cable Spreading Room A finding of very low safety significance and an associated non-cited violation of Technical Specification 6.8, Procedures, was identified by the inspectors for the failure to identify radiological and toxic hazards in the cable spreading area fire zone pre-fire strategy. These hazards were from a radioactively contaminated lead pipe in the fire zone that could melt during certain fire scenarios. As part of corrective actions, the licensee appropriately revised the strategy.
The issue was entered into the licensees corrective action program.
The finding is greater than minor because it was associated with the external factors - fire 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 provide adequate warnings and guidance in the pre-fire plan related to these hazards could have adversely impacted the fire brigades ability to properly respond to a fire. This impact could increase the likelihood of damage to equipment, causing an upset pf plant stability. NRC management review determined the finding to be of very low safety significance (Green),
due to the extensive training provided to fire brigade members to deal with unexpected contingencies. This finding has a cross-cutting aspect in the area of human performance because the licensee failed to provide complete, accurate, and up-to-date pre-fire strategies for the fire brigade to respond to a fire.
Inspection Report# : 2006005 (pdf)
Significance:      Jun 23, 2006 Identified By: NRC Item Type: NCV NonCited Violation Procedure for Reactor Startup Not Followed The inspectors identified a finding associated with a non-cited violation of Technical Specification 6.8.a (written procedures and administrative policies). The finding was for the licensees failure to follow approved procedures during a plant startup. The finding was of very low safety significance and there were three examples of the finding. The first example of a failure to follow approved procedures occurred when operators incorrectly marked a procedure step as not applicable and failed to execute the step. The second example of the failure to follow approved procedures occurred when operators executed procedure steps out of sequence. The third example occurred during the previous reactor startup conducted in November 2005 when operators performed procedure steps out of sequence in the same manner as executed during this plant startup. Corrective actions included placing Procedure N-0-01 on administrative hold until appropriate procedure changes could be made and training operating crews on procedure adherence.
This finding was of more than minor safety significance. Failure to comply with reactivity management requirements can lead to an uncontrolled reactivity event. In this particular event, the failure to follow the procedural sequence could have resulted in shutdown margin being less than that required by Technical Specifications. However, this finding is of very low significance because the actual shutdown margin did not go below the minimum required by Technical Specifications. This finding affected the cross-cutting issue of human performance.
Inspection Report# : 2006011 (pdf)
Significance:      Jun 23, 2006 Identified By: NRC Item Type: NCV NonCited Violation
 
4Q/2006 Inspection Findings - Kewaunee                                                                                Page 2 of 11 Inadequate Procedure for Reactor Startup The inspectors identified a finding associated with an non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, of very low safety significance associated with an event. The inspectors identified that Procedure N-0-01, Plant Startup from Cold Shutdown Condition to Hot Shutdown Condition, Revision BI, Step 4.45 was inadequate to start up the reactor for the conditions that existed on May 17, 2006. The procedure, as written, would have required the operators to dilute the reactor to a lower boron concentration than the Estimated Critical Position boron concentration prior to withdrawing the Shutdown Bank rods. Corrective actions to address this finding included placing Procedure N-0-01 on administrative hold until appropriate procedure changes could be implemented.
This finding was more than minor in safety significance because this issue, if left uncorrected, would have resulted in the core reactivity shutdown margin being less than that required by Technical Specifications. However, this finding is of very low significance because the procedure step was not executed and shutdown was never below that required by Technical Specifications. This finding affected the cross-cutting issue of human performance.
Inspection Report# : 2006011 (pdf)
Significance:        May 19, 2006 Identified By: NRC Item Type: NCV NonCited Violation Criterion XVI: Failed to Identify Causes and Corrective Actions to Preclude Repetition for Significant Conditions Adverse to Quality The NRC inspectors identified a finding of very low safety significance that involved a violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions. Specifically, for the turbine building flooding and auxiliary feedwater air entrainment performance deficiencies, which were significant conditions adverse to quality, the licensee failed to identify the causes, and to determine corrective actions to preclude repetition.
The finding was greater than minor because the failure to identify the cause and corrective actions to preclude repetition of significant conditions adverse to quality, which led to a degraded cornerstone could result in the NRC needing to take more significant action. The finding was determined to be of very low safety significance based on management review, and the determination that no additional instances of significant conditions adverse to quality have actually occurred due to the failure to identify the causes and corrective actions for the previous performance deficiencies. The cause of the finding was related to the evaluation aspect of the cross-cutting element of problem identification and resolution.
Inspection Report# : 2006007 (pdf)
Significance:        May 05, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Incorporate Operating Experience Into Preventive Maintenance Procedures The inspectors identified a finding associated with a non-cited violation (NCV) of 10 CFR 50.65 (the Maintenance Rule),
having very low safety significance for the licensees failure to incorporate into station procedures available internal and external operating experience pertaining to 4.16-kilovolt (kV) switchgear mechanically operated contact (MOC) switch linkage assemblies. As a result, preventive maintenance procedures for 4.16-kV safety- and nonsafety-related switchgear breaker cubicles were inadequate and had not been upgraded to incorporate important MOC switch linkage measurements and adjustments to be used during periodic breaker/cubicle maintenance. The licensee entered the problem with the procedures into its corrective action program for resolution. Corrective action included the revision of the procedures to incorporate the need to inspect the linkage and adjust it to within specified values.
The finding is greater than minor because it is associated with the procedure adequacy 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 operation. The finding was determined to be of very low safety significance because the transient initiator contributor is a reactor trip that did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available. The cause of the finding is related to the cross-cutting element of problem identification and resolution.
Inspection Report# : 2006010 (pdf)
 
4Q/2006 Inspection Findings - Kewaunee                                                                              Page 3 of 11 Significance:      Mar 31, 2006 Identified By: NRC Item Type: FIN Finding Failure to Control Loose Materials Within the Protected Area in Response to Adverse Weather Conditions A finding of very low safety significance was identified by the inspectors for the licensees failure to control loose materials within the protected area south of the transformer bays in response to adverse weather conditions. The material could have been blown into the transformers and initiate a transient. The primary cause of this finding was related to the cross-cutting area of problem identification and resolution for the failure to implement effective corrective actions in response to a similar, previous inspection finding (Inspection Report 05000305/2005008). No violation of regulatory requirements occurred.
The licensee entered this issue into its corrective action program and removed the loose material from the transformer bays.
The finding is more than minor because, if left uncorrected, the loose items would become a more significant safety concern by becoming missile hazards; thereby, increasing the likelihood of an initiating event. Additionally, the inspectors determined that this issue was associated with the procedure quality 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 because the station procedure used to control potential airborne material was too narrow in scope. The finding was of very low safety significance because the inspectors answered no to all the screening questions in the Significance Determination Process Phase 1 Screening Worksheet under the Initiating Events column.
Inspection Report# : 2006002 (pdf)
Significance:      Mar 30, 2006 Identified By: NRC Item Type: FIN Finding Failure to Adequately Evaluate an Inoperative Indicating Lamp for a Turbine Control Valve A finding of very low safety significance was identified by the inspectors for the failure to adequately evaluate an inoperative indicating lamp associated with the turbine control valves. The primary cause of this finding was attributed to the cross-cutting area of human performance because procedures were available, but not followed, that would have facilitated proper performance of the task.
The licensee entered this item into its corrective action program and reviewed open work orders, provided a status update to management, and increased communications of related expectations.
The finding is greater than minor because the failure to adequately evaluate deficient conditions, if left uncorrected, would become a more significant safety concern. The finding was of very low safety significance because the inspectors answered no to all the questions in the Significance Determination Process Phase 1 Screening Worksheet under the Initiating Events column.
Inspection Report# : 2006002 (pdf)
Mitigating Systems Significance:      Dec 14, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify Emergency Diesel Generator Air Intake Temperature Limitations Impact Upon Ability to Meet Technical Specification Surveillance Requirements A finding of very low safety significance was identified by the inspectors for a violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action. The licensee failed to identify the impact of air intake temperature limitation on the
 
4Q/2006 Inspection Findings - Kewaunee                                                                            Page 4 of 11 ability of the emergency diesel generators to meet Technical Specification surveillance loading requirements at elevated temperatures. Once identified, the licensee established 75 degrees Fahrenheit as a maximum outside temperature for emergency diesel generator operability. The primary cause of this violation was related to the cross-cutting area of Problem Identification and Resolution, because the licensee failed to ensure that an issue potentially impacting nuclear safety was promptly identified, fully evaluated, and that actions were taken to address safety issues in a timely manner, commensurate with their significance.
The issue was more than minor because the failure to identify that the emergency diesel generators would not be able to meet Technical Specification surveillance requirements at elevated temperatures could have resulted in the emergency diesel generators being considered operable when, in fact, they had less operational margin than required by Technical Specifications. The issue was of very low safety significance because both of the emergency diesel generators were determined to be capable of carrying their respective design basis accident loads below the outside temperature limitations that the licensee had in place. The issue was a NCV of 10 CFR Part 50, Appendix B, Criterion XVI, which required that conditions adverse to quality are promptly identified and corrected.
Inspection Report# : 2006016 (pdf)
Significance:      Dec 14, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Provide Suppression for Safe Shutdown Equipment in Appendix R, III.G.3 Area A finding of very low safety significance was identified by the inspectors for a violation of 10 CFR Part 50, Appendix R, Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979. The licensee failed to provide required fire suppression coverage in fire zone AX-32 for the safe shutdown functions of source range monitoring, isolation of a steam generator blowdown line, and pressurizer level instrumentation. Once identified, the licensee entered the issue into their corrective action program and implemented compensatory measures.
This issue was more than minor because the failure to provide suppression for redundant trains of safe shutdown equipment increased the likelihood that alternative shutdown methods would have to be used in the event of a fire. The issue was of very low safety significance because of the mitigating systems, which would have remained available in the event of a fire.
The issue was a NCV of 10 CFR Part 50, Appendix R, Section III.G.3, which required fixed suppression systems for alternative shutdown areas such as fire zone AX-32.
Inspection Report# : 2006016 (pdf)
Significance:      Oct 23, 2006 Identified By: NRC Item Type: NCV NonCited Violation Scaffolding in Contact with the Safety Injection System Affects Operability A finding of very low safety significance and an associated non-cited violation of Technical Specification 6.8, Procedures, was identified by the inspectors on October 23, 2006, for the failure to install scaffolding in accordance with station procedures. Specifically, scaffolding was installed inside containment that was too close to, or was in contact with, safety injection system components and piping. As part of corrective actions, the licensee removed the scaffolding and enhance the station procedure for scaffolding. The issue was entered into the licensees corrective action program.
This finding is greater than minor because it was 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). Specifically, improperly positioned scaffolding could have impeded or prevented proper operation of the safety injection system during an accident.
The finding was of very low safety significance (Green) because it did not require a quantitative assessment. This finding has a cross-cutting aspect in the area of human performances because personnel did not follow the procedure for scaffolding.
Inspection Report# : 2006005 (pdf)
Significance:      Oct 11, 2006 Identified By: NRC
 
4Q/2006 Inspection Findings - Kewaunee                                                                              Page 5 of 11 Item Type: NCV NonCited Violation Inadvertent Drain Down of the Reactor Coolant system During Fill and Vent of the Containment Spray System A finding of very low safety significance and an associated non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed during performance of a plant safety-related procedure to fill and vent the containment spray system, on October 11, 2006, when water was inappropriately diverted from the reactor coolant system to the residual heat removal system. As part of corrective actions, the licensee revised the procedure to ensure the systems were properly aligned during fill and vent activities. The issue was entered into the licensees corrective action program.
This finding is greater than minor because if left uncorrected it would become a more significant safety concern in that the use of other inadequate procedures could have rendered inoperable important mitigating equipment, such as the containment spray and residual heat removal systems. Additionally, the finding was associated with the procedure quality and configuration control attributes 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 was determined to be of very low safety significance (Green) because it did not require a quantitative assessment. This finding has a cross-cutting aspect in the area of human performances because the licensee failed to provide complete, accurate, and up-to-date procedures to fill and vent the containment spray system.
Inspection Report# : 2006005 (pdf)
Significance:        Sep 30, 2006 Identified By: NRC Item Type: FIN Finding Technical Specification LCO Not Entered for Diesel Generators Inoperable While in Refueling Shutdown A finding of very low safety significance (Green) was identified by the inspectors when the licensee failed to properly apply shutdown Technical Specifications (TSs) for the residual heat removal (RHR) system with both emergency diesel generators (EDGs) declared inoperable. While reviewing startup preparations being made for a mode change, the inspectors identified that TSs required both RHR systems to be operable and that both EDGs were inoperable due to tornado failure susceptibilities, thereby rendering both trains of RHR inoperable as required by the related power requirements TS. The licensee concurred with the inspectors observations, prevented the mode change, and issued the related licensee event report. Corrective actions, to date, included restoration of EDG operability prior to making a mode change and procedural enhancements.
The inspectors determined that the finding is greater than minor because if left uncorrected it would become a more significant safety issue: the licensee would have made a mode change without the required operable equipment. This finding was of very low safety significance because the licensee returned the EDGs to operability prior to making any mode changes, no violation of NRC requirements was identified, and the finding did not require a quantitative assessment using Check List 4 for PWR Shutdown Operation with Time to Boil >2 hours and Inventory in the Pressurizer. The cause of this finding was related to the crosscutting area of human performance because procedures, specifically TSs, were available but not followed, that would have facilitated the proper performance of the task.
Inspection Report# : 2006004 (pdf)
Significance:        Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Reactor Protection System Surveillance Procedure Revised Without Proper Review The inspectors identified a finding of very low safety significance and an associated non-cited violation of Technical Specification 6.8, Procedures, during a review of a procedure. The licensee had changed the procedure to allow the turbine-driven auxiliary feedwater (TDAFW) pump to be considered available for risk management purposes while the pump control switch was in pull-to-lock during the performance of the surveillance procedure; however, the required Plant Operating Review Committee review and approval for the change was not obtained. Corrective actions, to date, included review of the surveillance procedure by the Plant Operating Review Committee and inclusion into the procedure of additional provisions to ensure availability of the TDAFW pump while the control switch is in pull-to-lock during performance of the procedure. The cause of this finding is related to the cross-cutting area of human performance because of the licensees failure to follow a plant procedure regarding the review and approval of safety-related procedures.
 
4Q/2006 Inspection Findings - Kewaunee                                                                              Page 6 of 11 The finding is greater than minor because if left uncorrected the finding would become a more significant safety concern.
Specifically, improper application of the temporary procedure change process could lead to a more significant unreviewed, improper procedure change. Additionally, this issue is associated with the procedure quality 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). Specifically, the failure to provide adequate review and approval of a safety-related surviellance procedure prior to issuance for use and the failure to include adequate provisions to ensure availability of a safety-related component in the surveillance procedure potentially impacted equipment availability. The finding is of very low safety significance because the answer to all the screening questions in the significance determination process Phase 1 screening worksheet in the Mitigating Systems column was no.
Inspection Report# : 2006003 (pdf)
Significance:        Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Leak Developed in Service Water Pipe after Wall Thinning Evaluation was Cancelled A self-revealed finding of very low safety significance and an associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, were identified on April 25, 2006, when a leak due to pipe-wall thinning was identified in a 90&deg; elbow in a service water (SW) line to the B emergency diesel generator. This wall-thinning and leak, a condition adverse to quality, resulted in the need to declare the emergency diesel generator inoperable and a shut down of the reactor to allow repair of the leak. In April 2004, a work order to inspect the elbow for wall-thinning was cancelled after wall thickness in a nearby elbow was evaluated by the licensee and deemed acceptable. The extrapolation of inspection results from one elbow to the other elbow was inappropriate. Corrective actions taken by the licensee included replacement of the failed section of SW piping, performance of additional inspections on SW piping, and replacement of other safety-related sections of SW piping. The cause of this finding is related to the cross-cutting area of problem identification and resolution because the licensee failed to promptly identify an issue potentially impacting safety-related piping.
The finding is greater than minor because it is associated with the equipment performance attribute of the Mitigating System 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). Specifically, the failure to conduct a wall-thinning evaluation in April 2004 resulted in the need to take the emergency diesel generator out-of-service and shut down the reactor to allow repair of the pipe. Additionally, the failure to inspect and correct, as necessary, wall-thinning in a safety-related system, if left uncorrected, would become a more significant safety concern through the possible development of a large system leak or the complication of the operations of a safety-related system. The finding is of very low safety significance because the answer to all the screening questions in the significance determination process Phase 1 screening worksheet in the Mitigating Systems column was no.
Inspection Report# : 2006003 (pdf)
Significance:        May 19, 2006 Identified By: NRC Item Type: NCV NonCited Violation Criterion V: Failed to Incorporate Appropriate Acceptance Criteria for Assessing Operability of the AFW Pump The NRC inspectors identified a finding of very low safety significance that involved a violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings. Specifically, the licensee failed to incorporate appropriate acceptance criteria for assessing operability of the auxiliary feedwater pump following identification of a piping obstruction.
The finding was greater than minor because the finding was associated with the Mitigating Systems cornerstone attribute of procedure quality which affected the cornerstone objective. Specifically, the relevant procedure was not adequate to ensure the availability, reliability, and capability of the auxiliary feedwater system to respond to initiating events. The finding was determined to be of very low safety significance because subsequent evaluation of the pipe occlusions, using appropriate acceptance criteria, supported past operability of the pump. The cause of the finding was related to the evaluation aspect of the cross-cutting element of problem identification and resolution.
Inspection Report# : 2006007 (pdf)
 
4Q/2006 Inspection Findings - Kewaunee                                                                                Page 7 of 11 Significance:      May 19, 2006 Identified By: NRC Item Type: NCV NonCited Violation Criterion III: Failed to Correctly Translate Containment Sump Volume into Design The NRC inspectors identified a finding of very low safety significance that involved a violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to ensure that design basis calculations correctly translated the containment sump volume at the time of the switch over from the refueling water storage tank to the containment sump to ensure adequate available net positive suction head and vortex suppression for the residual heat removal pumps.
The finding was greater than minor because the finding was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective because the inadequate calculation impacted the design requirements for the new containment strainers being installed to resolve Generic Safety Issue 191. The finding was determined to be of very low safety significance because (1) the licensee normally kept the refueling water storage tank at a level above the Technical Specification minimum; (2) new strainers were not yet installed; and (3) inspector-independent calculations indicated that the pumps had adequate net positive suction head and vortex suppression, with the additional non-conservatisms incorporated. The cause of the finding was related to the corrective action aspect of the cross-cutting element of problem identification and resolution.
Inspection Report# : 2006007 (pdf)
Significance:      May 19, 2006 Identified By: NRC Item Type: NCV NonCited Violation Criterion III: Failed to Verify or Check the Adequacy of the Design Canceling Design Change Request 2548 The NRC inspectors identified a finding of very low safety significance that involved a violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to properly evaluate the minimum flow requirements of the high head safety injection pumps.
The finding was greater than minor because the finding was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective as providing inadequate minimum flow to the SI pumps could result in the pumps failing under certain accident scenarios. The finding was determined to be of very low safety significance because both the licensee and the inspectors determined that the safety injection pumps remained operable with the 47 gpm minimum flow rate. The cause of the finding was related to the corrective action of the cross-cutting element of problem identification and resolution.
Inspection Report# : 2006007 (pdf)
Significance:      May 05, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Separation of Cables The inspectors identified a finding associated with a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, that pertained to a modification that failed to incorporate applicable design requirements for cable separation. Nonsafety-related cables associated with train B reactor coolant pump (RCP) safety-related cable trays and cables were bundled inside the RCP breaker cubicles with train A RCP safety-related cables feeding the reactor protection system (RPS). Consequently, a fault in the train B cable/cable tray could propagate to train A. The licensee entered the problem into its corrective action program for resolution. Corrective actions included encasing the nonsafety-related cables in flexible metal conduit and confirming that other safety-related cables were not affected.
The finding is greater than minor 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). The finding was determined to be of very low safety significance because of the redundancy and coincident logic in the RPS design; and it did not represent a loss of system safety function, an actual loss of safety function of a single train, an actual loss of safety function of one or more non-technical specification trains of equipment, designated as risk significant per 10 CFR 50.65, for greater than 24
 
4Q/2006 Inspection Findings - Kewaunee                                                                              Page 8 of 11 hours, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event.
Inspection Report# : 2006010 (pdf)
Significance:      Mar 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation Ineffective Corrective Actions to Resolve Boric Acid Leakage from the 1A RHR Pump Flange Studs and Nuts A finding of very low safety significance and an associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified by the inspectors for ineffective identification and the initiation of corrective actions to resolve boric acid leakage from the 1A residual heat removal (RHR) pump flange studs and nuts. The primary cause of this finding was attributed to the cross-cutting area of problem identification and resolution. During a review of corrective actions associated with the licensees identification of a moderate amount of boric acid around various pump flange studs and nuts, the inspectors found that numerous prior occasions existed where the licensee had identified similar conditions yet failed to adequately identify and initiate actions to evaluate or correct this condition adverse to quality.
The licensee entered this item into its corrective action program and wrote a work order to replace the pump casing flange gasket.
The finding is greater than minor because it is associated with the equipment performance attribute of the Mitigating System 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). Additionally, failure to correct a condition adverse to quality in a safety-related system, if left uncorrected, would become a more significant safety concern.
The finding was of very low safety significance because the inspectors answered no to all the screening questions in the Significance Determination Process Phase 1 Screening Worksheet under the Mitigating Systems column.
Inspection Report# : 2006002 (pdf)
Significance:      Mar 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Apply Appropriate Quality Classification to TSC Diesel Generator Modifications as Required by Procedures A finding of very low safety significance and an associated non-cited violation of the Kewaunee Technical Specifications, Section 6.8, Procedures, was identified by the inspectors during a review of plant modification Design Change Request 3490, which replaced the existing Technical Support Center diesel generator fuel oil day tank level switches with new level switches of a different design. The inspectors determined that, in accordance with procedure GNP-01.01.01, Determination of Nuclear Safety Designed Classifications, QA [Quality Assurance] Type and EQ [Environmental Qualification] Type, the new level switches should have been designated as Augmented Quality. Contrary to this, the new switches were not designated as augmented quality. The primary cause of this finding was attributed to the cross-cutting area of problem identification and resolution because of the licensees failure to take effective corrective actions for previously identified problems with its quality assurance program.
The licensee entered this item into its corrective action program and conducted supplemental audits of quality-designated equipment, added additional related elements to an upcoming quality assurance group audit of the quality assurance program, and the conduct of a cause evaluation of related issues.
The finding is greater than minor because it is 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). Additionally, failure to comply with the provisions of nuclear safety-related procedures, if left uncorrected, would become a more significant safety concern.
The finding is of very low safety significance because the inspectors answered no to all the screening questions in the Significance Determination Process Phase 1 Screening Worksheet under the Mitigating Systems column.
Inspection Report# : 2006002 (pdf)
Significance:      Mar 30, 2006
 
4Q/2006 Inspection Findings - Kewaunee                                                                              Page 9 of 11 Identified By: NRC Item Type: FIN Finding Failure to Adequately Evaluate the Extent-of-Condition of Degraded Fuses in Installed Equipment A finding of very low safety significance was identified by the inspectors for the failure to adequately evaluate the extent-of-condition relative to installed equipment for a 10 CFR Part 21 notification for degraded Bussmann fuses. The primary cause of the finding was attributed to the cross-cutting area of human performance because procedures were available, but not followed, that would have facilitated proper performance of the task.
The licensee entered this item into its corrective action program and planned to review other installed fuses and to conduct an evaluation of original problem.
The finding was greater than minor because the failure to adequately evaluate the impact of potentially degraded safety-related fuses on installed equipment, if left uncorrected, would become a significant safety concern. Specifically, the condition could cause premature circuit interruptions of safety-related or risk significant mitigating components, when called upon to perform the related functions, and this is an undesirable condition. The finding was of very low safety significance because the inspectors answered no to all the screening questions in the Significance Determination Process Phase 1 Screening Worksheet under the Mitigating Systems column.
Inspection Report# : 2006002 (pdf)
Barrier Integrity Significance: SL-IV Dec 14, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Fully Update Updated Safety Analysis Report A finding of very low safety significance was identified for the licensees failure to adequately update the Update Safety Analysis Report (USAR) in accordance to 10 CFR 50.71, Maintenance of Records, Making of Reports. The licensee failed to update the USAR to fully reflect changes and analyses made in response to NRC Generic Letter (GL) 96-06, Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions. Once identified, the licensee entered this issue into their corrective action program. The primary cause of this violation was related to the cross-cutting area of Human Performance because the licensee failed to ensure that personnel, equipment, procedures, and other resources were available and adequate to assure nuclear safety. Specifically, the licensee failed to provide adequate engineering procedural guidance concerning the required content of USAR updates.
Because this issue potentially impacted the NRCs ability to perform its regulatory function, this finding was evaluated using the traditional enforcement process. The finding was determined to be more than minor because of the failure to provide complete licensing and design basis information in the USAR could result in either the licensee making an inappropriate licensing interpretation or the NRC making an inappropriate regulatory decision based on incomplete information in the USAR. The issue was of very low safety significance because no instances were identified where the failure to appropriately update the USAR impeded or influenced a regulatory decision, or resulted in an actual loss of safety function. The issue was a NCV of 10 CFR 50.71(e) which required that the USAR be updated to include the effects of all analyses of new safety issues performed by or on behalf of the licensee at Commission request.
Inspection Report# : 2006016 (pdf)
Significance:        May 19, 2006 Identified By: NRC Item Type: NCV NonCited Violation Criterion III: Failed to Properly Translate the ICS Design Basis into the Technical Specifications The NRC inspectors identified a finding of very low safety significance that involved a violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to ensure that design basis calculations correctly translated the internal containment spray flow requirements into the Technical Specification allowed number of blocked internal containment spray nozzles.
 
4Q/2006 Inspection Findings - Kewaunee                                                                            Page 10 of 11 The finding was greater than minor because the containment spray system could have been inoperable with the allowable pump degradation and allowable number of blocked containment spray nozzles. The finding was determined to be of very low safety significance because the internal containment spray system was determined to be operable. The cause of the finding was related to the evaluation aspect of the cross-cutting element of problem identification and resolution.
Inspection Report# : 2006007 (pdf)
Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Significance:      Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Properly Calibrate the Waste Discharge Liquid and the Steam Generator Blowdown Radiation Monitors The inspectors identified a finding of very low safety significance and an associated violation of NRC requirements for the failure to comply with technical specification and Offsite Dose Calculation Manual (ODCM) requirements in the calibration of two liquid discharge radiation monitors listed in the ODCM. Specifically, the radiation monitor high alarm trip functions were not verified with radiation sources during instrument calibration.
The finding is greater than minor because it is associated with the plant facilities/equipment and instrumentation attribute 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 released into the public domain. Specifically, not verifying the proper operation of a radiation monitor at its high alarm trip setpoint could result in the use of a monitor that does not properly operate at the high alarm setpoint and the consequent unintended release of radioactive material to the environment in excess of regulatory limits. The finding is of very low safety significance because actual effluent discharges were adequately analyzed for radioactive content by the licensee prior to release, and the licensees ability to assess dose from radioactive waste (radwaste) liquid discharges was not impaired, nor were regulatory dose limits or As-Low-As-Is-Reasonably-Achievable dose constraints exceeded due to liquid effluent discharges.
Inspection Report# : 2006003 (pdf)
Significance:      Mar 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Evaluate Degraded Flow Conditions on a SW System Radiation Monitor A finding of very low safety significance and an associated non-cited violation of the Kewaunee Technical Specifications, Section 6.8, Procedures, was identified by the inspectors for the failure to adequately evaluate degraded flow in a service water system radiation monitor. The primary cause of this finding was attributed to the cross-cutting area of human performance because procedures were available, but not followed, that would have facilitated proper performance of the task.
The licensee entered this item into its corrective action program and planned to conduct inspections of other radiation monitor sample chambers, assess the need for an in-line filter, and assess the need for a modification to correct the recurring problem with the service water radiation monitor.
The finding was greater than minor because the finding involved conditions contrary to those required by the offsite dose
 
4Q/2006 Inspection Findings - Kewaunee                                                                        Page 11 of 11 calculation manual. Specifically, sampling requirements that were required to be initiated when the related radiation monitoring instrumentation should have been declared inoperable were not accomplished. The finding was of very low safety significance because no radiological releases were possible from the indicated pathways when the condition existed.
Inspection Report# : 2006002 (pdf)
Physical Protection Physical Protection information not publicly available.
Miscellaneous Last modified : March 01, 2007
 
Kewaunee 1Q/2007 Plant Inspection Findings Initiating Events Significance:      Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Implement E-0-05, "Response to Natural Events," During a High Wind Advisory The inspectors identified a finding of very low safety significance (Green) and an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to adequately implement procedure E-O-05, Response to Natural Events, during a high wind advisory. Specifically, on February 22, 2007, during the advisory, the inspectors identified several items stored outdoors near the plant main output transformer that could become missile hazards during actual high winds. As part of corrective actions, the licensee removed the items. The issue was entered into the licensees corrective action program.
The inspectors determined that the finding is greater than minor because, if left uncorrected, the loose items could become a more significant safety concern by allowing the accumulation of missile hazards in these areas, thereby increasing the likelihood of an initiating event. The inspectors determined that the finding warranted evaluation using the Significance Determination Process (SDP) because the finding was associated with an increase in the likelihood of an initiating event.
The finding was determined to be of very low safety significance (Green). This finding has a cross-cutting aspect in the area of human performance because the licensee failed to communicate decisions and the basis for decisions to personnel who have a need to know the information in order to perform work safely and in a timely manner.
Inspection Report# : 2007002 (pdf)
Significance:      Dec 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Pre-Fire Strategy Identified in Cable Spreading Room A finding of very low safety significance and an associated non-cited violation of Technical Specification 6.8, Procedures, was identified by the inspectors for the failure to identify radiological and toxic hazards in the cable spreading area fire zone pre-fire strategy. These hazards were from a radioactively contaminated lead pipe in the fire zone that could melt during certain fire scenarios. As part of corrective actions, the licensee appropriately revised the strategy.
The issue was entered into the licensees corrective action program.
The finding is greater than minor because it was associated with the external factors - fire 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 provide adequate warnings and guidance in the pre-fire plan related to these hazards could have adversely impacted the fire brigades ability to properly respond to a fire. This impact could increase the likelihood of damage to equipment, causing an upset pf plant stability. NRC management review determined the finding to be of very low safety significance (Green),
due to the extensive training provided to fire brigade members to deal with unexpected contingencies. This finding has a cross-cutting aspect in the area of human performance because the licensee failed to provide complete, accurate, and up-to-date pre-fire strategies for the fire brigade to respond to a fire.
Inspection Report# : 2006005 (pdf)
Significance:      Jun 23, 2006 Identified By: NRC Item Type: NCV NonCited Violation Procedure for Reactor Startup Not Followed The inspectors identified a finding associated with a non-cited violation of Technical Specification 6.8.a (written procedures and administrative policies). The finding was for the licensees failure to follow approved procedures during a
 
plant startup. The finding was of very low safety significance and there were three examples of the finding. The first example of a failure to follow approved procedures occurred when operators incorrectly marked a procedure step as not applicable and failed to execute the step. The second example of the failure to follow approved procedures occurred when operators executed procedure steps out of sequence. The third example occurred during the previous reactor startup conducted in November 2005 when operators performed procedure steps out of sequence in the same manner as executed during this plant startup. Corrective actions included placing Procedure N-0-01 on administrative hold until appropriate procedure changes could be made and training operating crews on procedure adherence.
This finding was of more than minor safety significance. Failure to comply with reactivity management requirements can lead to an uncontrolled reactivity event. In this particular event, the failure to follow the procedural sequence could have resulted in shutdown margin being less than that required by Technical Specifications. However, this finding is of very low significance because the actual shutdown margin did not go below the minimum required by Technical Specifications. This finding affected the cross-cutting issue of human performance.
Inspection Report# : 2006011 (pdf)
Significance:        Jun 23, 2006 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedure for Reactor Startup The inspectors identified a finding associated with an non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, of very low safety significance associated with an event. The inspectors identified that Procedure N-0-01, Plant Startup from Cold Shutdown Condition to Hot Shutdown Condition, Revision BI, Step 4.45 was inadequate to start up the reactor for the conditions that existed on May 17, 2006. The procedure, as written, would have required the operators to dilute the reactor to a lower boron concentration than the Estimated Critical Position boron concentration prior to withdrawing the Shutdown Bank rods. Corrective actions to address this finding included placing Procedure N-0-01 on administrative hold until appropriate procedure changes could be implemented.
This finding was more than minor in safety significance because this issue, if left uncorrected, would have resulted in the core reactivity shutdown margin being less than that required by Technical Specifications. However, this finding is of very low significance because the procedure step was not executed and shutdown was never below that required by Technical Specifications. This finding affected the cross-cutting issue of human performance.
Inspection Report# : 2006011 (pdf)
Significance:        May 19, 2006 Identified By: NRC Item Type: NCV NonCited Violation Criterion XVI: Failed to Identify Causes and Corrective Actions to Preclude Repetition for Significant Conditions Adverse to Quality The NRC inspectors identified a finding of very low safety significance that involved a violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions. Specifically, for the turbine building flooding and auxiliary feedwater air entrainment performance deficiencies, which were significant conditions adverse to quality, the licensee failed to identify the causes, and to determine corrective actions to preclude repetition.
The finding was greater than minor because the failure to identify the cause and corrective actions to preclude repetition of significant conditions adverse to quality, which led to a degraded cornerstone could result in the NRC needing to take more significant action. The finding was determined to be of very low safety significance based on management review, and the determination that no additional instances of significant conditions adverse to quality have actually occurred due to the failure to identify the causes and corrective actions for the previous performance deficiencies. The cause of the finding was related to the evaluation aspect of the cross-cutting element of problem identification and resolution.
Inspection Report# : 2006007 (pdf)
Significance:        May 05, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Incorporate Operating Experience Into Preventive Maintenance Procedures
 
The inspectors identified a finding associated with a non-cited violation (NCV) of 10 CFR 50.65 (the Maintenance Rule),
having very low safety significance for the licensees failure to incorporate into station procedures available internal and external operating experience pertaining to 4.16-kilovolt (kV) switchgear mechanically operated contact (MOC) switch linkage assemblies. As a result, preventive maintenance procedures for 4.16-kV safety- and nonsafety-related switchgear breaker cubicles were inadequate and had not been upgraded to incorporate important MOC switch linkage measurements and adjustments to be used during periodic breaker/cubicle maintenance. The licensee entered the problem with the procedures into its corrective action program for resolution. Corrective action included the revision of the procedures to incorporate the need to inspect the linkage and adjust it to within specified values.
The finding is greater than minor because it is associated with the procedure adequacy 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 operation. The finding was determined to be of very low safety significance because the transient initiator contributor is a reactor trip that did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available. The cause of the finding is related to the cross-cutting element of problem identification and resolution.
Inspection Report# : 2006010 (pdf)
Mitigating Systems Significance:        Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inappropriate QA Class Components Installed in TSC Diesel Generator The inspectors identified a finding of very low safety significance (Green) and an associated Non-Cited Violation of Technical Specification 6.8, Procedures, during a review on January 27, 2007, of maintenance performed on the station blackout diesel generator. The maintenance, which was conducted to repair a cooling water leak, inappropriately replaced existing parts with commercial grade components. The inspectors determined that, in accordance with procedure GNP-01.01.01, Determination of Nuclear Safety Designed Classifications, QA [Quality Assurance] Type and EQ
[Environmental Qualification] Type, the new components should have been designated as augmented quality. As part of corrective actions, the licensee revised its parts database to show the appropriate classification for parts for the diesel. The issue was entered into the corrective action program.
The inspectors determined that the finding is greater than minor 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). Specifically, the installation of parts in equipment with a lower quality designation than required potentially impacted equipment reliability. The finding was determined to be of very low safety significance (Green). This finding has a cross-cutting aspect in the area of problem identification and resolution because the licensee failed to take timely effective corrective actions for a similar prior occurrence. Barriers to prevent recurrence had not been established during supervisory reviews that granted multiple extensions to the corrective actions for the prior occurrence.
Inspection Report# : 2007002 (pdf)
Significance:        Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Nuclear Instrument Test Performed Contrary to Procedural Requirements The inspectors identified a finding of very low safety significance (Green) and an associated Non-Cited Violation of Technical Specification 6.8, Procedures, when the licensee failed, on January 8, 2007, to follow procedures for performing the monthly surveillance test on power range instrument N-42 and failed to obtain an approved procedure change as required by administrative procedures when the technicians established an alternate ground point contrary to procedural requirements. As part of corrective actions, the licensee counseled the technicians involved and discussed the event with all members of the instrument and control department. The issue was entered into the corrective action program.
 
The inspectors determined that the finding is greater than minor because it was associated with the procedure quality 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).
Specifically, the procedure required the use of the ground associated with the related card to verify proper continuity within the circuit and the use of an alternate ground point was a substantive change to the procedure. The finding was determined to be of very low safety significance (Green). This finding has a cross-cutting aspect in the area of human performance because personnel did not follow procedures.
Inspection Report# : 2007002 (pdf)
Significance:        Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Seismic Storage requirements in Station Housekeeping Procedure The inspectors identified a finding of very low safety significance (Green) and an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, during a review of procedures related to the control and storage of material. On March 21, 2007, the inspectors identified a number of unsecured equipment carts located in the vicinity of the seismically-classified, safety-related auxiliary building special ventilation system. The inspectors concluded that, although this was allowed by plant procedure GNP-01.31.01, Plant Cleanliness and Storage, it was a condition that potentially affected quality (safe operation of the ventilation system during a seismic event) and should not have been allowed by the procedure. As part of corrective action, the licensee properly secured the carts and evaluated other carts positioned near safety-related equipment. The issue was entered into the corrective action program.
The inspectors determined that the finding is greater than minor because it was associated with the procedure quality 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).
Specifically, the procedure allowed uncontrolled storage of materials in the vicinity of the auxiliary building special ventilation system that could render the system inoperable during a seismic event. The finding was determined to be of very low safety significance (Green). This finding has a cross-cutting aspect in the area of human performance because the licensee failed to provide accurate procedures to assure the operability of safety-related equipment was maintained.
Inspection Report# : 2007002 (pdf)
Significance:        Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Evaluate the Potentially Inadequate Design of the Service Water System The inspectors identified a finding of very low safety significance (Green) and associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, Instruction, Procedures, and Drawings, when the licensee failed to have in place adequate procedures to preclude a common mode failure of both trains of the safety-related service water (SW) system. Specifically, adequate procedures were not established for the maintenance of the SW system to prevent corrosion and degradation of the plant equipment water (PEW) filter vessels from affecting the safety-related SW bearing water supply components. As part of corrective actions, the licensee wrote the appropriate maintenance procedures. The issue was entered into the corrective action program.
The inspectors concluded that this finding is greater than minor because it was associated with the procedure quality and equipment performance attributes 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.
Specifically, the lack of appropriate procedures allowed the degradation of PEW components to cause the inoperability of two safety-related SW pumps. The finding was determined to be of very low safety significance. This finding has a cross-cutting aspect in the area of problem identification and resolution because the licensee failed to thoroughly evaluate problems such that the resolutions addressed causes and extent of condition, as necessary.
Inspection Report# : 2007002 (pdf)
Significance:        Mar 31, 2007 Identified By: NRC
 
Item Type: NCV NonCited Violation Foreign Material in Containment as a Result of Inadequate Containment Closure Inspections The inspectors identified a finding of very low safety significance (Green) and an associated Non-Cited Violation of Technical Specification 6.8, Procedures, on February 28, 2007, when the licensee failed to adequately perform a containment closeout inspection to ensure that debris and foreign materials were identified and removed in accordance with plant procedures. Specifically, inspectors identified unsecured metal sheets inside containment during a walkdown. As part of corrective actions, the sheets were removed from containment. The issue was entered into the corrective action program.
This finding is greater 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, the failure to identify and remove the steel sheets from containment could have affected the availability of both trains of the residual heat removal system (the accident recirculation sump) during a loss-of-coolant accident because of increased debris generation caused by the unsecured sheets. The finding was determined to be of very low safety significance (Green). This finding has a cross-cutting aspect in the area of human performance because personnel did not follow procedures, causing a condition to exist that potentially impacted the operability of both trains of the residual heat removal system.
Inspection Report# : 2007002 (pdf)
Significance: SL-IV Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Notify NRC of Licensee medical Condition Change in Accordance with 10 CFR50.74 The inspectors identified a finding of very low safety significance and an associated Severity Level IV, Non-Cited Violation of 10 CFR 50.74 for the licensees failure to notify the NRC that one of its licensed operators was taking prescribed medication for a potentially disqualifying medical condition (hypertension). After a review of the licensed operators medical status was completed by the NRCs medical review officer, a condition was added to the operators license requiring him to take the medication as prescribed. The facility licensee entered this issue in their corrective action program. They required the individual licensed operator to take the medication as prescribed and incorporated these lessons learned in their requalification training program to ensure all licensed operators are aware of the requirement to notify the NRC of changes in their medical status.
Because violations of 10 CFR 50.74 affect the NRCs ability to perform its regulatory function, this finding was evaluated using the traditional enforcement process. In accordance with the NRC Enforcement Policy, this finding was determined to be greater than minor because the medical condition that was not reported required a change to the operators NRC license.
Because the operator was always in the presence of other licensed operators while performing licensed duties and made no operational errors while he was taking the prescribed medication before his license had been appropriately revised, NRC management has determined this issue is a Green finding, of very low safety significance. This issue is considered an NCV because it was entered into the licensees corrective action program. This finding also has a cross-cutting aspect in the area of human performance because a standard, specifically American National Standards Institute/American Nuclear Society (ANSI/ANS) 3.4, Medical Certification and Monitoring of Personnel Requiring Operator Licenses for Nuclear Power Plants, was available but not correctly implemented. The correct implementation of the standard would have led to a proper notification of the NRC and timely conditioning of the operators NRC license.
Inspection Report# : 2007002 (pdf)
Significance:      Jan 31, 2007 Identified By: NRC Item Type: VIO Violation Failure to Evaluate Operability of the "A" EDG when a Fuel Oil Leak was Identified A finding that was preliminarily determined to be of substantial safety significance (Yellow), and an associated apparent violation of Technical Specification 6.8, Procedures, was identified for a fuel oil leak on the A emergency diesel generator (EDG) that was identified on June 28, 2006, but was not repaired until 51 days later on August 17. In December 2006, the licensee tested the fitting and copper tubing that was the source of the leak to assess the leaks effect on the operability of the diesel. The licensee concluded that the leak rendered the diesel inoperable for those 51 days. As part of corrective action, the licensee replaced the leaking fuel oil line and reinforced with plant personnel the procedural requirements to properly evaluate equipment problems. The licensee also entered the issue into its corrective action program.
 
The finding was more than minor because if left uncorrected it would become a more significant safety concern during use of the A EDG to mitigate a loss of offsite power event. Specifically, the A EDG would have failed after approximately four hours due to the loss of fuel oil through the failed fuel line tubing, and the systems that respond to accidents and are powered by the A EDG would not be available. A Significance Determination Process Phase 3 risk analysis preliminarily determined that this finding was of substantial safety significance (Yellow). This finding has a cross-cutting aspect in the area of human performance because procedures were available, but not followed, that could have resulted in the leak being promptly repaired.
After considering the information developed during the inspection, the NRC has concluded that the inspection finding is appropriately characterized as Yellow, i.e., an issue with substantial safety significance that will result in additional NRC inspection and potentially other NRC action.
Inspection Report# : 2007007 (pdf)
Inspection Report# : 2007009 (pdf)
Significance:      Dec 14, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify Emergency Diesel Generator Air Intake Temperature Limitations Impact Upon Ability to Meet Technical Specification Surveillance Requirements A finding of very low safety significance was identified by the inspectors for a violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action. The licensee failed to identify the impact of air intake temperature limitation on the ability of the emergency diesel generators to meet Technical Specification surveillance loading requirements at elevated temperatures. Once identified, the licensee established 75 degrees Fahrenheit as a maximum outside temperature for emergency diesel generator operability. The primary cause of this violation was related to the cross-cutting area of Problem Identification and Resolution, because the licensee failed to ensure that an issue potentially impacting nuclear safety was promptly identified, fully evaluated, and that actions were taken to address safety issues in a timely manner, commensurate with their significance.
The issue was more than minor because the failure to identify that the emergency diesel generators would not be able to meet Technical Specification surveillance requirements at elevated temperatures could have resulted in the emergency diesel generators being considered operable when, in fact, they had less operational margin than required by Technical Specifications. The issue was of very low safety significance because both of the emergency diesel generators were determined to be capable of carrying their respective design basis accident loads below the outside temperature limitations that the licensee had in place. The issue was a NCV of 10 CFR Part 50, Appendix B, Criterion XVI, which required that conditions adverse to quality are promptly identified and corrected.
Inspection Report# : 2006016 (pdf)
Significance:      Dec 14, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Provide Suppression for Safe Shutdown Equipment in Appendix R, III.G.3 Area A finding of very low safety significance was identified by the inspectors for a violation of 10 CFR Part 50, Appendix R, Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979. The licensee failed to provide required fire suppression coverage in fire zone AX-32 for the safe shutdown functions of source range monitoring, isolation of a steam generator blowdown line, and pressurizer level instrumentation. Once identified, the licensee entered the issue into their corrective action program and implemented compensatory measures.
This issue was more than minor because the failure to provide suppression for redundant trains of safe shutdown equipment increased the likelihood that alternative shutdown methods would have to be used in the event of a fire. The issue was of very low safety significance because of the mitigating systems, which would have remained available in the event of a fire.
The issue was a NCV of 10 CFR Part 50, Appendix R, Section III.G.3, which required fixed suppression systems for alternative shutdown areas such as fire zone AX-32.
Inspection Report# : 2006016 (pdf)
 
Significance:        Oct 23, 2006 Identified By: NRC Item Type: NCV NonCited Violation Scaffolding in Contact with the Safety Injection System Affects Operability A finding of very low safety significance and an associated non-cited violation of Technical Specification 6.8, Procedures, was identified by the inspectors on October 23, 2006, for the failure to install scaffolding in accordance with station procedures. Specifically, scaffolding was installed inside containment that was too close to, or was in contact with, safety injection system components and piping. As part of corrective actions, the licensee removed the scaffolding and enhance the station procedure for scaffolding. The issue was entered into the licensees corrective action program.
This finding is greater than minor because it was 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). Specifically, improperly positioned scaffolding could have impeded or prevented proper operation of the safety injection system during an accident.
The finding was of very low safety significance (Green) because it did not require a quantitative assessment. This finding has a cross-cutting aspect in the area of human performances because personnel did not follow the procedure for scaffolding.
Inspection Report# : 2006005 (pdf)
Significance:        Oct 11, 2006 Identified By: NRC Item Type: NCV NonCited Violation Inadvertent Drain Down of the Reactor Coolant system During Fill and Vent of the Containment Spray System A finding of very low safety significance and an associated non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed during performance of a plant safety-related procedure to fill and vent the containment spray system, on October 11, 2006, when water was inappropriately diverted from the reactor coolant system to the residual heat removal system. As part of corrective actions, the licensee revised the procedure to ensure the systems were properly aligned during fill and vent activities. The issue was entered into the licensees corrective action program.
This finding is greater than minor because if left uncorrected it would become a more significant safety concern in that the use of other inadequate procedures could have rendered inoperable important mitigating equipment, such as the containment spray and residual heat removal systems. Additionally, the finding was associated with the procedure quality and configuration control attributes 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 was determined to be of very low safety significance (Green) because it did not require a quantitative assessment. This finding has a cross-cutting aspect in the area of human performances because the licensee failed to provide complete, accurate, and up-to-date procedures to fill and vent the containment spray system.
Inspection Report# : 2006005 (pdf)
Significance:        Sep 30, 2006 Identified By: NRC Item Type: FIN Finding Technical Specification LCO Not Entered for Diesel Generators Inoperable While in Refueling Shutdown A finding of very low safety significance (Green) was identified by the inspectors when the licensee failed to properly apply shutdown Technical Specifications (TSs) for the residual heat removal (RHR) system with both emergency diesel generators (EDGs) declared inoperable. While reviewing startup preparations being made for a mode change, the inspectors identified that TSs required both RHR systems to be operable and that both EDGs were inoperable due to tornado failure susceptibilities, thereby rendering both trains of RHR inoperable as required by the related power requirements TS. The licensee concurred with the inspectors observations, prevented the mode change, and issued the related licensee event report. Corrective actions, to date, included restoration of EDG operability prior to making a mode change and procedural enhancements.
The inspectors determined that the finding is greater than minor because if left uncorrected it would become a more significant safety issue: the licensee would have made a mode change without the required operable equipment. This
 
finding was of very low safety significance because the licensee returned the EDGs to operability prior to making any mode changes, no violation of NRC requirements was identified, and the finding did not require a quantitative assessment using Check List 4 for PWR Shutdown Operation with Time to Boil >2 hours and Inventory in the Pressurizer. The cause of this finding was related to the crosscutting area of human performance because procedures, specifically TSs, were available but not followed, that would have facilitated the proper performance of the task.
Inspection Report# : 2006004 (pdf)
Significance:      Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Reactor Protection System Surveillance Procedure Revised Without Proper Review The inspectors identified a finding of very low safety significance and an associated non-cited violation of Technical Specification 6.8, Procedures, during a review of a procedure. The licensee had changed the procedure to allow the turbine-driven auxiliary feedwater (TDAFW) pump to be considered available for risk management purposes while the pump control switch was in pull-to-lock during the performance of the surveillance procedure; however, the required Plant Operating Review Committee review and approval for the change was not obtained. Corrective actions, to date, included review of the surveillance procedure by the Plant Operating Review Committee and inclusion into the procedure of additional provisions to ensure availability of the TDAFW pump while the control switch is in pull-to-lock during performance of the procedure. The cause of this finding is related to the cross-cutting area of human performance because of the licensees failure to follow a plant procedure regarding the review and approval of safety-related procedures.
The finding is greater than minor because if left uncorrected the finding would become a more significant safety concern.
Specifically, improper application of the temporary procedure change process could lead to a more significant unreviewed, improper procedure change. Additionally, this issue is associated with the procedure quality 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). Specifically, the failure to provide adequate review and approval of a safety-related surviellance procedure prior to issuance for use and the failure to include adequate provisions to ensure availability of a safety-related component in the surveillance procedure potentially impacted equipment availability. The finding is of very low safety significance because the answer to all the screening questions in the significance determination process Phase 1 screening worksheet in the Mitigating Systems column was no.
Inspection Report# : 2006003 (pdf)
Significance:      Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Leak Developed in Service Water Pipe after Wall Thinning Evaluation was Cancelled A self-revealed finding of very low safety significance and an associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, were identified on April 25, 2006, when a leak due to pipe-wall thinning was identified in a 90&deg; elbow in a service water (SW) line to the B emergency diesel generator. This wall-thinning and leak, a condition adverse to quality, resulted in the need to declare the emergency diesel generator inoperable and a shut down of the reactor to allow repair of the leak. In April 2004, a work order to inspect the elbow for wall-thinning was cancelled after wall thickness in a nearby elbow was evaluated by the licensee and deemed acceptable. The extrapolation of inspection results from one elbow to the other elbow was inappropriate. Corrective actions taken by the licensee included replacement of the failed section of SW piping, performance of additional inspections on SW piping, and replacement of other safety-related sections of SW piping. The cause of this finding is related to the cross-cutting area of problem identification and resolution because the licensee failed to promptly identify an issue potentially impacting safety-related piping.
The finding is greater than minor because it is associated with the equipment performance attribute of the Mitigating System 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). Specifically, the failure to conduct a wall-thinning evaluation in April 2004 resulted in the need to take the emergency diesel generator out-of-service and shut down the reactor to allow repair of the pipe. Additionally, the failure to inspect and correct, as necessary, wall-thinning in a safety-related system, if left uncorrected, would become a more significant safety concern through the possible development of a large system leak or the complication of the operations of a safety-related system. The finding is of very low safety significance because the answer to all the screening questions in the significance determination process Phase 1
 
screening worksheet in the Mitigating Systems column was no.
Inspection Report# : 2006003 (pdf)
Significance:        May 19, 2006 Identified By: NRC Item Type: NCV NonCited Violation Criterion V: Failed to Incorporate Appropriate Acceptance Criteria for Assessing Operability of the AFW Pump The NRC inspectors identified a finding of very low safety significance that involved a violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings. Specifically, the licensee failed to incorporate appropriate acceptance criteria for assessing operability of the auxiliary feedwater pump following identification of a piping obstruction.
The finding was greater than minor because the finding was associated with the Mitigating Systems cornerstone attribute of procedure quality which affected the cornerstone objective. Specifically, the relevant procedure was not adequate to ensure the availability, reliability, and capability of the auxiliary feedwater system to respond to initiating events. The finding was determined to be of very low safety significance because subsequent evaluation of the pipe occlusions, using appropriate acceptance criteria, supported past operability of the pump. The cause of the finding was related to the evaluation aspect of the cross-cutting element of problem identification and resolution.
Inspection Report# : 2006007 (pdf)
Significance:        May 19, 2006 Identified By: NRC Item Type: NCV NonCited Violation Criterion III: Failed to Correctly Translate Containment Sump Volume into Design The NRC inspectors identified a finding of very low safety significance that involved a violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to ensure that design basis calculations correctly translated the containment sump volume at the time of the switch over from the refueling water storage tank to the containment sump to ensure adequate available net positive suction head and vortex suppression for the residual heat removal pumps.
The finding was greater than minor because the finding was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective because the inadequate calculation impacted the design requirements for the new containment strainers being installed to resolve Generic Safety Issue 191. The finding was determined to be of very low safety significance because (1) the licensee normally kept the refueling water storage tank at a level above the Technical Specification minimum; (2) new strainers were not yet installed; and (3) inspector-independent calculations indicated that the pumps had adequate net positive suction head and vortex suppression, with the additional non-conservatisms incorporated. The cause of the finding was related to the corrective action aspect of the cross-cutting element of problem identification and resolution.
Inspection Report# : 2006007 (pdf)
Significance:        May 19, 2006 Identified By: NRC Item Type: NCV NonCited Violation Criterion III: Failed to Verify or Check the Adequacy of the Design Canceling Design Change Request 2548 The NRC inspectors identified a finding of very low safety significance that involved a violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to properly evaluate the minimum flow requirements of the high head safety injection pumps.
The finding was greater than minor because the finding was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective as providing inadequate minimum flow to the SI pumps could result in the pumps failing under certain accident scenarios. The finding was determined to be of very low safety significance because both the licensee and the inspectors determined that the safety injection pumps remained operable with the 47 gpm minimum flow rate. The cause of the finding was related to the corrective action of the cross-cutting element of problem identification and resolution.
Inspection Report# : 2006007 (pdf)
 
Significance:        May 05, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Maintain Separation of Cables The inspectors identified a finding associated with a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, that pertained to a modification that failed to incorporate applicable design requirements for cable separation. Nonsafety-related cables associated with train B reactor coolant pump (RCP) safety-related cable trays and cables were bundled inside the RCP breaker cubicles with train A RCP safety-related cables feeding the reactor protection system (RPS). Consequently, a fault in the train B cable/cable tray could propagate to train A. The licensee entered the problem into its corrective action program for resolution. Corrective actions included encasing the nonsafety-related cables in flexible metal conduit and confirming that other safety-related cables were not affected.
The finding is greater than minor 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). The finding was determined to be of very low safety significance because of the redundancy and coincident logic in the RPS design; and it did not represent a loss of system safety function, an actual loss of safety function of a single train, an actual loss of safety function of one or more non-technical specification trains of equipment, designated as risk significant per 10 CFR 50.65, for greater than 24 hours, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event.
Inspection Report# : 2006010 (pdf)
Barrier Integrity Significance: SL-IV Dec 14, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Fully Update Updated Safety Analysis Report A finding of very low safety significance was identified for the licensees failure to adequately update the Update Safety Analysis Report (USAR) in accordance to 10 CFR 50.71, Maintenance of Records, Making of Reports. The licensee failed to update the USAR to fully reflect changes and analyses made in response to NRC Generic Letter (GL) 96-06, Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions. Once identified, the licensee entered this issue into their corrective action program. The primary cause of this violation was related to the cross-cutting area of Human Performance because the licensee failed to ensure that personnel, equipment, procedures, and other resources were available and adequate to assure nuclear safety. Specifically, the licensee failed to provide adequate engineering procedural guidance concerning the required content of USAR updates.
Because this issue potentially impacted the NRCs ability to perform its regulatory function, this finding was evaluated using the traditional enforcement process. The finding was determined to be more than minor because of the failure to provide complete licensing and design basis information in the USAR could result in either the licensee making an inappropriate licensing interpretation or the NRC making an inappropriate regulatory decision based on incomplete information in the USAR. The issue was of very low safety significance because no instances were identified where the failure to appropriately update the USAR impeded or influenced a regulatory decision, or resulted in an actual loss of safety function. The issue was a NCV of 10 CFR 50.71(e) which required that the USAR be updated to include the effects of all analyses of new safety issues performed by or on behalf of the licensee at Commission request.
Inspection Report# : 2006016 (pdf)
Significance:        May 19, 2006 Identified By: NRC Item Type: NCV NonCited Violation Criterion III: Failed to Properly Translate the ICS Design Basis into the Technical Specifications The NRC inspectors identified a finding of very low safety significance that involved a violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to ensure that design basis calculations correctly translated the internal containment spray flow requirements into the Technical Specification allowed number of
 
blocked internal containment spray nozzles.
The finding was greater than minor because the containment spray system could have been inoperable with the allowable pump degradation and allowable number of blocked containment spray nozzles. The finding was determined to be of very low safety significance because the internal containment spray system was determined to be operable. The cause of the finding was related to the evaluation aspect of the cross-cutting element of problem identification and resolution.
Inspection Report# : 2006007 (pdf)
Emergency Preparedness Occupational Radiation Safety Public Radiation Safety Significance:      Jun 30, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Properly Calibrate the Waste Discharge Liquid and the Steam Generator Blowdown Radiation Monitors The inspectors identified a finding of very low safety significance and an associated violation of NRC requirements for the failure to comply with technical specification and Offsite Dose Calculation Manual (ODCM) requirements in the calibration of two liquid discharge radiation monitors listed in the ODCM. Specifically, the radiation monitor high alarm trip functions were not verified with radiation sources during instrument calibration.
The finding is greater than minor because it is associated with the plant facilities/equipment and instrumentation attribute 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 released into the public domain. Specifically, not verifying the proper operation of a radiation monitor at its high alarm trip setpoint could result in the use of a monitor that does not properly operate at the high alarm setpoint and the consequent unintended release of radioactive material to the environment in excess of regulatory limits. The finding is of very low safety significance because actual effluent discharges were adequately analyzed for radioactive content by the licensee prior to release, and the licensees ability to assess dose from radioactive waste (radwaste) liquid discharges was not impaired, nor were regulatory dose limits or As-Low-As-Is-Reasonably-Achievable dose constraints exceeded due to liquid effluent discharges.
Inspection Report# : 2006003 (pdf)
Physical Protection Physical Protection information not publicly available.
Miscellaneous Last modified : June 01, 2007
 
Kewaunee 2Q/2007 Plant Inspection Findings Initiating Events Significance:      Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation No Analysis or Procedures to Establish Operability of the TAT Source The inspectors identified a finding having very low significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to evaluate the capability of the 345 kV offsite power supply when isolated from the 138 kV switchyard and to translate this criteria into procedures.
This issue was more than minor because procedures allowed operation of the station in unanalyzed configurations for which operability of one offsite source could not be assured and new calculations were needed to ensure that the design basis was met. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
Significance:      Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Incorporate and Effectively Implement Operating Experience in RTB Maintenance Activities The inspectors identified a finding having very low significance and an associated NCV of 10 CFR 50.65(a)(3) for the failure to incorporate external and internal operating experience into preventive maintenance activities for the reactor trip breakers. 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 previous breaker issues and did not perform adequate extent of condition reviews. Specifically, the licensee initiated several corrective action documents in response to identified issues; however, did not perform adequate evaluations of the conditions to address the cause or resolve the identified issue. (P.1.(c))
This issue was more than minor because the licensee failed to ensure that the RTBs, and their associated cell assemblies, had been maintained in a continuous state of operational readiness by performing effective maintenance and surveillance activities in accordance with relevant vendor specifications and available operating experience. The issue was of very low safety significance based on a Phase 1 screening because the finding did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available.
Inspection Report# : 2007006 (pdf)
Significance:      Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Acceptance Criteria Not Met Due to Failure to Follow Procedure The inspectors identified a finding having very low significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings. Specifically, on May 22, 2006 during the performance of PMP-47-01, maintenance technician recorded a trip bar force of 32 ounces when testing RTB S/N 850-027-1, which exceeded the acceptance criteria; however, no further actions were taken as required by the test. This finding has a cross-cutting aspect in the area of Human Performance, Work Practices because the licensee did not perform an adequate peer check of the surveillance results. Specifically, several individuals including the person performing the task did not identify that the RTB trip bar force exceeded the acceptance criteria. (H.4.(c))
This issue was more than minor because not meeting the acceptance for the trip bar force impacted the reliability of
 
the RBTs because excessive force could result in a failure to trip the breaker. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
Significance:      Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Implement E-0-05, "Response to Natural Events," During a High Wind Advisory The inspectors identified a finding of very low safety significance (Green) and an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to adequately implement procedure E-O-05, Response to Natural Events, during a high wind advisory. Specifically, on February 22, 2007, during the advisory, the inspectors identified several items stored outdoors near the plant main output transformer that could become missile hazards during actual high winds. As part of corrective actions, the licensee removed the items. The issue was entered into the licensees corrective action program.
The inspectors determined that the finding is greater than minor because, if left uncorrected, the loose items could become a more significant safety concern by allowing the accumulation of missile hazards in these areas, thereby increasing the likelihood of an initiating event. The inspectors determined that the finding warranted evaluation using the Significance Determination Process (SDP) because the finding was associated with an increase in the likelihood of an initiating event. The finding was determined to be of very low safety significance (Green). This finding has a cross-cutting aspect in the area of human performance because the licensee failed to communicate decisions and the basis for decisions to personnel who have a need to know the information in order to perform work safely and in a timely manner.
Inspection Report# : 2007002 (pdf)
Significance:      Dec 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Pre-Fire Strategy Identified in Cable Spreading Room A finding of very low safety significance and an associated non-cited violation of Technical Specification 6.8, Procedures, was identified by the inspectors for the failure to identify radiological and toxic hazards in the cable spreading area fire zone pre-fire strategy. These hazards were from a radioactively contaminated lead pipe in the fire zone that could melt during certain fire scenarios. As part of corrective actions, the licensee appropriately revised the strategy. The issue was entered into the licensees corrective action program.
The finding is greater than minor because it was associated with the external factors - fire 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 provide adequate warnings and guidance in the pre-fire plan related to these hazards could have adversely impacted the fire brigades ability to properly respond to a fire. This impact could increase the likelihood of damage to equipment, causing an upset pf plant stability. NRC management review determined the finding to be of very low safety significance (Green), due to the extensive training provided to fire brigade members to deal with unexpected contingencies. This finding has a cross-cutting aspect in the area of human performance because the licensee failed to provide complete, accurate, and up-to-date pre-fire strategies for the fire brigade to respond to a fire.
Inspection Report# : 2006005 (pdf)
Mitigating Systems Significance: SL-IV Jun 30, 2007 Identified By: NRC Item Type: VIO Violation Failure to Perform a 10 CFR 50.59 Evaluation for Compensatory Measures Associated with a Procedure Change
 
The inspectors identified a finding having very low safety significance and an associated Severity Level IV, Cited Violation of 10 CFR 50.59 while reviewing unresolved items URI 05000305/2006003-04, Adequacy of Compensatory Actions for Potential Turbine Missile Strike of Control Room Ventilation Cooling; and URI 05000305/2006016-01, Adequacy of 10 CFR 50.59 Screening for Procedure Change. Specifically, the licensee failed to properly interpret design and licensing basis requirements associated with protection against external events and as a result did not perform a 10 CFR 50.59 evaluation. The cause of this finding is related to the cross-cutting area of problem identification and resolution because the licensee had similar prior problems that, if effectively evaluated and resolved, could have prevented this issue. (P.1(c))
This finding was determined to be more than minor because the inspectors determined that the procedure change would have ultimately required NRC approval. The procedure changes, in the form of compensatory operator actions, adversely impacted the operation of control room recirculation system following a tornado. A Phase 1 significance determination of this finding using IMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," using the Severe Weather Screening Criteria questions was completed. Since the loss of the control room recirculation system would not result in an initiating event or degrade two or more trains of a multi-train safety system, the issue screened as Green.
Inspection Report# : 2007003 (pdf)
Significance:      May 18, 2007 Identified By: NRC Item Type: NCV NonCited Violation Procedure Non-Compliance The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to adequately implement procedure DNAP-1604, Cause Evaluation Program, and the Cause Evaluation Handbook during investigative analyses of root cause, collective significance, and apparent cause evaluations. The licensee subsequently revised several apparent cause evaluations (ACEs), such as ACE 3374 on the diesel generator B fuel rack shaft binding, and completed industry benchmarking to improve root cause evaluation and ACE investigative analysis.
This finding was associated with the Mitigating Systems Cornerstone. The finding was more than minor because, if left uncorrected, the licensees analyses of conditions adverse to quality, such as the investigation of the diesel generator B fuel rack shaft binding, as documented in ACE 3374, would not be performed at an appropriate investigative depth for cause determination. The inspectors assessed the significance of this finding as very low safety significance (Green) because the finding did not represent an actual loss of safety function of the equipment. The finding was associated with cross-cutting aspect P.1(c), in the area of problem identification and resolution, corrective action program, because the licensee failed to thoroughly analyze the sequence of events and the cause and effect relationships potentially impacting the causal determination of CAP evaluations.
Inspection Report# : 2007008 (pdf)
Significance:      Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation No Motor Starting Analyses for Offsite Power Supply The inspectors identified a finding having very low significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to perform motor starting studies to demonstrate that motors would successfully start when connected to the offsite power supply. Upon discovery, the licensee provided additional data and compensatory measures to justify operability.
The inspectors determined that the performance deficiency was more than minor because the lack of a formal motor starting calculations resulted in the adequacy of important aspects of the design not being demonstrated, such that further evaluation needed to be performed in order to demonstrate that the equipment could perform its safety function. Although, by the end of the inspection, the licensee was able to demonstrate operability, at the time of discovery there was reasonable doubt on the operability of motors. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
 
Significance:      Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Increased Cable Resistance Due to Accident Temperatures The inspectors identified a finding having very low significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to consider the effects of accident temperatures on cable resistance in voltage drop calculations. Upon discovery, the licensee performed preliminary calculations to verify operability of the circuits.
This issue was more than minor because the calculational errors had more than a minimal effect on the outcome of the calculation, considerably impacting the available margin of the system such that further evaluation needed to be performed in order to demonstrate that the equipment could perform its safety function. Although, by the end of the inspection, the licensee was able to demonstrate operability; at the time of discovery there was reasonable doubt on the operability of the circuits. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
Significance:      Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Ensure Adequate 125 Vdc Breaker Interrupting Short Circuit Current Capability The inspectors identified a finding having very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to ensure that four of the 125 VDC circuit breakers had adequate interrupting short circuit fault current capability. Upon discovery, the licensee performed a preliminary evaluation, and verified that the most likely fault would result in a lower short circuit fault current than the breakers rating.
This issue was more than minor because the failure could have affected the operability of the breaker/DC Bus and could have resulted in the loss of DC power to safe shutdown equipment in the event of short circuit faults. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
Significance:      Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Use Actual Minimum Voltage Value in 125Vdc Voltage Drop Calculation The inspectors identified a finding having very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to use correct design input data into the 125 VDC safeguard battery calculation. The licensee used a battery terminal voltage value of 117.49 volts for BRA-101 and 118.95 volts for BRB-101, for the first minute, and did not compensate for worse case conditions. Upon discovery, the licensee performed preliminary evaluation and verified that safe shutdown equipment have adequate voltage using the battery terminal voltage value of 113.87 volts.
This issue was more than minor because the failure to use correct design input had more than a minimal effect on the outcome of the voltage drop calculation, considerably impacting the available margin of the system such that further evaluation needed to be performed in order to demonstrate that equipment could perform its safety function.
Although, during the inspection, the licensee was able to demonstrate operability; at the time of discovery there was reasonable doubt on the operability of circuits. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
 
Significance:      Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Acceptance Criteria in 125 Vdc Station Battery Load Tests Procedures The inspectors identified a finding having very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings. Specifically, the licensee failed to include the acceptable minimum battery terminal voltage, during the first minute, into the acceptance criteria for battery load test procedures SP-38-102A/B Station Battery Load Test. Upon discovery, the licensee entered the issue into its corrective action program to revise the acceptance criteria of procedures SP-38-102A/B to include this requirement.
This issue was more than minor because the failure to ensure that the battery terminal voltage during the first minute battery discharge did not drop below the design input value could have affected the operability of safety related equipments in the event of a design basis accident and or station blackout conditions. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
Significance:      Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Adequate Control Voltage for 4160V Breaker's Closing Coil was not Assured The inspectors identified a finding having very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to assure that the minimum available control voltage at the 4160V breakers was adequate to energize the closing coils during all conditions. Upon discovery, the licensee performed preliminary calculation and verified operability of the emergency diesel generators 4160V breakers following loss of all AC power conditions.
This finding was more than minor because the failure to assure adequate control voltage was available to close the 4160V breakers would have affected the capability of emergency diesel generators and other safety related equipments to respond to initiating events. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
Significance:      Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Safeguard Battery Load Profile Did Not Include LOOP/LOCA Loads The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control having very low safety significance for the licensees failure to assure that safeguard battery loads profile was adequate to meet all USAR requirements. Specifically, the licensee failed to verify that the battery loading profile for loss of coolant accident (LOCA) coincide with loss of all AC power condition was bounded by the station blackout condition loading to ensure adequate battery sizing and testing. Upon discovery, the licensee was able to show that the charger will be available upon the start of the emergency diesel generator and will provide additional support. This issue was entered into the licensees corrective action program to revise the battery calculation to include the LOCA loads.
This finding was more than minor because the failure to include the LOCA loads in the battery sizing and testing did not ensure the capability of the battery to provide adequate DC power in accordance with USAR requirements. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
Significance:      Apr 17, 2007 Identified By: NRC
 
Item Type: NCV NonCited Violation Electrolytic Capacitors in Spare Safeguard Battery Charger Not Periodically Energized The inspectors identified a finding having very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action. Specifically, the licensee failed to incorporate previously identified vendor recommendation to periodically energize the spare 125 VDC safeguard battery charger for at least a half-hour every 18 months to ensure the operability of the electrolytic capacitor in the charger. The licensee has previously entered the vendor recommendation into their corrective action in 2002, however, all actions were closed but the recommendation was never implemented. Following discovery, the licensee entered the issue into its corrective action program and declared the spare charger inoperable. The primary cause of this violation was related to the cross-cutting area of problem identification and resolution because the licensee failed to take appropriate corrective actions to address a previously failed charger. Specifically, the licensee developed corrective actions which included incorporating pertinent vendor recommendation into the preventive maintenance program but closed the action without ensuring completion (P.1.d)
This issue was more than minor because the failure to periodically energize the spare charger did not ensure the operability and reliability of the spare charger when needed. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
Significance:      Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Diesel Loading Calculations Non Conservative The inspectors identified a finding having very low significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to properly account for all loads on the diesel generators. Upon discovery, the licensee provided additional data and initiated procedure changes to ensure diesels were loaded within their ratings.
The inspectors determined that the performance deficiency was more than minor because the lack of adequate diesel generator loading calculations resulted in some diesel loads not being properly accounted for, such that further evaluation needed to be performed in order to demonstrate that the equipment could perform its safety function.
Although, by the end of the inspection, the licensee was able to demonstrate operability, at the time of discovery there was reasonable doubt on the operability of equipment. The inspectors screened the finding using IMC 0609, Appendix A. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
Significance:      Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation RWST Level Instruments Do Not Protect SI and RHR Pumps from Excessive Air Entrainment The inspectors identified a finding having very low significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to incorporate the results of design calculations with respect to minimum refueling water storage tank (RWST) level and transfer of suction sources into the appropriate emergency operating. Procedures allowed operators to transfer suction at 4 percent indicated level in the RWST; however, at this level, significant air entrainment may damage the pumps. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program because the licensee did not thoroughly evaluate problems such that the resolution addresses the extent of condition (P.1.c).
This issue was more than minor because the existing margin was already low and as a consequence, the large error associated with the level instrument resulted in eliminating the entire margin, and jeopardized the functionality of the pumps taking suction from the RWST due to excessive air entrainment. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations SDP Phase 1.
Inspection Report# : 2007006 (pdf)
 
Significance:        Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Non-Conservative Assumption Used in Service Water Flow Model Calculation The inspectors identified a finding having very low significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to appropriately account for service water strainer plugging in the service water system flow model. Upon discovery, the licensee placed this issue into their corrective action program and planned to formally revise the service water system flow model to reflect plugging of both strainers in a train.
The issue was more than minor because the error had more than a minimal effect on the outcome of the calculation, considerably impacting the available margin of the system such that further evaluation needed to be performed in order to demonstrate that the service water system could perform its safety function. The issue was of very low safety significance because the issue was a design issue confirmed to not result in a loss of operability.
Inspection Report# : 2007006 (pdf)
Significance:        Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Screen House Ventilation Damper Maintenance The inspectors identified a finding having very low significance and an associated NCV of 10 CFR Part 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, Paragraph (b)(2), for the licensees failure to scope the closing function of the screenhouse ventilation dampers into the monitoring program.
Specifically, the degraded screen-house dampers fail to close and maintain ambient temperatures > 60 &deg;F such that service water system would remain operable and available after a station blackout event with severely cold outside temperatures. Following discovery, the licensee entered the issue into its corrective action program for resolution.
This issue was more than minor because the licensee had not included the closing function of the screen-house ventilation dampers within the scope of its program for implementation of the Maintenance Rule. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
Significance:        Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Non Conservative Assumption Used for "B" CCW Pump Room Heat Gain Calculation The inspectors identified a finding having very low significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III. Specifically, the licensee failed to account for component cooling water (CCW) piping temperatures as high as 176&deg;F in the CCW B pump room and the impact upon the temperature in the CCW B pump room. As a result, the licensee used the non-conservative results in an operability evaluation for the auxiliary building fan coil unit (FCU). Upon discovery, the licensee placed this issue into their corrective action program, performed an immediate operability evaluation, and planned to perform a more thorough evaluation. This finding has a cross-cutting aspect in the area of human performance associated with decision making because the licensee did not use conservative assumptions. Specifically, the licensee failed to account for higher CCW piping temperatures because the licensee did not model the CCW room properly and did not use the maximum expected temperature under accident conditions when revising calculation C11156 (H.1.b).
The issue was more than minor because the error because, if left uncorrected, the finding would become a more safety significant concern. The use of a non-conservative value as a basis for operability could allow FCU performance to degrade to unacceptable levels without being detected and corrected. The issue was of very low safety significance because the issue was a design issue confirmed to not result in a loss of operability.
Inspection Report# : 2007006 (pdf)
 
Significance:      Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Safety Injection Pump Lube Oil Coolers Testing deficiencies The inspectors identified a finding having very low significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control. Specifically, the licensee failed to establish a testing program capable of identifying an unacceptable condition of the safety injection (SI) lube oil coolers. Upon discovery, the licensee initiated a change to the test program methodology and performed back-flushing and inspection on the two SI lube oil coolers. The licensee also assessed that as a result of the very cold temperature of the water of Lake Michigan during the inspection, the cooler was considered operable. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with self- and independent assessments because during a 2005 audit of licensing commitments, the licensee failed to identify that the commitment to perform inspection and maintenance of the SI lube oil coolers in accordance with the licensee's response to Generic Letter 89-13 was not kept (P.3.a).
This issue was more than minor because when later assessed, the licensee realized that the coolers would have failed previous tests when reevaluated performance factors were less than the acceptance criterion of 0.9. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
Significance:      Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inappropriate QA Class Components Installed in TSC Diesel Generator The inspectors identified a finding of very low safety significance (Green) and an associated Non-Cited Violation of Technical Specification 6.8, Procedures, during a review on January 27, 2007, of maintenance performed on the station blackout diesel generator. The maintenance, which was conducted to repair a cooling water leak, inappropriately replaced existing parts with commercial grade components. The inspectors determined that, in accordance with procedure GNP-01.01.01, Determination of Nuclear Safety Designed Classifications, QA [Quality Assurance] Type and EQ [Environmental Qualification] Type, the new components should have been designated as augmented quality. As part of corrective actions, the licensee revised its parts database to show the appropriate classification for parts for the diesel. The issue was entered into the corrective action program.
The inspectors determined that the finding is greater than minor 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). Specifically, the installation of parts in equipment with a lower quality designation than required potentially impacted equipment reliability. The finding was determined to be of very low safety significance (Green). This finding has a cross-cutting aspect in the area of problem identification and resolution because the licensee failed to take timely effective corrective actions for a similar prior occurrence. Barriers to prevent recurrence had not been established during supervisory reviews that granted multiple extensions to the corrective actions for the prior occurrence.
Inspection Report# : 2007002 (pdf)
Significance:      Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Nuclear Instrument Test Performed Contrary to Procedural Requirements The inspectors identified a finding of very low safety significance (Green) and an associated Non-Cited Violation of Technical Specification 6.8, Procedures, when the licensee failed, on January 8, 2007, to follow procedures for performing the monthly surveillance test on power range instrument N-42 and failed to obtain an approved procedure change as required by administrative procedures when the technicians established an alternate ground point contrary to procedural requirements. As part of corrective actions, the licensee counseled the technicians involved and discussed the event with all members of the instrument and control department. The issue was entered into the corrective action program.
 
The inspectors determined that the finding is greater than minor because it was associated with the procedure quality 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). Specifically, the procedure required the use of the ground associated with the related card to verify proper continuity within the circuit and the use of an alternate ground point was a substantive change to the procedure. The finding was determined to be of very low safety significance (Green). This finding has a cross-cutting aspect in the area of human performance because personnel did not follow procedures.
Inspection Report# : 2007002 (pdf)
Significance:      Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Seismic Storage requirements in Station Housekeeping Procedure The inspectors identified a finding of very low safety significance (Green) and an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, during a review of procedures related to the control and storage of material. On March 21, 2007, the inspectors identified a number of unsecured equipment carts located in the vicinity of the seismically-classified, safety-related auxiliary building special ventilation system.
The inspectors concluded that, although this was allowed by plant procedure GNP-01.31.01, Plant Cleanliness and Storage, it was a condition that potentially affected quality (safe operation of the ventilation system during a seismic event) and should not have been allowed by the procedure. As part of corrective action, the licensee properly secured the carts and evaluated other carts positioned near safety-related equipment. The issue was entered into the corrective action program.
The inspectors determined that the finding is greater than minor because it was associated with the procedure quality 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). Specifically, the procedure allowed uncontrolled storage of materials in the vicinity of the auxiliary building special ventilation system that could render the system inoperable during a seismic event. The finding was determined to be of very low safety significance (Green). This finding has a cross-cutting aspect in the area of human performance because the licensee failed to provide accurate procedures to assure the operability of safety-related equipment was maintained.
Inspection Report# : 2007002 (pdf)
Significance:      Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Evaluate the Potentially Inadequate Design of the Service Water System The inspectors identified a finding of very low safety significance (Green) and associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, Instruction, Procedures, and Drawings, when the licensee failed to have in place adequate procedures to preclude a common mode failure of both trains of the safety-related service water (SW) system. Specifically, adequate procedures were not established for the maintenance of the SW system to prevent corrosion and degradation of the plant equipment water (PEW) filter vessels from affecting the safety-related SW bearing water supply components. As part of corrective actions, the licensee wrote the appropriate maintenance procedures. The issue was entered into the corrective action program.
The inspectors concluded that this finding is greater than minor because it was associated with the procedure quality and equipment performance attributes 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. Specifically, the lack of appropriate procedures allowed the degradation of PEW components to cause the inoperability of two safety-related SW pumps. The finding was determined to be of very low safety significance.
This finding has a cross-cutting aspect in the area of problem identification and resolution because the licensee failed to thoroughly evaluate problems such that the resolutions addressed causes and extent of condition, as necessary.
Inspection Report# : 2007002 (pdf)
 
Significance:      Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Foreign Material in Containment as a Result of Inadequate Containment Closure Inspections The inspectors identified a finding of very low safety significance (Green) and an associated Non-Cited Violation of Technical Specification 6.8, Procedures, on February 28, 2007, when the licensee failed to adequately perform a containment closeout inspection to ensure that debris and foreign materials were identified and removed in accordance with plant procedures. Specifically, inspectors identified unsecured metal sheets inside containment during a walkdown. As part of corrective actions, the sheets were removed from containment. The issue was entered into the corrective action program.
This finding is greater 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, the failure to identify and remove the steel sheets from containment could have affected the availability of both trains of the residual heat removal system (the accident recirculation sump) during a loss-of-coolant accident because of increased debris generation caused by the unsecured sheets. The finding was determined to be of very low safety significance (Green). This finding has a cross-cutting aspect in the area of human performance because personnel did not follow procedures, causing a condition to exist that potentially impacted the operability of both trains of the residual heat removal system.
Inspection Report# : 2007002 (pdf)
Significance: SL-IV Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Notify NRC of Licensee medical Condition Change in Accordance with 10 CFR50.74 The inspectors identified a finding of very low safety significance and an associated Severity Level IV, Non-Cited Violation of 10 CFR 50.74 for the licensees failure to notify the NRC that one of its licensed operators was taking prescribed medication for a potentially disqualifying medical condition (hypertension). After a review of the licensed operators medical status was completed by the NRCs medical review officer, a condition was added to the operators license requiring him to take the medication as prescribed. The facility licensee entered this issue in their corrective action program. They required the individual licensed operator to take the medication as prescribed and incorporated these lessons learned in their requalification training program to ensure all licensed operators are aware of the requirement to notify the NRC of changes in their medical status.
Because violations of 10 CFR 50.74 affect the NRCs ability to perform its regulatory function, this finding was evaluated using the traditional enforcement process. In accordance with the NRC Enforcement Policy, this finding was determined to be greater than minor because the medical condition that was not reported required a change to the operators NRC license. Because the operator was always in the presence of other licensed operators while performing licensed duties and made no operational errors while he was taking the prescribed medication before his license had been appropriately revised, NRC management has determined this issue is a Green finding, of very low safety significance. This issue is considered an NCV because it was entered into the licensees corrective action program.
This finding also has a cross-cutting aspect in the area of human performance because a standard, specifically American National Standards Institute/American Nuclear Society (ANSI/ANS) 3.4, Medical Certification and Monitoring of Personnel Requiring Operator Licenses for Nuclear Power Plants, was available but not correctly implemented. The correct implementation of the standard would have led to a proper notification of the NRC and timely conditioning of the operators NRC license.
Inspection Report# : 2007002 (pdf)
Significance:      Jan 31, 2007 Identified By: NRC Item Type: VIO Violation Failure to Evaluate Operability of the "A" EDG when a Fuel Oil Leak was Identified A finding that was preliminarily determined to be of substantial safety significance (Yellow), and an associated apparent violation of Technical Specification 6.8, Procedures, was identified for a fuel oil leak on the A emergency diesel generator (EDG) that was identified on June 28, 2006, but was not repaired until 51 days later on
 
August 17. In December 2006, the licensee tested the fitting and copper tubing that was the source of the leak to assess the leaks effect on the operability of the diesel. The licensee concluded that the leak rendered the diesel inoperable for those 51 days. As part of corrective action, the licensee replaced the leaking fuel oil line and reinforced with plant personnel the procedural requirements to properly evaluate equipment problems. The licensee also entered the issue into its corrective action program.
The finding was more than minor because if left uncorrected it would become a more significant safety concern during use of the A EDG to mitigate a loss of offsite power event. Specifically, the A EDG would have failed after approximately four hours due to the loss of fuel oil through the failed fuel line tubing, and the systems that respond to accidents and are powered by the A EDG would not be available. A Significance Determination Process Phase 3 risk analysis preliminarily determined that this finding was of substantial safety significance (Yellow). This finding has a cross-cutting aspect in the area of human performance because procedures were available, but not followed, that could have resulted in the leak being promptly repaired.
After considering the information developed during the inspection, the NRC has concluded that the inspection finding is appropriately characterized as Yellow, i.e., an issue with substantial safety significance that will result in additional NRC inspection and potentially other NRC action.
Inspection Report# : 2007007 (pdf)
Inspection Report# : 2007009 (pdf)
Significance:        Dec 14, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify Emergency Diesel Generator Air Intake Temperature Limitations Impact Upon Ability to Meet Technical Specification Surveillance Requirements A finding of very low safety significance was identified by the inspectors for a violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action. The licensee failed to identify the impact of air intake temperature limitation on the ability of the emergency diesel generators to meet Technical Specification surveillance loading requirements at elevated temperatures. Once identified, the licensee established 75 degrees Fahrenheit as a maximum outside temperature for emergency diesel generator operability. The primary cause of this violation was related to the cross-cutting area of Problem Identification and Resolution, because the licensee failed to ensure that an issue potentially impacting nuclear safety was promptly identified, fully evaluated, and that actions were taken to address safety issues in a timely manner, commensurate with their significance.
The issue was more than minor because the failure to identify that the emergency diesel generators would not be able to meet Technical Specification surveillance requirements at elevated temperatures could have resulted in the emergency diesel generators being considered operable when, in fact, they had less operational margin than required by Technical Specifications. The issue was of very low safety significance because both of the emergency diesel generators were determined to be capable of carrying their respective design basis accident loads below the outside temperature limitations that the licensee had in place. The issue was a NCV of 10 CFR Part 50, Appendix B, Criterion XVI, which required that conditions adverse to quality are promptly identified and corrected.
Inspection Report# : 2006016 (pdf)
Significance:        Dec 14, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Provide Suppression for Safe Shutdown Equipment in Appendix R, III.G.3 Area A finding of very low safety significance was identified by the inspectors for a violation of 10 CFR Part 50, Appendix R, Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979. The licensee failed to provide required fire suppression coverage in fire zone AX-32 for the safe shutdown functions of source range monitoring, isolation of a steam generator blowdown line, and pressurizer level instrumentation. Once identified, the licensee entered the issue into their corrective action program and implemented compensatory measures.
This issue was more than minor because the failure to provide suppression for redundant trains of safe shutdown equipment increased the likelihood that alternative shutdown methods would have to be used in the event of a fire.
The issue was of very low safety significance because of the mitigating systems, which would have remained
 
available in the event of a fire. The issue was a NCV of 10 CFR Part 50, Appendix R, Section III.G.3, which required fixed suppression systems for alternative shutdown areas such as fire zone AX-32.
Inspection Report# : 2006016 (pdf)
Significance:        Oct 23, 2006 Identified By: NRC Item Type: NCV NonCited Violation Scaffolding in Contact with the Safety Injection System Affects Operability A finding of very low safety significance and an associated non-cited violation of Technical Specification 6.8, Procedures, was identified by the inspectors on October 23, 2006, for the failure to install scaffolding in accordance with station procedures. Specifically, scaffolding was installed inside containment that was too close to, or was in contact with, safety injection system components and piping. As part of corrective actions, the licensee removed the scaffolding and enhance the station procedure for scaffolding. The issue was entered into the licensees corrective action program.
This finding is greater than minor because it was 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).
Specifically, improperly positioned scaffolding could have impeded or prevented proper operation of the safety injection system during an accident. The finding was of very low safety significance (Green) because it did not require a quantitative assessment. This finding has a cross-cutting aspect in the area of human performances because personnel did not follow the procedure for scaffolding.
Inspection Report# : 2006005 (pdf)
Significance:        Oct 11, 2006 Identified By: NRC Item Type: NCV NonCited Violation Inadvertent Drain Down of the Reactor Coolant system During Fill and Vent of the Containment Spray System A finding of very low safety significance and an associated non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed during performance of a plant safety-related procedure to fill and vent the containment spray system, on October 11, 2006, when water was inappropriately diverted from the reactor coolant system to the residual heat removal system. As part of corrective actions, the licensee revised the procedure to ensure the systems were properly aligned during fill and vent activities. The issue was entered into the licensees corrective action program.
This finding is greater than minor because if left uncorrected it would become a more significant safety concern in that the use of other inadequate procedures could have rendered inoperable important mitigating equipment, such as the containment spray and residual heat removal systems. Additionally, the finding was associated with the procedure quality and configuration control attributes 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 was determined to be of very low safety significance (Green) because it did not require a quantitative assessment. This finding has a cross-cutting aspect in the area of human performances because the licensee failed to provide complete, accurate, and up-to-date procedures to fill and vent the containment spray system.
Inspection Report# : 2006005 (pdf)
Significance:        Sep 30, 2006 Identified By: NRC Item Type: FIN Finding Technical Specification LCO Not Entered for Diesel Generators Inoperable While in Refueling Shutdown A finding of very low safety significance (Green) was identified by the inspectors when the licensee failed to properly apply shutdown Technical Specifications (TSs) for the residual heat removal (RHR) system with both emergency diesel generators (EDGs) declared inoperable. While reviewing startup preparations being made for a mode change, the inspectors identified that TSs required both RHR systems to be operable and that both EDGs were inoperable due to tornado failure susceptibilities, thereby rendering both trains of RHR inoperable as required by the related power
 
requirements TS. The licensee concurred with the inspectors observations, prevented the mode change, and issued the related licensee event report. Corrective actions, to date, included restoration of EDG operability prior to making a mode change and procedural enhancements.
The inspectors determined that the finding is greater than minor because if left uncorrected it would become a more significant safety issue: the licensee would have made a mode change without the required operable equipment. This finding was of very low safety significance because the licensee returned the EDGs to operability prior to making any mode changes, no violation of NRC requirements was identified, and the finding did not require a quantitative assessment using Check List 4 for PWR Shutdown Operation with Time to Boil >2 hours and Inventory in the Pressurizer. The cause of this finding was related to the crosscutting area of human performance because procedures, specifically TSs, were available but not followed, that would have facilitated the proper performance of the task.
Inspection Report# : 2006004 (pdf)
Barrier Integrity Significance:      Jun 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedure for Surveillance Testing of Auxiliary Building Special Ventilation Zone The inspectors identified a finding having very low safety significance and an associated non-cited Violation of 10 CFR 50, Appendix B, Criterion V, Instructions , Procedures, and Drawings,while reviewing surveillance testing procedures for the auxiliary building special ventilation zone (Zone SV). Specifically, the licensee procedure for tracking the amount of in-leakage into the Zone SV did not have adequate criteria to capture degraded conditions, nor ensure that the acceptance criteria reflected the design requirements of the system. The cause of this finding is related to the cross-cutting area of problem identification and resolution because the licensee failed to properly evaluate multiple condition reports for operability and extent of condition. (P1(c))
This finding was determined to be more than minor because, if left uncorrected, the failure to evaluate barrier breaches that did not have breach permits could become a more significant safety concern. Specifically, if left unmonitored the breaches without barrier permits could potentially exceed the allowable design limits. The finding was evaluated using IMC 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations. The answer to Question 1 in the Significance Determination Process Phase 1 Screening Worksheet in the Containment Barrier Cornerstone column was yes; therefore, this finding is of very low safety significance (Green). Corrective actions to date included revisions to procedure FPP-08-09, to track barrier breaches that result from degraded conditions and provide conservative acceptance criteria.
Inspection Report# : 2007003 (pdf)
Significance:      Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Loss of Coolant Environment Improperly Considered in Containment Fan Coil Unit Calculation The inspectors identified a finding having very low significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, "Design Control." Specifically, the licensee failed to use the correct data when determining the most limiting conditions on the safety related motors of the containment fan coil units (CFCU). The engineers failed to use the combination of the greatest density of the air-steam mixture following a loss of coolant accident (LOCA) with the greatest flow rate attributed to the fans by testing. As a result, the licensee was not aware that post LOCA, the motors will be operating at 113 percent of their design rating, and drawing 13 additional kW from each diesel generator.
Upon discovery, the licensee recalculated the motors' horsepower, recalculated the service factor (percent above continuous design rating) at which the motors will be operating, and recalculated the elevated current that will be drawn by the motors, and the elevated current at degraded voltage. In addition, the licensee had to reevaluate whether the over-current trip setpoint of the motors will be exceeded.
This issue was more than minor because the assumed power drawn by the motors was significantly less, the existing
 
margin was already low, and as a consequence, the error resulted in a significant reduction in margin. This issue also impacted the capability of the emergency diesel generators to supply the required power to the CFCU's motors. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
Significance: SL-IV Dec 14, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Fully Update Updated Safety Analysis Report A finding of very low safety significance was identified for the licensees failure to adequately update the Update Safety Analysis Report (USAR) in accordance to 10 CFR 50.71, Maintenance of Records, Making of Reports. The licensee failed to update the USAR to fully reflect changes and analyses made in response to NRC Generic Letter (GL) 96-06, Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions. Once identified, the licensee entered this issue into their corrective action program. The primary cause of this violation was related to the cross-cutting area of Human Performance because the licensee failed to ensure that personnel, equipment, procedures, and other resources were available and adequate to assure nuclear safety.
Specifically, the licensee failed to provide adequate engineering procedural guidance concerning the required content of USAR updates.
Because this issue potentially impacted the NRCs ability to perform its regulatory function, this finding was evaluated using the traditional enforcement process. The finding was determined to be more than minor because of the failure to provide complete licensing and design basis information in the USAR could result in either the licensee making an inappropriate licensing interpretation or the NRC making an inappropriate regulatory decision based on incomplete information in the USAR. The issue was of very low safety significance because no instances were identified where the failure to appropriately update the USAR impeded or influenced a regulatory decision, or resulted in an actual loss of safety function. The issue was a NCV of 10 CFR 50.71(e) which required that the USAR be updated to include the effects of all analyses of new safety issues performed by or on behalf of the licensee at Commission request.
Inspection Report# : 2006016 (pdf)
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.
Miscellaneous
 
Last modified : August 24, 2007 Kewaunee 3Q/2007 Plant Inspection Findings Initiating Events Significance:      Sep 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement Maintenance Rule (a)(1) in Corrective Actions on the "G" Instrument Air Compressor The inspectors identified a finding of very low safety significance and an associated non-cited violation of 10 CFR 50.65(a)(1), Requirements for monitoring the effectiveness of maintenance at nuclear power plants. Specifically, as of August 25, 2007, the licensee failed to implement the Maintenance Rule (a)(1) action plan which had been incorporated into plant procedure N-AS-01 to preclude a loss of the G air compressor. The licensee entered the issue into their corrective action program. Corrective actions have included implementation of the procedural requirements of N-AS-01 for both the G and F air compressors.
The finding is greater than minor because it relates to a licensee failure to implement prescribed significant compensatory measures to manage risk and implement the 10 CFR 50.65(a)(1) action plan. Additionally, the finding is 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. The inspectors evaluated the finding using Inspection Manual Chapter 0609, Appendix A, Significance Determination Process, and determined that this finding is of very low safety significance by answering No to all questions in the Initiating Events Cornerstone column.
Inspection Report# : 2007004 (pdf)
Significance:      Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation No Analysis or Procedures to Establish Operability of the TAT Source The inspectors identified a finding having very low significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to evaluate the capability of the 345 kV offsite power supply when isolated from the 138 kV switchyard and to translate this criteria into procedures.
This issue was more than minor because procedures allowed operation of the station in unanalyzed configurations for which operability of one offsite source could not be assured and new calculations were needed to ensure that the design basis was met. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
Significance:      Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Incorporate and Effectively Implement Operating Experience in RTB Maintenance Activities The inspectors identified a finding having very low significance and an associated NCV of 10 CFR 50.65(a)(3) for the failure to incorporate external and internal operating experience into preventive maintenance activities for the reactor trip breakers. 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 previous breaker issues and did not perform adequate extent of condition reviews. Specifically, the licensee initiated several corrective action documents in response to identified issues; however, did not perform adequate evaluations of the conditions to address the cause or resolve the identified issue. (P.1.(c))
 
This issue was more than minor because the licensee failed to ensure that the RTBs, and their associated cell assemblies, had been maintained in a continuous state of operational readiness by performing effective maintenance and surveillance activities in accordance with relevant vendor specifications and available operating experience. The issue was of very low safety significance based on a Phase 1 screening because the finding did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available.
Inspection Report# : 2007006 (pdf)
Significance:      Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Acceptance Criteria Not Met Due to Failure to Follow Procedure The inspectors identified a finding having very low significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings. Specifically, on May 22, 2006 during the performance of PMP-47-01, maintenance technician recorded a trip bar force of 32 ounces when testing RTB S/N 850-027-1, which exceeded the acceptance criteria; however, no further actions were taken as required by the test. This finding has a cross-cutting aspect in the area of Human Performance, Work Practices because the licensee did not perform an adequate peer check of the surveillance results. Specifically, several individuals including the person performing the task did not identify that the RTB trip bar force exceeded the acceptance criteria. (H.4.(c))
This issue was more than minor because not meeting the acceptance for the trip bar force impacted the reliability of the RBTs because excessive force could result in a failure to trip the breaker. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
Significance:      Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Implement E-0-05, "Response to Natural Events," During a High Wind Advisory The inspectors identified a finding of very low safety significance (Green) and an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to adequately implement procedure E-O-05, Response to Natural Events, during a high wind advisory. Specifically, on February 22, 2007, during the advisory, the inspectors identified several items stored outdoors near the plant main output transformer that could become missile hazards during actual high winds. As part of corrective actions, the licensee removed the items. The issue was entered into the licensees corrective action program.
The inspectors determined that the finding is greater than minor because, if left uncorrected, the loose items could become a more significant safety concern by allowing the accumulation of missile hazards in these areas, thereby increasing the likelihood of an initiating event. The inspectors determined that the finding warranted evaluation using the Significance Determination Process (SDP) because the finding was associated with an increase in the likelihood of an initiating event. The finding was determined to be of very low safety significance (Green). This finding has a cross-cutting aspect in the area of human performance because the licensee failed to communicate decisions and the basis for decisions to personnel who have a need to know the information in order to perform work safely and in a timely manner.
Inspection Report# : 2007002 (pdf)
Significance:      Dec 31, 2006 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Pre-Fire Strategy Identified in Cable Spreading Room A finding of very low safety significance and an associated non-cited violation of Technical Specification 6.8, Procedures, was identified by the inspectors for the failure to identify radiological and toxic hazards in the cable spreading area fire zone pre-fire strategy. These hazards were from a radioactively contaminated lead pipe in the fire zone that could melt during certain fire scenarios. As part of corrective actions, the licensee appropriately revised the strategy. The issue was entered into the licensees corrective action program.
 
The finding is greater than minor because it was associated with the external factors - fire 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 provide adequate warnings and guidance in the pre-fire plan related to these hazards could have adversely impacted the fire brigades ability to properly respond to a fire. This impact could increase the likelihood of damage to equipment, causing an upset pf plant stability. NRC management review determined the finding to be of very low safety significance (Green), due to the extensive training provided to fire brigade members to deal with unexpected contingencies. This finding has a cross-cutting aspect in the area of human performance because the licensee failed to provide complete, accurate, and up-to-date pre-fire strategies for the fire brigade to respond to a fire.
Inspection Report# : 2006005 (pdf)
Mitigating Systems Significance: SL-IV Sep 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Update the Updated Safety Evaluation Report The inspectors identified a finding of very low safety significance for the licensee's failure to adequately update the Updated Safety Analysis Report (USAR) in accordance to 10 CFR 50.71, Maintenance of records, making of reports. The licensee failed to update the USAR to fully reflect changes and analyses made in response to license amendment 184. Once identified, the licensee entered this issue into its corrective action program.
Because this issue potentially impacted the NRC's ability to perform its regulatory function, this finding was evaluated using the traditional enforcement process. The finding is greater than minor because of the failure to provide complete licensing and design basis information in the USAR could result in either the licensee making an inappropriate licensing interpretation or the NRC making an inappropriate regulatory decision based on incomplete information in the USAR. The issue is of very low safety significance based upon a Phase 2 significance determination analysis of the associated technical issue. The issue was a NCV of 10 CFR 50.71(e), which required that the USAR be updated to include the effects of all safely evaluations performed by the licensee in support of requested license amendments.
The primary cause of this violation is related to the cross-cutting area of problem identification and resolution because the extent of condition review performed for a recent and similar violation failed to identify the issue even though it was within the scope of the extent of condition review which had been performed Inspection Report# : 2007004 (pdf)
Significance: SL-IV Jun 30, 2007 Identified By: NRC Item Type: VIO Violation Failure to Perform a 10 CFR 50.59 Evaluation for Compensatory Measures Associated with a Procedure Change The inspectors identified a finding having very low safety significance and an associated Severity Level IV, Cited Violation of 10 CFR 50.59 while reviewing unresolved items URI 05000305/2006003-04, Adequacy of Compensatory Actions for Potential Turbine Missile Strike of Control Room Ventilation Cooling; and URI 05000305/2006016-01, Adequacy of 10 CFR 50.59 Screening for Procedure Change. Specifically, the licensee failed to properly interpret design and licensing basis requirements associated with protection against external events and as a result did not perform a 10 CFR 50.59 evaluation. The cause of this finding is related to the cross-cutting area of problem identification and resolution because the licensee had similar prior problems that, if effectively evaluated and resolved, could have prevented this issue. (P.1(c))
This finding was determined to be more than minor because the inspectors determined that the procedure change would have ultimately required NRC approval. The procedure changes, in the form of compensatory operator actions, adversely impacted the operation of control room recirculation system following a tornado. A Phase 1 significance determination of this finding using IMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," using the Severe Weather Screening Criteria questions was completed. Since the loss of the control room recirculation system would not result in an initiating event or degrade two or more trains of a multi-train safety system, the issue screened as Green.
 
Inspection Report# : 2007003 (pdf)
Significance:      May 18, 2007 Identified By: NRC Item Type: NCV NonCited Violation Procedure Non-Compliance The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to adequately implement procedure DNAP-1604, Cause Evaluation Program, and the Cause Evaluation Handbook during investigative analyses of root cause, collective significance, and apparent cause evaluations. The licensee subsequently revised several apparent cause evaluations (ACEs), such as ACE 3374 on the diesel generator B fuel rack shaft binding, and completed industry benchmarking to improve root cause evaluation and ACE investigative analysis.
This finding was associated with the Mitigating Systems Cornerstone. The finding was more than minor because, if left uncorrected, the licensees analyses of conditions adverse to quality, such as the investigation of the diesel generator B fuel rack shaft binding, as documented in ACE 3374, would not be performed at an appropriate investigative depth for cause determination. The inspectors assessed the significance of this finding as very low safety significance (Green) because the finding did not represent an actual loss of safety function of the equipment. The finding was associated with cross-cutting aspect P.1(c), in the area of problem identification and resolution, corrective action program, because the licensee failed to thoroughly analyze the sequence of events and the cause and effect relationships potentially impacting the causal determination of CAP evaluations.
Inspection Report# : 2007008 (pdf)
Significance:      Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation No Motor Starting Analyses for Offsite Power Supply The inspectors identified a finding having very low significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to perform motor starting studies to demonstrate that motors would successfully start when connected to the offsite power supply. Upon discovery, the licensee provided additional data and compensatory measures to justify operability.
The inspectors determined that the performance deficiency was more than minor because the lack of a formal motor starting calculations resulted in the adequacy of important aspects of the design not being demonstrated, such that further evaluation needed to be performed in order to demonstrate that the equipment could perform its safety function. Although, by the end of the inspection, the licensee was able to demonstrate operability, at the time of discovery there was reasonable doubt on the operability of motors. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
Significance:      Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Increased Cable Resistance Due to Accident Temperatures The inspectors identified a finding having very low significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to consider the effects of accident temperatures on cable resistance in voltage drop calculations. Upon discovery, the licensee performed preliminary calculations to verify operability of the circuits.
This issue was more than minor because the calculational errors had more than a minimal effect on the outcome of the calculation, considerably impacting the available margin of the system such that further evaluation needed to be performed in order to demonstrate that the equipment could perform its safety function. Although, by the end of the inspection, the licensee was able to demonstrate operability; at the time of discovery there was reasonable doubt on the operability of the circuits. The issue was of very low safety significance based on a Phase 1 screening in
 
accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
Significance:      Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Ensure Adequate 125 Vdc Breaker Interrupting Short Circuit Current Capability The inspectors identified a finding having very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to ensure that four of the 125 VDC circuit breakers had adequate interrupting short circuit fault current capability. Upon discovery, the licensee performed a preliminary evaluation, and verified that the most likely fault would result in a lower short circuit fault current than the breakers rating.
This issue was more than minor because the failure could have affected the operability of the breaker/DC Bus and could have resulted in the loss of DC power to safe shutdown equipment in the event of short circuit faults. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
Significance:      Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Use Actual Minimum Voltage Value in 125Vdc Voltage Drop Calculation The inspectors identified a finding having very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to use correct design input data into the 125 VDC safeguard battery calculation. The licensee used a battery terminal voltage value of 117.49 volts for BRA-101 and 118.95 volts for BRB-101, for the first minute, and did not compensate for worse case conditions. Upon discovery, the licensee performed preliminary evaluation and verified that safe shutdown equipment have adequate voltage using the battery terminal voltage value of 113.87 volts.
This issue was more than minor because the failure to use correct design input had more than a minimal effect on the outcome of the voltage drop calculation, considerably impacting the available margin of the system such that further evaluation needed to be performed in order to demonstrate that equipment could perform its safety function.
Although, during the inspection, the licensee was able to demonstrate operability; at the time of discovery there was reasonable doubt on the operability of circuits. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
Significance:      Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Acceptance Criteria in 125 Vdc Station Battery Load Tests Procedures The inspectors identified a finding having very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings. Specifically, the licensee failed to include the acceptable minimum battery terminal voltage, during the first minute, into the acceptance criteria for battery load test procedures SP-38-102A/B Station Battery Load Test. Upon discovery, the licensee entered the issue into its corrective action program to revise the acceptance criteria of procedures SP-38-102A/B to include this requirement.
This issue was more than minor because the failure to ensure that the battery terminal voltage during the first minute battery discharge did not drop below the design input value could have affected the operability of safety related equipments in the event of a design basis accident and or station blackout conditions. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
 
Inspection Report# : 2007006 (pdf)
Significance:      Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Adequate Control Voltage for 4160V Breaker's Closing Coil was not Assured The inspectors identified a finding having very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to assure that the minimum available control voltage at the 4160V breakers was adequate to energize the closing coils during all conditions. Upon discovery, the licensee performed preliminary calculation and verified operability of the emergency diesel generators 4160V breakers following loss of all AC power conditions.
This finding was more than minor because the failure to assure adequate control voltage was available to close the 4160V breakers would have affected the capability of emergency diesel generators and other safety related equipments to respond to initiating events. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
Significance:      Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Safeguard Battery Load Profile Did Not Include LOOP/LOCA Loads The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control having very low safety significance for the licensees failure to assure that safeguard battery loads profile was adequate to meet all USAR requirements. Specifically, the licensee failed to verify that the battery loading profile for loss of coolant accident (LOCA) coincide with loss of all AC power condition was bounded by the station blackout condition loading to ensure adequate battery sizing and testing. Upon discovery, the licensee was able to show that the charger will be available upon the start of the emergency diesel generator and will provide additional support. This issue was entered into the licensees corrective action program to revise the battery calculation to include the LOCA loads.
This finding was more than minor because the failure to include the LOCA loads in the battery sizing and testing did not ensure the capability of the battery to provide adequate DC power in accordance with USAR requirements. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
Significance:      Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Electrolytic Capacitors in Spare Safeguard Battery Charger Not Periodically Energized The inspectors identified a finding having very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action. Specifically, the licensee failed to incorporate previously identified vendor recommendation to periodically energize the spare 125 VDC safeguard battery charger for at least a half-hour every 18 months to ensure the operability of the electrolytic capacitor in the charger. The licensee has previously entered the vendor recommendation into their corrective action in 2002, however, all actions were closed but the recommendation was never implemented. Following discovery, the licensee entered the issue into its corrective action program and declared the spare charger inoperable. The primary cause of this violation was related to the cross-cutting area of problem identification and resolution because the licensee failed to take appropriate corrective actions to address a previously failed charger. Specifically, the licensee developed corrective actions which included incorporating pertinent vendor recommendation into the preventive maintenance program but closed the action without ensuring completion (P.1.d)
This issue was more than minor because the failure to periodically energize the spare charger did not ensure the operability and reliability of the spare charger when needed. The issue was of very low safety significance based on a
 
Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
Significance:        Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Diesel Loading Calculations Non Conservative The inspectors identified a finding having very low significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to properly account for all loads on the diesel generators. Upon discovery, the licensee provided additional data and initiated procedure changes to ensure diesels were loaded within their ratings.
The inspectors determined that the performance deficiency was more than minor because the lack of adequate diesel generator loading calculations resulted in some diesel loads not being properly accounted for, such that further evaluation needed to be performed in order to demonstrate that the equipment could perform its safety function.
Although, by the end of the inspection, the licensee was able to demonstrate operability, at the time of discovery there was reasonable doubt on the operability of equipment. The inspectors screened the finding using IMC 0609, Appendix A. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
Significance:        Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation RWST Level Instruments Do Not Protect SI and RHR Pumps from Excessive Air Entrainment The inspectors identified a finding having very low significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to incorporate the results of design calculations with respect to minimum refueling water storage tank (RWST) level and transfer of suction sources into the appropriate emergency operating. Procedures allowed operators to transfer suction at 4 percent indicated level in the RWST; however, at this level, significant air entrainment may damage the pumps. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program because the licensee did not thoroughly evaluate problems such that the resolution addresses the extent of condition (P.1.c).
This issue was more than minor because the existing margin was already low and as a consequence, the large error associated with the level instrument resulted in eliminating the entire margin, and jeopardized the functionality of the pumps taking suction from the RWST due to excessive air entrainment. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations SDP Phase 1.
Inspection Report# : 2007006 (pdf)
Significance:        Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Non-Conservative Assumption Used in Service Water Flow Model Calculation The inspectors identified a finding having very low significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to appropriately account for service water strainer plugging in the service water system flow model. Upon discovery, the licensee placed this issue into their corrective action program and planned to formally revise the service water system flow model to reflect plugging of both strainers in a train.
The issue was more than minor because the error had more than a minimal effect on the outcome of the calculation, considerably impacting the available margin of the system such that further evaluation needed to be performed in order to demonstrate that the service water system could perform its safety function. The issue was of very low safety significance because the issue was a design issue confirmed to not result in a loss of operability.
 
Inspection Report# : 2007006 (pdf)
Significance:        Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Screen House Ventilation Damper Maintenance The inspectors identified a finding having very low significance and an associated NCV of 10 CFR Part 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, Paragraph (b)(2), for the licensees failure to scope the closing function of the screenhouse ventilation dampers into the monitoring program.
Specifically, the degraded screen-house dampers fail to close and maintain ambient temperatures > 60 &deg;F such that service water system would remain operable and available after a station blackout event with severely cold outside temperatures. Following discovery, the licensee entered the issue into its corrective action program for resolution.
This issue was more than minor because the licensee had not included the closing function of the screen-house ventilation dampers within the scope of its program for implementation of the Maintenance Rule. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
Significance:        Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Non Conservative Assumption Used for "B" CCW Pump Room Heat Gain Calculation The inspectors identified a finding having very low significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III. Specifically, the licensee failed to account for component cooling water (CCW) piping temperatures as high as 176&deg;F in the CCW B pump room and the impact upon the temperature in the CCW B pump room. As a result, the licensee used the non-conservative results in an operability evaluation for the auxiliary building fan coil unit (FCU). Upon discovery, the licensee placed this issue into their corrective action program, performed an immediate operability evaluation, and planned to perform a more thorough evaluation. This finding has a cross-cutting aspect in the area of human performance associated with decision making because the licensee did not use conservative assumptions. Specifically, the licensee failed to account for higher CCW piping temperatures because the licensee did not model the CCW room properly and did not use the maximum expected temperature under accident conditions when revising calculation C11156 (H.1.b).
The issue was more than minor because the error because, if left uncorrected, the finding would become a more safety significant concern. The use of a non-conservative value as a basis for operability could allow FCU performance to degrade to unacceptable levels without being detected and corrected. The issue was of very low safety significance because the issue was a design issue confirmed to not result in a loss of operability.
Inspection Report# : 2007006 (pdf)
Significance:        Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Safety Injection Pump Lube Oil Coolers Testing deficiencies The inspectors identified a finding having very low significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control. Specifically, the licensee failed to establish a testing program capable of identifying an unacceptable condition of the safety injection (SI) lube oil coolers. Upon discovery, the licensee initiated a change to the test program methodology and performed back-flushing and inspection on the two SI lube oil coolers. The licensee also assessed that as a result of the very cold temperature of the water of Lake Michigan during the inspection, the cooler was considered operable. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with self- and independent assessments because during a 2005 audit of licensing commitments, the licensee failed to identify that the commitment to perform inspection and maintenance of the SI lube oil coolers in accordance with the licensee's response to Generic Letter 89-13 was not kept (P.3.a).
This issue was more than minor because when later assessed, the licensee realized that the coolers would have failed
 
previous tests when reevaluated performance factors were less than the acceptance criterion of 0.9. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
Significance:      Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inappropriate QA Class Components Installed in TSC Diesel Generator The inspectors identified a finding of very low safety significance (Green) and an associated Non-Cited Violation of Technical Specification 6.8, Procedures, during a review on January 27, 2007, of maintenance performed on the station blackout diesel generator. The maintenance, which was conducted to repair a cooling water leak, inappropriately replaced existing parts with commercial grade components. The inspectors determined that, in accordance with procedure GNP-01.01.01, Determination of Nuclear Safety Designed Classifications, QA [Quality Assurance] Type and EQ [Environmental Qualification] Type, the new components should have been designated as augmented quality. As part of corrective actions, the licensee revised its parts database to show the appropriate classification for parts for the diesel. The issue was entered into the corrective action program.
The inspectors determined that the finding is greater than minor 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). Specifically, the installation of parts in equipment with a lower quality designation than required potentially impacted equipment reliability. The finding was determined to be of very low safety significance (Green). This finding has a cross-cutting aspect in the area of problem identification and resolution because the licensee failed to take timely effective corrective actions for a similar prior occurrence. Barriers to prevent recurrence had not been established during supervisory reviews that granted multiple extensions to the corrective actions for the prior occurrence.
Inspection Report# : 2007002 (pdf)
Significance:      Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Nuclear Instrument Test Performed Contrary to Procedural Requirements The inspectors identified a finding of very low safety significance (Green) and an associated Non-Cited Violation of Technical Specification 6.8, Procedures, when the licensee failed, on January 8, 2007, to follow procedures for performing the monthly surveillance test on power range instrument N-42 and failed to obtain an approved procedure change as required by administrative procedures when the technicians established an alternate ground point contrary to procedural requirements. As part of corrective actions, the licensee counseled the technicians involved and discussed the event with all members of the instrument and control department. The issue was entered into the corrective action program.
The inspectors determined that the finding is greater than minor because it was associated with the procedure quality 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). Specifically, the procedure required the use of the ground associated with the related card to verify proper continuity within the circuit and the use of an alternate ground point was a substantive change to the procedure. The finding was determined to be of very low safety significance (Green). This finding has a cross-cutting aspect in the area of human performance because personnel did not follow procedures.
Inspection Report# : 2007002 (pdf)
Significance:      Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Seismic Storage requirements in Station Housekeeping Procedure The inspectors identified a finding of very low safety significance (Green) and an associated Non-Cited Violation of
 
10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, during a review of procedures related to the control and storage of material. On March 21, 2007, the inspectors identified a number of unsecured equipment carts located in the vicinity of the seismically-classified, safety-related auxiliary building special ventilation system.
The inspectors concluded that, although this was allowed by plant procedure GNP-01.31.01, Plant Cleanliness and Storage, it was a condition that potentially affected quality (safe operation of the ventilation system during a seismic event) and should not have been allowed by the procedure. As part of corrective action, the licensee properly secured the carts and evaluated other carts positioned near safety-related equipment. The issue was entered into the corrective action program.
The inspectors determined that the finding is greater than minor because it was associated with the procedure quality 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). Specifically, the procedure allowed uncontrolled storage of materials in the vicinity of the auxiliary building special ventilation system that could render the system inoperable during a seismic event. The finding was determined to be of very low safety significance (Green). This finding has a cross-cutting aspect in the area of human performance because the licensee failed to provide accurate procedures to assure the operability of safety-related equipment was maintained.
Inspection Report# : 2007002 (pdf)
Significance:      Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Evaluate the Potentially Inadequate Design of the Service Water System The inspectors identified a finding of very low safety significance (Green) and associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, Instruction, Procedures, and Drawings, when the licensee failed to have in place adequate procedures to preclude a common mode failure of both trains of the safety-related service water (SW) system. Specifically, adequate procedures were not established for the maintenance of the SW system to prevent corrosion and degradation of the plant equipment water (PEW) filter vessels from affecting the safety-related SW bearing water supply components. As part of corrective actions, the licensee wrote the appropriate maintenance procedures. The issue was entered into the corrective action program.
The inspectors concluded that this finding is greater than minor because it was associated with the procedure quality and equipment performance attributes 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. Specifically, the lack of appropriate procedures allowed the degradation of PEW components to cause the inoperability of two safety-related SW pumps. The finding was determined to be of very low safety significance.
This finding has a cross-cutting aspect in the area of problem identification and resolution because the licensee failed to thoroughly evaluate problems such that the resolutions addressed causes and extent of condition, as necessary.
Inspection Report# : 2007002 (pdf)
Significance:      Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Foreign Material in Containment as a Result of Inadequate Containment Closure Inspections The inspectors identified a finding of very low safety significance (Green) and an associated Non-Cited Violation of Technical Specification 6.8, Procedures, on February 28, 2007, when the licensee failed to adequately perform a containment closeout inspection to ensure that debris and foreign materials were identified and removed in accordance with plant procedures. Specifically, inspectors identified unsecured metal sheets inside containment during a walkdown. As part of corrective actions, the sheets were removed from containment. The issue was entered into the corrective action program.
This finding is greater 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, the failure to identify and remove the steel sheets from containment could have affected the availability of both trains of the residual heat removal system (the accident recirculation sump) during a loss-of-coolant accident because of increased
 
debris generation caused by the unsecured sheets. The finding was determined to be of very low safety significance (Green). This finding has a cross-cutting aspect in the area of human performance because personnel did not follow procedures, causing a condition to exist that potentially impacted the operability of both trains of the residual heat removal system.
Inspection Report# : 2007002 (pdf)
Significance: SL-IV Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Notify NRC of Licensee medical Condition Change in Accordance with 10 CFR50.74 The inspectors identified a finding of very low safety significance and an associated Severity Level IV, Non-Cited Violation of 10 CFR 50.74 for the licensees failure to notify the NRC that one of its licensed operators was taking prescribed medication for a potentially disqualifying medical condition (hypertension). After a review of the licensed operators medical status was completed by the NRCs medical review officer, a condition was added to the operators license requiring him to take the medication as prescribed. The facility licensee entered this issue in their corrective action program. They required the individual licensed operator to take the medication as prescribed and incorporated these lessons learned in their requalification training program to ensure all licensed operators are aware of the requirement to notify the NRC of changes in their medical status.
Because violations of 10 CFR 50.74 affect the NRCs ability to perform its regulatory function, this finding was evaluated using the traditional enforcement process. In accordance with the NRC Enforcement Policy, this finding was determined to be greater than minor because the medical condition that was not reported required a change to the operators NRC license. Because the operator was always in the presence of other licensed operators while performing licensed duties and made no operational errors while he was taking the prescribed medication before his license had been appropriately revised, NRC management has determined this issue is a Green finding, of very low safety significance. This issue is considered an NCV because it was entered into the licensees corrective action program.
This finding also has a cross-cutting aspect in the area of human performance because a standard, specifically American National Standards Institute/American Nuclear Society (ANSI/ANS) 3.4, Medical Certification and Monitoring of Personnel Requiring Operator Licenses for Nuclear Power Plants, was available but not correctly implemented. The correct implementation of the standard would have led to a proper notification of the NRC and timely conditioning of the operators NRC license.
Inspection Report# : 2007002 (pdf)
Significance:        Jan 31, 2007 Identified By: NRC Item Type: VIO Violation Failure to Evaluate Operability of the "A" EDG when a Fuel Oil Leak was Identified A finding that was preliminarily determined to be of substantial safety significance (Yellow), and an associated apparent violation of Technical Specification 6.8, Procedures, was identified for a fuel oil leak on the A emergency diesel generator (EDG) that was identified on June 28, 2006, but was not repaired until 51 days later on August 17. In December 2006, the licensee tested the fitting and copper tubing that was the source of the leak to assess the leaks effect on the operability of the diesel. The licensee concluded that the leak rendered the diesel inoperable for those 51 days. As part of corrective action, the licensee replaced the leaking fuel oil line and reinforced with plant personnel the procedural requirements to properly evaluate equipment problems. The licensee also entered the issue into its corrective action program.
The finding was more than minor because if left uncorrected it would become a more significant safety concern during use of the A EDG to mitigate a loss of offsite power event. Specifically, the A EDG would have failed after approximately four hours due to the loss of fuel oil through the failed fuel line tubing, and the systems that respond to accidents and are powered by the A EDG would not be available. A Significance Determination Process Phase 3 risk analysis preliminarily determined that this finding was of substantial safety significance (Yellow). This finding has a cross-cutting aspect in the area of human performance because procedures were available, but not followed, that could have resulted in the leak being promptly repaired.
After considering the information developed during the inspection, the NRC has concluded that the inspection finding is appropriately characterized as Yellow, i.e., an issue with substantial safety significance that will result in additional NRC inspection and potentially other NRC action.
 
Inspection Report# : 2007007 (pdf)
Inspection Report# : 2007009 (pdf)
Significance:      Dec 14, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Identify Emergency Diesel Generator Air Intake Temperature Limitations Impact Upon Ability to Meet Technical Specification Surveillance Requirements A finding of very low safety significance was identified by the inspectors for a violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action. The licensee failed to identify the impact of air intake temperature limitation on the ability of the emergency diesel generators to meet Technical Specification surveillance loading requirements at elevated temperatures. Once identified, the licensee established 75 degrees Fahrenheit as a maximum outside temperature for emergency diesel generator operability. The primary cause of this violation was related to the cross-cutting area of Problem Identification and Resolution, because the licensee failed to ensure that an issue potentially impacting nuclear safety was promptly identified, fully evaluated, and that actions were taken to address safety issues in a timely manner, commensurate with their significance.
The issue was more than minor because the failure to identify that the emergency diesel generators would not be able to meet Technical Specification surveillance requirements at elevated temperatures could have resulted in the emergency diesel generators being considered operable when, in fact, they had less operational margin than required by Technical Specifications. The issue was of very low safety significance because both of the emergency diesel generators were determined to be capable of carrying their respective design basis accident loads below the outside temperature limitations that the licensee had in place. The issue was a NCV of 10 CFR Part 50, Appendix B, Criterion XVI, which required that conditions adverse to quality are promptly identified and corrected.
Inspection Report# : 2006016 (pdf)
Significance:      Dec 14, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Provide Suppression for Safe Shutdown Equipment in Appendix R, III.G.3 Area A finding of very low safety significance was identified by the inspectors for a violation of 10 CFR Part 50, Appendix R, Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979. The licensee failed to provide required fire suppression coverage in fire zone AX-32 for the safe shutdown functions of source range monitoring, isolation of a steam generator blowdown line, and pressurizer level instrumentation. Once identified, the licensee entered the issue into their corrective action program and implemented compensatory measures.
This issue was more than minor because the failure to provide suppression for redundant trains of safe shutdown equipment increased the likelihood that alternative shutdown methods would have to be used in the event of a fire.
The issue was of very low safety significance because of the mitigating systems, which would have remained available in the event of a fire. The issue was a NCV of 10 CFR Part 50, Appendix R, Section III.G.3, which required fixed suppression systems for alternative shutdown areas such as fire zone AX-32.
Inspection Report# : 2006016 (pdf)
Significance:      Oct 23, 2006 Identified By: NRC Item Type: NCV NonCited Violation Scaffolding in Contact with the Safety Injection System Affects Operability A finding of very low safety significance and an associated non-cited violation of Technical Specification 6.8, Procedures, was identified by the inspectors on October 23, 2006, for the failure to install scaffolding in accordance with station procedures. Specifically, scaffolding was installed inside containment that was too close to, or was in contact with, safety injection system components and piping. As part of corrective actions, the licensee removed the scaffolding and enhance the station procedure for scaffolding. The issue was entered into the licensees corrective action program.
This finding is greater than minor because it was 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).
Specifically, improperly positioned scaffolding could have impeded or prevented proper operation of the safety injection system during an accident. The finding was of very low safety significance (Green) because it did not require a quantitative assessment. This finding has a cross-cutting aspect in the area of human performances because personnel did not follow the procedure for scaffolding.
Inspection Report# : 2006005 (pdf)
Significance:        Oct 11, 2006 Identified By: NRC Item Type: NCV NonCited Violation Inadvertent Drain Down of the Reactor Coolant system During Fill and Vent of the Containment Spray System A finding of very low safety significance and an associated non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed during performance of a plant safety-related procedure to fill and vent the containment spray system, on October 11, 2006, when water was inappropriately diverted from the reactor coolant system to the residual heat removal system. As part of corrective actions, the licensee revised the procedure to ensure the systems were properly aligned during fill and vent activities. The issue was entered into the licensees corrective action program.
This finding is greater than minor because if left uncorrected it would become a more significant safety concern in that the use of other inadequate procedures could have rendered inoperable important mitigating equipment, such as the containment spray and residual heat removal systems. Additionally, the finding was associated with the procedure quality and configuration control attributes 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 was determined to be of very low safety significance (Green) because it did not require a quantitative assessment. This finding has a cross-cutting aspect in the area of human performances because the licensee failed to provide complete, accurate, and up-to-date procedures to fill and vent the containment spray system.
Inspection Report# : 2006005 (pdf)
Barrier Integrity Significance:        Sep 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Preconditioning of Safety-Related Motor-Operated Valves Prior to Performance of Technical Specification Required Surveillance Testing The inspectors identified a finding of very low safety significance and an associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, during plant preparations to perform Surveillance Procedure SP-23-100B, Train B Containment Spray Pump and Valve Test - IST. Specifically, the inspectors noted on August 8, 2007, that shortly prior to performing the surveillance procedure, the plant had hung safety tags on the containment spray system in order to perform repair activities on IDS-102, a check valve in that system. These tags required that normally open motor- operated valves IDS-202 and IDS-2B be cycled closed and tagged in order to isolate the check valve. Because these motor-operated valves were required to be stroke and time-tested during the performance of the surveillance procedure, and the effects of preconditioning on these valves was not considered prior to implementation of the maintenance activity, the inspectors determined that plant procedures were inadequate to assess preconditioning implications associated with station activities. The licensee entered the issue into their corrective action program. Corrective actions included completion of the surveillance procedure with acceptable results and a evaluation of the test results, which determined that the surveillance test was acceptable.
The finding is greater than minor because it was associated with the configuration control attribute of the Barrier Integrity Cornerstone and 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. The inspectors evaluated the finding using IMC 0609, Appendix A, Significance Determination Process, and determined that this finding is of very low safety significance by answering No to all questions in the containment barriers cornerstone column. The inspectors also determined that the primary cause for this finding is related to the cross-cutting area of human performance. Specifically, under the component of resources, procedures to assess and prevent preconditioning of safety-related components were not complete, accurate, and up-to-date Inspection Report# : 2007004 (pdf)
Significance:        Jun 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedure for Surveillance Testing of Auxiliary Building Special Ventilation Zone The inspectors identified a finding having very low safety significance and an associated non-cited Violation of 10 CFR 50, Appendix B, Criterion V, Instructions , Procedures, and Drawings,while reviewing surveillance testing procedures for the auxiliary building special ventilation zone (Zone SV). Specifically, the licensee procedure for tracking the amount of in-leakage into the Zone SV did not have adequate criteria to capture degraded conditions, nor ensure that the acceptance criteria reflected the design requirements of the system. The cause of this finding is related to the cross-cutting area of problem identification and resolution because the licensee failed to properly evaluate multiple condition reports for operability and extent of condition. (P1(c))
This finding was determined to be more than minor because, if left uncorrected, the failure to evaluate barrier breaches that did not have breach permits could become a more significant safety concern. Specifically, if left unmonitored the breaches without barrier permits could potentially exceed the allowable design limits. The finding was evaluated using IMC 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations. The answer to Question 1 in the Significance Determination Process Phase 1 Screening Worksheet in the Containment Barrier Cornerstone column was yes; therefore, this finding is of very low safety significance (Green). Corrective actions to date included revisions to procedure FPP-08-09, to track barrier breaches that result from degraded conditions and provide conservative acceptance criteria.
Inspection Report# : 2007003 (pdf)
Significance:        Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Loss of Coolant Environment Improperly Considered in Containment Fan Coil Unit Calculation The inspectors identified a finding having very low significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, "Design Control." Specifically, the licensee failed to use the correct data when determining the most limiting conditions on the safety related motors of the containment fan coil units (CFCU). The engineers failed to use the combination of the greatest density of the air-steam mixture following a loss of coolant accident (LOCA) with the greatest flow rate attributed to the fans by testing. As a result, the licensee was not aware that post LOCA, the motors will be operating at 113 percent of their design rating, and drawing 13 additional kW from each diesel generator.
Upon discovery, the licensee recalculated the motors' horsepower, recalculated the service factor (percent above continuous design rating) at which the motors will be operating, and recalculated the elevated current that will be drawn by the motors, and the elevated current at degraded voltage. In addition, the licensee had to reevaluate whether the over-current trip setpoint of the motors will be exceeded.
This issue was more than minor because the assumed power drawn by the motors was significantly less, the existing margin was already low, and as a consequence, the error resulted in a significant reduction in margin. This issue also impacted the capability of the emergency diesel generators to supply the required power to the CFCU's motors. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
Significance: SL-IV Dec 14, 2006 Identified By: NRC Item Type: NCV NonCited Violation Failure to Fully Update Updated Safety Analysis Report
 
A finding of very low safety significance was identified for the licensees failure to adequately update the Update Safety Analysis Report (USAR) in accordance to 10 CFR 50.71, Maintenance of Records, Making of Reports. The licensee failed to update the USAR to fully reflect changes and analyses made in response to NRC Generic Letter (GL) 96-06, Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions. Once identified, the licensee entered this issue into their corrective action program. The primary cause of this violation was related to the cross-cutting area of Human Performance because the licensee failed to ensure that personnel, equipment, procedures, and other resources were available and adequate to assure nuclear safety.
Specifically, the licensee failed to provide adequate engineering procedural guidance concerning the required content of USAR updates.
Because this issue potentially impacted the NRCs ability to perform its regulatory function, this finding was evaluated using the traditional enforcement process. The finding was determined to be more than minor because of the failure to provide complete licensing and design basis information in the USAR could result in either the licensee making an inappropriate licensing interpretation or the NRC making an inappropriate regulatory decision based on incomplete information in the USAR. The issue was of very low safety significance because no instances were identified where the failure to appropriately update the USAR impeded or influenced a regulatory decision, or resulted in an actual loss of safety function. The issue was a NCV of 10 CFR 50.71(e) which required that the USAR be updated to include the effects of all analyses of new safety issues performed by or on behalf of the licensee at Commission request.
Inspection Report# : 2006016 (pdf)
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.
Miscellaneous Last modified : December 07, 2007
 
Kewaunee 4Q/2007 Plant Inspection Findings Initiating Events Significance:      Sep 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement Maintenance Rule (a)(1) in Corrective Actions on the "G" Instrument Air Compressor The inspectors identified a finding of very low safety significance and an associated non-cited violation of 10 CFR 50.65(a)(1), Requirements for monitoring the effectiveness of maintenance at nuclear power plants. Specifically, as of August 25, 2007, the licensee failed to implement the Maintenance Rule (a)(1) action plan which had been incorporated into plant procedure N-AS-01 to preclude a loss of the G air compressor. The licensee entered the issue into their corrective action program. Corrective actions have included implementation of the procedural requirements of N-AS-01 for both the G and F air compressors.
The finding is greater than minor because it relates to a licensee failure to implement prescribed significant compensatory measures to manage risk and implement the 10 CFR 50.65(a)(1) action plan. Additionally, the finding is 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. The inspectors evaluated the finding using Inspection Manual Chapter 0609, Appendix A, Significance Determination Process, and determined that this finding is of very low safety significance by answering No to all questions in the Initiating Events Cornerstone column.
Inspection Report# : 2007004 (pdf)
Significance:      Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation No Analysis or Procedures to Establish Operability of the TAT Source The inspectors identified a finding having very low significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to evaluate the capability of the 345 kV offsite power supply when isolated from the 138 kV switchyard and to translate this criteria into procedures.
This issue was more than minor because procedures allowed operation of the station in unanalyzed configurations for which operability of one offsite source could not be assured and new calculations were needed to ensure that the design basis was met. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
Significance:      Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Incorporate and Effectively Implement Operating Experience in RTB Maintenance Activities The inspectors identified a finding having very low significance and an associated NCV of 10 CFR 50.65(a)(3) for the failure to incorporate external and internal operating experience into preventive maintenance activities for the reactor trip breakers. 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 previous breaker issues and did not perform adequate extent of condition reviews. Specifically, the licensee initiated several corrective action documents in response to identified issues; however, did not perform adequate evaluations of the conditions to address the cause or resolve the identified issue. (P.1.(c))
 
This issue was more than minor because the licensee failed to ensure that the RTBs, and their associated cell assemblies, had been maintained in a continuous state of operational readiness by performing effective maintenance and surveillance activities in accordance with relevant vendor specifications and available operating experience. The issue was of very low safety significance based on a Phase 1 screening because the finding did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available.
Inspection Report# : 2007006 (pdf)
Significance:      Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Acceptance Criteria Not Met Due to Failure to Follow Procedure The inspectors identified a finding having very low significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings. Specifically, on May 22, 2006 during the performance of PMP-47-01, maintenance technician recorded a trip bar force of 32 ounces when testing RTB S/N 850-027-1, which exceeded the acceptance criteria; however, no further actions were taken as required by the test. This finding has a cross-cutting aspect in the area of Human Performance, Work Practices because the licensee did not perform an adequate peer check of the surveillance results. Specifically, several individuals including the person performing the task did not identify that the RTB trip bar force exceeded the acceptance criteria. (H.4.(c))
This issue was more than minor because not meeting the acceptance for the trip bar force impacted the reliability of the RBTs because excessive force could result in a failure to trip the breaker. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
Significance:      Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Adequately Implement E-0-05, "Response to Natural Events," During a High Wind Advisory The inspectors identified a finding of very low safety significance (Green) and an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to adequately implement procedure E-O-05, Response to Natural Events, during a high wind advisory. Specifically, on February 22, 2007, during the advisory, the inspectors identified several items stored outdoors near the plant main output transformer that could become missile hazards during actual high winds. As part of corrective actions, the licensee removed the items. The issue was entered into the licensees corrective action program.
The inspectors determined that the finding is greater than minor because, if left uncorrected, the loose items could become a more significant safety concern by allowing the accumulation of missile hazards in these areas, thereby increasing the likelihood of an initiating event. The inspectors determined that the finding warranted evaluation using the Significance Determination Process (SDP) because the finding was associated with an increase in the likelihood of an initiating event. The finding was determined to be of very low safety significance (Green). This finding has a cross-cutting aspect in the area of human performance because the licensee failed to communicate decisions and the basis for decisions to personnel who have a need to know the information in order to perform work safely and in a timely manner.
Inspection Report# : 2007002 (pdf)
Mitigating Systems Significance: SL-IV Sep 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Update the Updated Safety Evaluation Report The inspectors identified a finding of very low safety significance for the licensee's failure to adequately update the Updated Safety Analysis Report (USAR) in accordance to 10 CFR 50.71, Maintenance of records, making of
 
reports. The licensee failed to update the USAR to fully reflect changes and analyses made in response to license amendment 184. Once identified, the licensee entered this issue into its corrective action program.
Because this issue potentially impacted the NRC's ability to perform its regulatory function, this finding was evaluated using the traditional enforcement process. The finding is greater than minor because of the failure to provide complete licensing and design basis information in the USAR could result in either the licensee making an inappropriate licensing interpretation or the NRC making an inappropriate regulatory decision based on incomplete information in the USAR. The issue is of very low safety significance based upon a Phase 2 significance determination analysis of the associated technical issue. The issue was a NCV of 10 CFR 50.71(e), which required that the USAR be updated to include the effects of all safely evaluations performed by the licensee in support of requested license amendments.
The primary cause of this violation is related to the cross-cutting area of problem identification and resolution because the extent of condition review performed for a recent and similar violation failed to identify the issue even though it was within the scope of the extent of condition review which had been performed Inspection Report# : 2007004 (pdf)
Significance: SL-IV Jun 30, 2007 Identified By: NRC Item Type: VIO Violation Failure to Perform a 10 CFR 50.59 Evaluation for Compensatory Measures Associated with a Procedure Change The inspectors identified a finding having very low safety significance and an associated Severity Level IV, Cited Violation of 10 CFR 50.59 while reviewing unresolved items URI 05000305/2006003-04, Adequacy of Compensatory Actions for Potential Turbine Missile Strike of Control Room Ventilation Cooling; and URI 05000305/2006016-01, Adequacy of 10 CFR 50.59 Screening for Procedure Change. Specifically, the licensee failed to properly interpret design and licensing basis requirements associated with protection against external events and as a result did not perform a 10 CFR 50.59 evaluation. The cause of this finding is related to the cross-cutting area of problem identification and resolution because the licensee had similar prior problems that, if effectively evaluated and resolved, could have prevented this issue. (P.1(c))
This finding was determined to be more than minor because the inspectors determined that the procedure change would have ultimately required NRC approval. The procedure changes, in the form of compensatory operator actions, adversely impacted the operation of control room recirculation system following a tornado. A Phase 1 significance determination of this finding using IMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," using the Severe Weather Screening Criteria questions was completed. Since the loss of the control room recirculation system would not result in an initiating event or degrade two or more trains of a multi-train safety system, the issue screened as Green.
Inspection Report# : 2007003 (pdf)
Significance:      May 18, 2007 Identified By: NRC Item Type: NCV NonCited Violation Procedure Non-Compliance The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to adequately implement procedure DNAP-1604, Cause Evaluation Program, and the Cause Evaluation Handbook during investigative analyses of root cause, collective significance, and apparent cause evaluations. The licensee subsequently revised several apparent cause evaluations (ACEs), such as ACE 3374 on the diesel generator B fuel rack shaft binding, and completed industry benchmarking to improve root cause evaluation and ACE investigative analysis.
This finding was associated with the Mitigating Systems Cornerstone. The finding was more than minor because, if left uncorrected, the licensees analyses of conditions adverse to quality, such as the investigation of the diesel generator B fuel rack shaft binding, as documented in ACE 3374, would not be performed at an appropriate investigative depth for cause determination. The inspectors assessed the significance of this finding as very low safety significance (Green) because the finding did not represent an actual loss of safety function of the equipment. The finding was associated with cross-cutting aspect P.1(c), in the area of problem identification and resolution, corrective action program, because the licensee failed to thoroughly analyze the sequence of events and the cause and effect relationships potentially impacting the causal determination of CAP evaluations.
 
Inspection Report# : 2007008 (pdf)
Significance:      Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation No Motor Starting Analyses for Offsite Power Supply The inspectors identified a finding having very low significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to perform motor starting studies to demonstrate that motors would successfully start when connected to the offsite power supply. Upon discovery, the licensee provided additional data and compensatory measures to justify operability.
The inspectors determined that the performance deficiency was more than minor because the lack of a formal motor starting calculations resulted in the adequacy of important aspects of the design not being demonstrated, such that further evaluation needed to be performed in order to demonstrate that the equipment could perform its safety function. Although, by the end of the inspection, the licensee was able to demonstrate operability, at the time of discovery there was reasonable doubt on the operability of motors. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
Significance:      Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Increased Cable Resistance Due to Accident Temperatures The inspectors identified a finding having very low significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to consider the effects of accident temperatures on cable resistance in voltage drop calculations. Upon discovery, the licensee performed preliminary calculations to verify operability of the circuits.
This issue was more than minor because the calculational errors had more than a minimal effect on the outcome of the calculation, considerably impacting the available margin of the system such that further evaluation needed to be performed in order to demonstrate that the equipment could perform its safety function. Although, by the end of the inspection, the licensee was able to demonstrate operability; at the time of discovery there was reasonable doubt on the operability of the circuits. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
Significance:      Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Ensure Adequate 125 Vdc Breaker Interrupting Short Circuit Current Capability The inspectors identified a finding having very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to ensure that four of the 125 VDC circuit breakers had adequate interrupting short circuit fault current capability. Upon discovery, the licensee performed a preliminary evaluation, and verified that the most likely fault would result in a lower short circuit fault current than the breakers rating.
This issue was more than minor because the failure could have affected the operability of the breaker/DC Bus and could have resulted in the loss of DC power to safe shutdown equipment in the event of short circuit faults. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
 
Significance:      Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Use Actual Minimum Voltage Value in 125Vdc Voltage Drop Calculation The inspectors identified a finding having very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to use correct design input data into the 125 VDC safeguard battery calculation. The licensee used a battery terminal voltage value of 117.49 volts for BRA-101 and 118.95 volts for BRB-101, for the first minute, and did not compensate for worse case conditions. Upon discovery, the licensee performed preliminary evaluation and verified that safe shutdown equipment have adequate voltage using the battery terminal voltage value of 113.87 volts.
This issue was more than minor because the failure to use correct design input had more than a minimal effect on the outcome of the voltage drop calculation, considerably impacting the available margin of the system such that further evaluation needed to be performed in order to demonstrate that equipment could perform its safety function.
Although, during the inspection, the licensee was able to demonstrate operability; at the time of discovery there was reasonable doubt on the operability of circuits. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
Significance:      Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Acceptance Criteria in 125 Vdc Station Battery Load Tests Procedures The inspectors identified a finding having very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings. Specifically, the licensee failed to include the acceptable minimum battery terminal voltage, during the first minute, into the acceptance criteria for battery load test procedures SP-38-102A/B Station Battery Load Test. Upon discovery, the licensee entered the issue into its corrective action program to revise the acceptance criteria of procedures SP-38-102A/B to include this requirement.
This issue was more than minor because the failure to ensure that the battery terminal voltage during the first minute battery discharge did not drop below the design input value could have affected the operability of safety related equipments in the event of a design basis accident and or station blackout conditions. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
Significance:      Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Adequate Control Voltage for 4160V Breaker's Closing Coil was not Assured The inspectors identified a finding having very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to assure that the minimum available control voltage at the 4160V breakers was adequate to energize the closing coils during all conditions. Upon discovery, the licensee performed preliminary calculation and verified operability of the emergency diesel generators 4160V breakers following loss of all AC power conditions.
This finding was more than minor because the failure to assure adequate control voltage was available to close the 4160V breakers would have affected the capability of emergency diesel generators and other safety related equipments to respond to initiating events. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
 
Significance:      Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Safeguard Battery Load Profile Did Not Include LOOP/LOCA Loads The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control having very low safety significance for the licensees failure to assure that safeguard battery loads profile was adequate to meet all USAR requirements. Specifically, the licensee failed to verify that the battery loading profile for loss of coolant accident (LOCA) coincide with loss of all AC power condition was bounded by the station blackout condition loading to ensure adequate battery sizing and testing. Upon discovery, the licensee was able to show that the charger will be available upon the start of the emergency diesel generator and will provide additional support. This issue was entered into the licensees corrective action program to revise the battery calculation to include the LOCA loads.
This finding was more than minor because the failure to include the LOCA loads in the battery sizing and testing did not ensure the capability of the battery to provide adequate DC power in accordance with USAR requirements. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
Significance:      Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Electrolytic Capacitors in Spare Safeguard Battery Charger Not Periodically Energized The inspectors identified a finding having very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action. Specifically, the licensee failed to incorporate previously identified vendor recommendation to periodically energize the spare 125 VDC safeguard battery charger for at least a half-hour every 18 months to ensure the operability of the electrolytic capacitor in the charger. The licensee has previously entered the vendor recommendation into their corrective action in 2002, however, all actions were closed but the recommendation was never implemented. Following discovery, the licensee entered the issue into its corrective action program and declared the spare charger inoperable. The primary cause of this violation was related to the cross-cutting area of problem identification and resolution because the licensee failed to take appropriate corrective actions to address a previously failed charger. Specifically, the licensee developed corrective actions which included incorporating pertinent vendor recommendation into the preventive maintenance program but closed the action without ensuring completion (P.1.d)
This issue was more than minor because the failure to periodically energize the spare charger did not ensure the operability and reliability of the spare charger when needed. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
Significance:      Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Diesel Loading Calculations Non Conservative The inspectors identified a finding having very low significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to properly account for all loads on the diesel generators. Upon discovery, the licensee provided additional data and initiated procedure changes to ensure diesels were loaded within their ratings.
The inspectors determined that the performance deficiency was more than minor because the lack of adequate diesel generator loading calculations resulted in some diesel loads not being properly accounted for, such that further evaluation needed to be performed in order to demonstrate that the equipment could perform its safety function.
Although, by the end of the inspection, the licensee was able to demonstrate operability, at the time of discovery there was reasonable doubt on the operability of equipment. The inspectors screened the finding using IMC 0609, Appendix A. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609,
 
Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
Significance:        Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation RWST Level Instruments Do Not Protect SI and RHR Pumps from Excessive Air Entrainment The inspectors identified a finding having very low significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to incorporate the results of design calculations with respect to minimum refueling water storage tank (RWST) level and transfer of suction sources into the appropriate emergency operating. Procedures allowed operators to transfer suction at 4 percent indicated level in the RWST; however, at this level, significant air entrainment may damage the pumps. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program because the licensee did not thoroughly evaluate problems such that the resolution addresses the extent of condition (P.1.c).
This issue was more than minor because the existing margin was already low and as a consequence, the large error associated with the level instrument resulted in eliminating the entire margin, and jeopardized the functionality of the pumps taking suction from the RWST due to excessive air entrainment. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations SDP Phase 1.
Inspection Report# : 2007006 (pdf)
Significance:        Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Non-Conservative Assumption Used in Service Water Flow Model Calculation The inspectors identified a finding having very low significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to appropriately account for service water strainer plugging in the service water system flow model. Upon discovery, the licensee placed this issue into their corrective action program and planned to formally revise the service water system flow model to reflect plugging of both strainers in a train.
The issue was more than minor because the error had more than a minimal effect on the outcome of the calculation, considerably impacting the available margin of the system such that further evaluation needed to be performed in order to demonstrate that the service water system could perform its safety function. The issue was of very low safety significance because the issue was a design issue confirmed to not result in a loss of operability.
Inspection Report# : 2007006 (pdf)
Significance:        Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Screen House Ventilation Damper Maintenance The inspectors identified a finding having very low significance and an associated NCV of 10 CFR Part 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, Paragraph (b)(2), for the licensees failure to scope the closing function of the screenhouse ventilation dampers into the monitoring program.
Specifically, the degraded screen-house dampers fail to close and maintain ambient temperatures > 60 &deg;F such that service water system would remain operable and available after a station blackout event with severely cold outside temperatures. Following discovery, the licensee entered the issue into its corrective action program for resolution.
This issue was more than minor because the licensee had not included the closing function of the screen-house ventilation dampers within the scope of its program for implementation of the Maintenance Rule. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
 
Significance:        Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Non Conservative Assumption Used for "B" CCW Pump Room Heat Gain Calculation The inspectors identified a finding having very low significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III. Specifically, the licensee failed to account for component cooling water (CCW) piping temperatures as high as 176&deg;F in the CCW B pump room and the impact upon the temperature in the CCW B pump room. As a result, the licensee used the non-conservative results in an operability evaluation for the auxiliary building fan coil unit (FCU). Upon discovery, the licensee placed this issue into their corrective action program, performed an immediate operability evaluation, and planned to perform a more thorough evaluation. This finding has a cross-cutting aspect in the area of human performance associated with decision making because the licensee did not use conservative assumptions. Specifically, the licensee failed to account for higher CCW piping temperatures because the licensee did not model the CCW room properly and did not use the maximum expected temperature under accident conditions when revising calculation C11156 (H.1.b).
The issue was more than minor because the error because, if left uncorrected, the finding would become a more safety significant concern. The use of a non-conservative value as a basis for operability could allow FCU performance to degrade to unacceptable levels without being detected and corrected. The issue was of very low safety significance because the issue was a design issue confirmed to not result in a loss of operability.
Inspection Report# : 2007006 (pdf)
Significance:        Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Safety Injection Pump Lube Oil Coolers Testing deficiencies The inspectors identified a finding having very low significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control. Specifically, the licensee failed to establish a testing program capable of identifying an unacceptable condition of the safety injection (SI) lube oil coolers. Upon discovery, the licensee initiated a change to the test program methodology and performed back-flushing and inspection on the two SI lube oil coolers. The licensee also assessed that as a result of the very cold temperature of the water of Lake Michigan during the inspection, the cooler was considered operable. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with self- and independent assessments because during a 2005 audit of licensing commitments, the licensee failed to identify that the commitment to perform inspection and maintenance of the SI lube oil coolers in accordance with the licensee's response to Generic Letter 89-13 was not kept (P.3.a).
This issue was more than minor because when later assessed, the licensee realized that the coolers would have failed previous tests when reevaluated performance factors were less than the acceptance criterion of 0.9. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
Significance:        Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inappropriate QA Class Components Installed in TSC Diesel Generator The inspectors identified a finding of very low safety significance (Green) and an associated Non-Cited Violation of Technical Specification 6.8, Procedures, during a review on January 27, 2007, of maintenance performed on the station blackout diesel generator. The maintenance, which was conducted to repair a cooling water leak, inappropriately replaced existing parts with commercial grade components. The inspectors determined that, in accordance with procedure GNP-01.01.01, Determination of Nuclear Safety Designed Classifications, QA [Quality Assurance] Type and EQ [Environmental Qualification] Type, the new components should have been designated as augmented quality. As part of corrective actions, the licensee revised its parts database to show the appropriate classification for parts for the diesel. The issue was entered into the corrective action program.
The inspectors determined that the finding is greater than minor 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). Specifically, the installation of parts in equipment with a lower quality designation than required potentially impacted equipment reliability. The finding was determined to be of very low safety significance (Green). This finding has a cross-cutting aspect in the area of problem identification and resolution because the licensee failed to take timely effective corrective actions for a similar prior occurrence. Barriers to prevent recurrence had not been established during supervisory reviews that granted multiple extensions to the corrective actions for the prior occurrence.
Inspection Report# : 2007002 (pdf)
Significance:      Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Nuclear Instrument Test Performed Contrary to Procedural Requirements The inspectors identified a finding of very low safety significance (Green) and an associated Non-Cited Violation of Technical Specification 6.8, Procedures, when the licensee failed, on January 8, 2007, to follow procedures for performing the monthly surveillance test on power range instrument N-42 and failed to obtain an approved procedure change as required by administrative procedures when the technicians established an alternate ground point contrary to procedural requirements. As part of corrective actions, the licensee counseled the technicians involved and discussed the event with all members of the instrument and control department. The issue was entered into the corrective action program.
The inspectors determined that the finding is greater than minor because it was associated with the procedure quality 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). Specifically, the procedure required the use of the ground associated with the related card to verify proper continuity within the circuit and the use of an alternate ground point was a substantive change to the procedure. The finding was determined to be of very low safety significance (Green). This finding has a cross-cutting aspect in the area of human performance because personnel did not follow procedures.
Inspection Report# : 2007002 (pdf)
Significance:      Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Seismic Storage requirements in Station Housekeeping Procedure The inspectors identified a finding of very low safety significance (Green) and an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, during a review of procedures related to the control and storage of material. On March 21, 2007, the inspectors identified a number of unsecured equipment carts located in the vicinity of the seismically-classified, safety-related auxiliary building special ventilation system.
The inspectors concluded that, although this was allowed by plant procedure GNP-01.31.01, Plant Cleanliness and Storage, it was a condition that potentially affected quality (safe operation of the ventilation system during a seismic event) and should not have been allowed by the procedure. As part of corrective action, the licensee properly secured the carts and evaluated other carts positioned near safety-related equipment. The issue was entered into the corrective action program.
The inspectors determined that the finding is greater than minor because it was associated with the procedure quality 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). Specifically, the procedure allowed uncontrolled storage of materials in the vicinity of the auxiliary building special ventilation system that could render the system inoperable during a seismic event. The finding was determined to be of very low safety significance (Green). This finding has a cross-cutting aspect in the area of human performance because the licensee failed to provide accurate procedures to assure the operability of safety-related equipment was maintained.
Inspection Report# : 2007002 (pdf)
 
Significance:      Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Evaluate the Potentially Inadequate Design of the Service Water System The inspectors identified a finding of very low safety significance (Green) and associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, Instruction, Procedures, and Drawings, when the licensee failed to have in place adequate procedures to preclude a common mode failure of both trains of the safety-related service water (SW) system. Specifically, adequate procedures were not established for the maintenance of the SW system to prevent corrosion and degradation of the plant equipment water (PEW) filter vessels from affecting the safety-related SW bearing water supply components. As part of corrective actions, the licensee wrote the appropriate maintenance procedures. The issue was entered into the corrective action program.
The inspectors concluded that this finding is greater than minor because it was associated with the procedure quality and equipment performance attributes 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. Specifically, the lack of appropriate procedures allowed the degradation of PEW components to cause the inoperability of two safety-related SW pumps. The finding was determined to be of very low safety significance.
This finding has a cross-cutting aspect in the area of problem identification and resolution because the licensee failed to thoroughly evaluate problems such that the resolutions addressed causes and extent of condition, as necessary.
Inspection Report# : 2007002 (pdf)
Significance:      Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Foreign Material in Containment as a Result of Inadequate Containment Closure Inspections The inspectors identified a finding of very low safety significance (Green) and an associated Non-Cited Violation of Technical Specification 6.8, Procedures, on February 28, 2007, when the licensee failed to adequately perform a containment closeout inspection to ensure that debris and foreign materials were identified and removed in accordance with plant procedures. Specifically, inspectors identified unsecured metal sheets inside containment during a walkdown. As part of corrective actions, the sheets were removed from containment. The issue was entered into the corrective action program.
This finding is greater 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, the failure to identify and remove the steel sheets from containment could have affected the availability of both trains of the residual heat removal system (the accident recirculation sump) during a loss-of-coolant accident because of increased debris generation caused by the unsecured sheets. The finding was determined to be of very low safety significance (Green). This finding has a cross-cutting aspect in the area of human performance because personnel did not follow procedures, causing a condition to exist that potentially impacted the operability of both trains of the residual heat removal system.
Inspection Report# : 2007002 (pdf)
Significance: SL-IV Mar 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Notify NRC of Licensee medical Condition Change in Accordance with 10 CFR50.74 The inspectors identified a finding of very low safety significance and an associated Severity Level IV, Non-Cited Violation of 10 CFR 50.74 for the licensees failure to notify the NRC that one of its licensed operators was taking prescribed medication for a potentially disqualifying medical condition (hypertension). After a review of the licensed operators medical status was completed by the NRCs medical review officer, a condition was added to the operators license requiring him to take the medication as prescribed. The facility licensee entered this issue in their corrective action program. They required the individual licensed operator to take the medication as prescribed and incorporated these lessons learned in their requalification training program to ensure all licensed operators are aware of the requirement to notify the NRC of changes in their medical status.
 
Because violations of 10 CFR 50.74 affect the NRCs ability to perform its regulatory function, this finding was evaluated using the traditional enforcement process. In accordance with the NRC Enforcement Policy, this finding was determined to be greater than minor because the medical condition that was not reported required a change to the operators NRC license. Because the operator was always in the presence of other licensed operators while performing licensed duties and made no operational errors while he was taking the prescribed medication before his license had been appropriately revised, NRC management has determined this issue is a Green finding, of very low safety significance. This issue is considered an NCV because it was entered into the licensees corrective action program.
This finding also has a cross-cutting aspect in the area of human performance because a standard, specifically American National Standards Institute/American Nuclear Society (ANSI/ANS) 3.4, Medical Certification and Monitoring of Personnel Requiring Operator Licenses for Nuclear Power Plants, was available but not correctly implemented. The correct implementation of the standard would have led to a proper notification of the NRC and timely conditioning of the operators NRC license.
Inspection Report# : 2007002 (pdf)
Significance:        Jan 31, 2007 Identified By: NRC Item Type: VIO Violation Failure to Evaluate Operability of the "A" EDG when a Fuel Oil Leak was Identified A finding that was preliminarily determined to be of substantial safety significance (Yellow), and an associated apparent violation of Technical Specification 6.8, Procedures, was identified for a fuel oil leak on the A emergency diesel generator (EDG) that was identified on June 28, 2006, but was not repaired until 51 days later on August 17. In December 2006, the licensee tested the fitting and copper tubing that was the source of the leak to assess the leaks effect on the operability of the diesel. The licensee concluded that the leak rendered the diesel inoperable for those 51 days. As part of corrective action, the licensee replaced the leaking fuel oil line and reinforced with plant personnel the procedural requirements to properly evaluate equipment problems. The licensee also entered the issue into its corrective action program.
The finding was more than minor because if left uncorrected it would become a more significant safety concern during use of the A EDG to mitigate a loss of offsite power event. Specifically, the A EDG would have failed after approximately four hours due to the loss of fuel oil through the failed fuel line tubing, and the systems that respond to accidents and are powered by the A EDG would not be available. A Significance Determination Process Phase 3 risk analysis preliminarily determined that this finding was of substantial safety significance (Yellow). This finding has a cross-cutting aspect in the area of human performance because procedures were available, but not followed, that could have resulted in the leak being promptly repaired.
After considering the information developed during the inspection, the NRC has concluded that the inspection finding is appropriately characterized as Yellow, i.e., an issue with substantial safety significance that will result in additional NRC inspection and potentially other NRC action.
Inspection Report# : 2007007 (pdf)
Inspection Report# : 2007009 (pdf)
Barrier Integrity Significance:        Sep 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Preconditioning of Safety-Related Motor-Operated Valves Prior to Performance of Technical Specification Required Surveillance Testing The inspectors identified a finding of very low safety significance and an associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, during plant preparations to perform Surveillance Procedure SP-23-100B, Train B Containment Spray Pump and Valve Test - IST. Specifically, the inspectors noted on August 8, 2007, that shortly prior to performing the surveillance procedure, the plant had hung safety tags on the containment spray system in order to perform repair activities on IDS-102, a check valve in that
 
system. These tags required that normally open motor- operated valves IDS-202 and IDS-2B be cycled closed and tagged in order to isolate the check valve. Because these motor-operated valves were required to be stroke and time-tested during the performance of the surveillance procedure, and the effects of preconditioning on these valves was not considered prior to implementation of the maintenance activity, the inspectors determined that plant procedures were inadequate to assess preconditioning implications associated with station activities. The licensee entered the issue into their corrective action program. Corrective actions included completion of the surveillance procedure with acceptable results and a evaluation of the test results, which determined that the surveillance test was acceptable.
The finding is greater than minor because it was associated with the configuration control attribute of the Barrier Integrity Cornerstone and 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. The inspectors evaluated the finding using IMC 0609, Appendix A, Significance Determination Process, and determined that this finding is of very low safety significance by answering No to all questions in the containment barriers cornerstone column. The inspectors also determined that the primary cause for this finding is related to the cross-cutting area of human performance. Specifically, under the component of resources, procedures to assess and prevent preconditioning of safety-related components were not complete, accurate, and up-to-date Inspection Report# : 2007004 (pdf)
Significance:        Jun 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedure for Surveillance Testing of Auxiliary Building Special Ventilation Zone The inspectors identified a finding having very low safety significance and an associated non-cited Violation of 10 CFR 50, Appendix B, Criterion V, Instructions , Procedures, and Drawings,while reviewing surveillance testing procedures for the auxiliary building special ventilation zone (Zone SV). Specifically, the licensee procedure for tracking the amount of in-leakage into the Zone SV did not have adequate criteria to capture degraded conditions, nor ensure that the acceptance criteria reflected the design requirements of the system. The cause of this finding is related to the cross-cutting area of problem identification and resolution because the licensee failed to properly evaluate multiple condition reports for operability and extent of condition. (P1(c))
This finding was determined to be more than minor because, if left uncorrected, the failure to evaluate barrier breaches that did not have breach permits could become a more significant safety concern. Specifically, if left unmonitored the breaches without barrier permits could potentially exceed the allowable design limits. The finding was evaluated using IMC 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations. The answer to Question 1 in the Significance Determination Process Phase 1 Screening Worksheet in the Containment Barrier Cornerstone column was yes; therefore, this finding is of very low safety significance (Green). Corrective actions to date included revisions to procedure FPP-08-09, to track barrier breaches that result from degraded conditions and provide conservative acceptance criteria.
Inspection Report# : 2007003 (pdf)
Significance:        Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Loss of Coolant Environment Improperly Considered in Containment Fan Coil Unit Calculation The inspectors identified a finding having very low significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, "Design Control." Specifically, the licensee failed to use the correct data when determining the most limiting conditions on the safety related motors of the containment fan coil units (CFCU). The engineers failed to use the combination of the greatest density of the air-steam mixture following a loss of coolant accident (LOCA) with the greatest flow rate attributed to the fans by testing. As a result, the licensee was not aware that post LOCA, the motors will be operating at 113 percent of their design rating, and drawing 13 additional kW from each diesel generator.
Upon discovery, the licensee recalculated the motors' horsepower, recalculated the service factor (percent above continuous design rating) at which the motors will be operating, and recalculated the elevated current that will be drawn by the motors, and the elevated current at degraded voltage. In addition, the licensee had to reevaluate whether the over-current trip setpoint of the motors will be exceeded.
 
This issue was more than minor because the assumed power drawn by the motors was significantly less, the existing margin was already low, and as a consequence, the error resulted in a significant reduction in margin. This issue also impacted the capability of the emergency diesel generators to supply the required power to the CFCU's motors. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
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.
Miscellaneous Last modified : February 04, 2008
 
Kewaunee 1Q/2008 Plant Inspection Findings Initiating Events Significance:      Sep 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement Maintenance Rule (a)(1) in Corrective Actions on the "G" Instrument Air Compressor The inspectors identified a finding of very low safety significance and an associated non-cited violation of 10 CFR 50.65(a)(1), Requirements for monitoring the effectiveness of maintenance at nuclear power plants. Specifically, as of August 25, 2007, the licensee failed to implement the Maintenance Rule (a)(1) action plan which had been incorporated into plant procedure N-AS-01 to preclude a loss of the G air compressor. The licensee entered the issue into their corrective action program. Corrective actions have included implementation of the procedural requirements of N-AS-01 for both the G and F air compressors.
The finding is greater than minor because it relates to a licensee failure to implement prescribed significant compensatory measures to manage risk and implement the 10 CFR 50.65(a)(1) action plan. Additionally, the finding is 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. The inspectors evaluated the finding using Inspection Manual Chapter 0609, Appendix A, Significance Determination Process, and determined that this finding is of very low safety significance by answering No to all questions in the Initiating Events Cornerstone column.
Inspection Report# : 2007004 (pdf)
Significance:      Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation No Analysis or Procedures to Establish Operability of the Tertiary Auxiliary Transformer Source The inspectors identified a finding having very low significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to evaluate the capability of the 345 kV offsite power supply when isolated from the 138 kV switchyard and to translate this criteria into procedures.
This issue was more than minor because procedures allowed operation of the station in unanalyzed configurations for which operability of one offsite source could not be assured and new calculations were needed to ensure that the design basis was met. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
Significance:      Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Incorporate and Effectively Implement Operating Experience in Reactor Trip Breaker Maintenance Activities The inspectors identified a finding having very low significance and an associated NCV of 10 CFR 50.65(a)(3) for the failure to incorporate external and internal operating experience into preventive maintenance activities for the reactor trip breakers. 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 previous breaker issues and did not perform adequate extent of condition reviews. Specifically, the licensee initiated several corrective action documents in response to identified issues; however, did not perform adequate evaluations of the conditions to address the cause or resolve the identified issue. (P.1.(c))
 
This issue was more than minor because the licensee failed to ensure that the reactor trip breakers (RTBs), and their associated cell assemblies, had been maintained in a continuous state of operational readiness by performing effective maintenance and surveillance activities in accordance with relevant vendor specifications and available operating experience. The issue was of very low safety significance based on a Phase 1 screening because the finding did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available.
Inspection Report# : 2007006 (pdf)
Significance:        Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Acceptance Criteria Not Met Due to Failure to Follow Procedure The inspectors identified a finding having very low significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings. Specifically, on May 22, 2006 during the performance of PMP-47-01, maintenance technician recorded a trip bar force of 32 ounces when testing reactor trip breaker (RTB) serial number 850-027-1, which exceeded the acceptance criteria; however, no further actions were taken as required by the test. This finding has a cross-cutting aspect in the area of Human Performance, Work Practices because the licensee did not perform an adequate peer check of the surveillance results. Specifically, several individuals including the person performing the task did not identify that the RTB trip bar force exceeded the acceptance criteria. (H.4.(c))
This issue was more than minor because not meeting the acceptance for the trip bar force impacted the reliability of the RBTs because excessive force could result in a failure to trip the breaker. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
Mitigating Systems Significance:        Mar 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Scaffolding in Close Proximity to Multiple Safety-Related Systems Affects Operability A finding of very low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the licensees failure to install scaffolding in accordance with station procedures. Specifically, more than ten examples where scaffolding was built within 2-inches of safety-related systems without an engineering evaluation, and six examples where non-seismic scaffolding was built in safety-related areas were identified. The licensee suspended all scaffold building pending the completion of their corrective actions. The corrective actions included training scaffold builders on proper scaffold building techniques and how to identify operational and seismic concerns, revising procedures for scaffold building to address operations and engineering involvement in the scaffold building process, and a complete plant walkdown of all scaffolding by engineering or operations.
This finding was more than minor because it was 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. Specifically, the improperly installed scaffolding could have impeded or prevented proper operation of the safety-related components.
Using Attachment 4 of IMC 0609, the inspectors answered no to all the screening questions in the SDP Phase 1 Screening Worksheet in the Mitigating Systems column; therefore, this finding is of very low safety significance (Green). The inspectors determined that this finding had a cross cutting aspect in the area of problem identification and resolution, corrective action program, because the licensee did not take appropriate corrective actions to address safety issues and adverse trends in a timely manner.
Inspection Report# : 2008002 (pdf)
 
Significance:      Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Characterize and Manage Risk in Accordance with Maintenance Rule With the Turbine-Driven Auxiliary Feedwater Pump in Pull-to-Lock A finding of very low safety significance and an associated Non Cited Violation (NCV) of 10 CFR 50.65(a)(4),
Requirements for monitoring the effectiveness of maintenance at nuclear power plants, was identified by the inspectors during startup of the reactor following a plant shutdown to replace leaking hydrogen coolers on the turbine generator. The licensee entered this issue into its corrective action program.
The finding is greater than minor because if left uncorrected the finding would become a more significant safety concern. Specifically, the licensee failed to correctly characterize the risk on October 12 13, 2007. The inspectors evaluated the finding using Appendix K of Inspection Manual Chapter 0609, Maintenance Risk Assessment and Risk Management Significance Determination Process, and determined that this finding was of very low safety significance in accordance with Flowchart 1, Assessment of Risk Deficit.
Inspection Report# : 2007005 (pdf)
Significance:      Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Solenoid Valve Not Installed Properly A finding of very low safety significance and an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed when the B turbine building service water header isolation valve failed to fully cycle on demand. Specifically, the licensee failed to provide adequate procedures to support installation and maintenance for certain designs of solenoid valves used in the instrument air system. A failure of an instrument air system solenoid valve caused the service water valve to fail. The licensee repaired the solenoid valve and entered the issue into its corrective action program.
The finding is greater than minor because it is 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). Specifically, the licensee failed to provide adequate procedures to support the installation, maintenance, and operation of safety-related solenoid valve, SV-33044. Using Appendix A of Inspection Manual Chapter 0609, the inspectors answered no to all the screening questions in the Significance Determination Process Phase 1 Screening Worksheet in the Mitigating Systems column; therefore, this finding is of very low safety significance. This finding has a cross-cutting aspect in the area of human performance because related installation and maintenance procedures (resources) were inadequate and not up to date to ensure safety-related equipment was protected (H.2(c)).
Inspection Report# : 2007005 (pdf)
Significance:      Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedures for Post-Maintenance Testing of Steam Traps in the Turbine-Driven Auxiliary Feedwater Pump System A finding of very low safety significance and an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors during observation of a post-maintenance test on a steam trap associated with the turbine-driven auxiliary feedwater pump (TDAFWP) in accordance with plant procedure CMP-13-01 TD-Turbine Room Traps and Drains-Trap Maintenance. Specifically, the licensee failed to provide adequate procedures to support the testing of the steam trap. The licensee has entered the issue into the corrective action program and will be revising the appropriate procedures.
The finding is greater than minor because the finding is 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).
 
Specifically, the licensee failed to provide adequate procedures to support the post maintenance testing of steam traps in the TDAFWP system. Using Appendix A of Inspection Manual Chapter 0609, the inspectors answered no to all the screening questions in the Significance Determination Process Phase 1 Screening Worksheet in the Mitigating Systems column; therefore, this finding is of very low safety significance. This finding has a cross-cutting aspect in the area of human performance because the licensee had ample opportunity (resources) available to update procedure CMP-13-01 during multiple prior maintenance activities but did not (H.2(c)).
Inspection Report# : 2007005 (pdf)
Significance:        Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Seismic Storage Procedure A finding of very low safety significance and an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors during a review of procedures related to the control and storage of material. Specifically, procedure GNP 01.31.01, Plant Cleanliness and Storage, permitted uncontrolled storage of materials next to a Seismic Class 1 system. Additionally, opportunities existed to correct it and/or place compensatory measures in place after the NRC issued an NCV related to this issue in the first quarter of 2007. The licensee has entered the issue into the corrective action program and will be revising the procedure.
The finding is greater than minor because it is associated with the procedure quality 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). Specifically, the procedure allowed uncontrolled storage of materials in the vicinity of Seismic Class 1 systems that could render the systems inoperable during a seismic event. Using Appendix A of Inspection Manual Chapter 0609, the inspectors answered no to all the screening questions in the Significance Determination Process Phase 1 Screening Worksheet in the Mitigating Systems column; therefore, this finding is of very low safety significance. This finding has a cross-cutting aspect in the area of problem identification and resolution because the licensee failed to take appropriate timely corrective actions or put compensatory actions in place after the NRC issued a Non-Cited Violation relating to this issue in the first quarter of 2007 (P.1(d)).
Inspection Report# : 2007005 (pdf)
Significance:        Dec 19, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Extent of Condition Review for Fuel Leak.
The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, of very low safety significance, for failure by the licensee to follow procedural requirements for performing an adequate extent of condition for a diesel fuel line failure in 2006. Specifically, the licensee failed to complete an extent of condition which would have evaluated different systems where a similar failure mechanism (cyclic fatigue) could occur. The licensee entered the item into their corrective action program.
The issue is greater than minor because it was associated with the equipment performance attribute of the Mitigating System Cornerstone and affected the cornerstone objective of ensuring the availability of systems that respond to initiating events to prevent undesirable consequences. Specifically, it affected the equipment performance attribute for availability and reliability. Using Inspection Manual Chapter 0609, Significance Determination Process, the inspectors screened this issue as of very low safety significance (Green) because no loss of safety function occurred.
Inspection Report# : 2007011 (pdf)
Significance:        Dec 19, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Initiate Corrective Action Documents for Multiple Leaks in the Plant The inspections identified a non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, of very low safety significance. Specifically, the licensee failed to initiate corrective action documents
 
in accordance with plant procedures for multiple leaks found in the plant. The licensee entered this item into its corrective action program.
The finding is greater than minor because it was associated with the equipment performance attribute of the Mitigating System Cornerstone and affected the cornerstone objective of ensuring the availability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to identify and correct leakage on equipment important to safety could eventually lead to equipment unavailability during events that the equipment is designed to mitigate. The finding is of very low safety significance (Green), because the inspectors answered no to all five questions under the Mitigating Systems Cornerstone column of the Phase 1 worksheet in Inspection Manual Chapter 0609, Significance Determination Process. Specifically, at the time that the leakage was discovered, none of the leaks immediately impacted the functionality of the equipment affected. The finding has a cross-cutting aspect in the area of human performance because the licensee failed to effectively communicate expectations regarding procedural compliance for the corrective action program.
Inspection Report# : 2007011 (pdf)
Significance:        Dec 19, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Update the Updated Safety Analysis Report with Safety Analysis for Pressure Locking of Containment Sump Isolation Valves The inspectors identified a non cited violation of 10 CFR 50.71, of very low safety significance, for the licensees failure to update the Updated Safety Analysis Report (USAR). Specifically, the licensee failed to update the USAR to fully reflect the results of a safety analysis performed in response to Generic Letter 95 07, Pressure Locking and Thermal Binding of Safety Related Power Operated Gate Valves. The licensee entered this issue into its corrective action program.
Because this finding potentially impacted the NRCs ability to perform its regulatory function, it was evaluated using the traditional enforcement process. The finding is greater than minor because the failure to provide complete licensing and design basis information in the USAR could result in either the licensee making an inappropriate licensing interpretation or the NRC making an inappropriate regulatory decision based on incomplete information in the USAR. NRC management determined that this issue is of very low safety significance (Green) because it is a design issue confirmed not to result in a loss of operability.
Inspection Report# : 2007011 (pdf)
Significance:        Dec 19, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Extent of Condition Review of BF-66 Relays The inspectors identified a non cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, of very low safety significance, for failure by the licensee to follow procedural requirements for performing an adequate extent of condition following relay failures that led to reactor trips in 2006 and 2007.
Specifically, the licensee failed to perform an extent of condition action to inspect Engineered Safety Feature (ESF) relays when sufficient causal evidence was present that the same style relay in the ESF system (BF 66 relays) were susceptible to sulfidation, installation deficiencies, or manufacturing defects. The licensee entered this issue into its corrective action program.
The issue is greater than minor because, if left uncorrected, the failure to assess the other systems would become a more significant safety concern. Using Inspection Manual Chapter 0609, Significance Determination Process, the inspectors screened this issue as being of very low safety significance (Green) because no loss of safety function occurred.
Inspection Report# : 2007011 (pdf)
Significance: SL-IV Sep 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Update the Updated Safety Evaluation Report
 
The inspectors identified a finding of very low safety significance for the licensee's failure to adequately update the Updated Safety Analysis Report (USAR) in accordance to 10 CFR 50.71, Maintenance of records, making of reports. The licensee failed to update the USAR to fully reflect changes and analyses made in response to license amendment 184. Once identified, the licensee entered this issue into its corrective action program.
Because this issue potentially impacted the NRC's ability to perform its regulatory function, this finding was evaluated using the traditional enforcement process. The finding is greater than minor because of the failure to provide complete licensing and design basis information in the USAR could result in either the licensee making an inappropriate licensing interpretation or the NRC making an inappropriate regulatory decision based on incomplete information in the USAR. The issue is of very low safety significance based upon a Phase 2 significance determination analysis of the associated technical issue. The issue was a NCV of 10 CFR 50.71(e), which required that the USAR be updated to include the effects of all safely evaluations performed by the licensee in support of requested license amendments.
The primary cause of this violation is related to the cross-cutting area of problem identification and resolution because the extent of condition review performed for a recent and similar violation failed to identify the issue even though it was within the scope of the extent of condition review which had been performed Inspection Report# : 2007004 (pdf)
Significance: SL-IV Jun 30, 2007 Identified By: NRC Item Type: VIO Violation Failure to Perform a 10 CFR 50.59 Evaluation for Compensatory Measures Associated with a Procedure Change The inspectors identified a finding having very low safety significance and an associated Severity Level IV, Cited Violation of 10 CFR 50.59 while reviewing unresolved items URI 05000305/2006003-04, Adequacy of Compensatory Actions for Potential Turbine Missile Strike of Control Room Ventilation Cooling; and URI 05000305/2006016-01, Adequacy of 10 CFR 50.59 Screening for Procedure Change. Specifically, the licensee failed to properly interpret design and licensing basis requirements associated with protection against external events and as a result did not perform a 10 CFR 50.59 evaluation. The cause of this finding is related to the cross-cutting area of problem identification and resolution because the licensee had similar prior problems that, if effectively evaluated and resolved, could have prevented this issue. (P.1(c))
This finding was determined to be more than minor because the inspectors determined that the procedure change would have ultimately required NRC approval. The procedure changes, in the form of compensatory operator actions, adversely impacted the operation of control room recirculation system following a tornado. A Phase 1 significance determination of this finding using IMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," using the Severe Weather Screening Criteria questions was completed. Since the loss of the control room recirculation system would not result in an initiating event or degrade two or more trains of a multi-train safety system, the issue screened as Green.
Inspection Report# : 2007003 (pdf)
Significance:      May 18, 2007 Identified By: NRC Item Type: NCV NonCited Violation Procedure Non-Compliance The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to adequately implement procedure DNAP-1604, Cause Evaluation Program, and the Cause Evaluation Handbook during investigative analyses of root cause, collective significance, and apparent cause evaluations. The licensee subsequently revised several apparent cause evaluations (ACEs), such as ACE 3374 on the diesel generator B fuel rack shaft binding, and completed industry benchmarking to improve root cause evaluation and ACE investigative analysis.
This finding was associated with the Mitigating Systems Cornerstone. The finding was more than minor because, if left uncorrected, the licensees analyses of conditions adverse to quality, such as the investigation of the diesel generator B fuel rack shaft binding, as documented in ACE 3374, would not be performed at an appropriate investigative depth for cause determination. The inspectors assessed the significance of this finding as very low safety significance (Green) because the finding did not represent an actual loss of safety function of the equipment. The finding was associated with cross-cutting aspect P.1(c), in the area of problem identification and resolution, corrective action program, because the licensee failed to thoroughly analyze the sequence of events and the cause and effect
 
relationships potentially impacting the causal determination of CAP evaluations.
Inspection Report# : 2007008 (pdf)
Significance:      Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation No Motor Starting Analyses for Offsite Power Supply The inspectors identified a finding having very low significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to perform motor starting studies to demonstrate that motors would successfully start when connected to the offsite power supply. Upon discovery, the licensee provided additional data and compensatory measures to justify operability.
The inspectors determined that the performance deficiency was more than minor because the lack of a formal motor starting calculations resulted in the adequacy of important aspects of the design not being demonstrated, such that further evaluation needed to be performed in order to demonstrate that the equipment could perform its safety function. Although, by the end of the inspection, the licensee was able to demonstrate operability, at the time of discovery there was reasonable doubt on the operability of motors. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
Significance:      Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Increased Cable Resistance Due to Accident Temperatures The inspectors identified a finding having very low significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to consider the effects of accident temperatures on cable resistance in voltage drop calculations. Upon discovery, the licensee performed preliminary calculations to verify operability of the circuits.
This issue was more than minor because the calculational errors had more than a minimal effect on the outcome of the calculation, considerably impacting the available margin of the system such that further evaluation needed to be performed in order to demonstrate that the equipment could perform its safety function. Although, by the end of the inspection, the licensee was able to demonstrate operability; at the time of discovery there was reasonable doubt on the operability of the circuits. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
Significance:      Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Ensure Adequate 125-Volt Direct Current Breaker Interrupting Short Circuit Current Capability The inspectors identified a finding having very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to ensure that four of the 125-Volt Direct Current (DC) circuit breakers had adequate interrupting short circuit fault current capability. Upon discovery, the licensee performed a preliminary evaluation, and verified that the most likely fault would result in a lower short circuit fault current than the breakers rating.
This issue was more than minor because the failure could have affected the operability of the breaker/DC Bus and could have resulted in the loss of DC power to safe shutdown equipment in the event of short circuit faults. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
 
Significance:      Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Use Actual Minimum Voltage Value in 125-Volt Direct Current Voltage Drop Calculation The inspectors identified a finding having very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to use correct design input data into the 125-Volt Direct Current safeguard battery calculation. The licensee used a battery terminal voltage value of 117.49 volts for BRA-101 and 118.95 volts for BRB-101, for the first minute, and did not compensate for worse case conditions. Upon discovery, the licensee performed preliminary evaluation and verified that safe shutdown equipment have adequate voltage using the battery terminal voltage value of 113.87 volts.
This issue was more than minor because the failure to use correct design input had more than a minimal effect on the outcome of the voltage drop calculation, considerably impacting the available margin of the system such that further evaluation needed to be performed in order to demonstrate that equipment could perform its safety function.
Although, during the inspection, the licensee was able to demonstrate operability; at the time of discovery there was reasonable doubt on the operability of circuits. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
Significance:      Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Acceptance Criteria in 125-Volt Direct Current Station Battery Load Tests Procedures The inspectors identified a finding having very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings. Specifically, the licensee failed to include the acceptable minimum battery terminal voltage, during the first minute, into the acceptance criteria for battery load test procedures SP-38-102A/B Station Battery Load Test. Upon discovery, the licensee entered the issue into its corrective action program to revise the acceptance criteria of procedures SP-38-102A/B to include this requirement.
This issue was more than minor because the failure to ensure that the battery terminal voltage during the first minute battery discharge did not drop below the design input value could have affected the operability of safety related equipments in the event of a design basis accident and or station blackout conditions. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
Significance:      Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Adequate Control Voltage for 4160-Volt Breaker's Closing Coil Was Not Assured The inspectors identified a finding having very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to assure that the minimum available control voltage at the 4160-Volt breakers was adequate to energize the closing coils during all conditions. Upon discovery, the licensee performed preliminary calculation and verified operability of the emergency diesel generators 4160-Volt breakers following loss of all alternating current power conditions.
This finding was more than minor because the failure to assure adequate control voltage was available to close the 4160-Volt breakers would have affected the capability of emergency diesel generators and other safety related equipments to respond to initiating events. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
Significance:      Apr 17, 2007
 
Identified By: NRC Item Type: NCV NonCited Violation Safeguard Battery Load Profile Did Not Include Loss of Offsite Power/Loss of Coolant Accident Loads The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control having very low safety significance for the licensees failure to assure that safeguard battery loads profile was adequate to meet all Updated Safety Analysis Report (USAR) requirements. Specifically, the licensee failed to verify that the battery loading profile for loss of coolant accident (LOCA) coincide with loss of all alternating current power condition (loss of offsite power) was bounded by the station blackout condition loading to ensure adequate battery sizing and testing.
Upon discovery, the licensee was able to show that the charger will be available upon the start of the emergency diesel generator and will provide additional support. This issue was entered into the licensees corrective action program to revise the battery calculation to include the LOCA loads.
This finding was more than minor because the failure to include the LOCA loads in the battery sizing and testing did not ensure the capability of the battery to provide adequate direct current power in accordance with USAR requirements. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
Significance:      Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Electrolytic Capacitors in Spare Safeguard Battery Charger Not Periodically Energized The inspectors identified a finding having very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action. Specifically, the licensee failed to incorporate previously identified vendor recommendation to periodically energize the spare 125-volt direct current safeguard battery charger for at least a half-hour every 18 months to ensure the operability of the electrolytic capacitor in the charger. The licensee has previously entered the vendor recommendation into their corrective action in 2002, however, all actions were closed but the recommendation was never implemented. Following discovery, the licensee entered the issue into its corrective action program and declared the spare charger inoperable. The primary cause of this violation was related to the cross-cutting area of problem identification and resolution because the licensee failed to take appropriate corrective actions to address a previously failed charger. Specifically, the licensee developed corrective actions which included incorporating pertinent vendor recommendation into the preventive maintenance program but closed the action without ensuring completion (P.1.d)
This issue was more than minor because the failure to periodically energize the spare charger did not ensure the operability and reliability of the spare charger when needed. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
Significance:      Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Diesel Loading Calculations Non-Conservative The inspectors identified a finding having very low significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to properly account for all loads on the diesel generators. Upon discovery, the licensee provided additional data and initiated procedure changes to ensure diesels were loaded within their ratings.
The inspectors determined that the performance deficiency was more than minor because the lack of adequate diesel generator loading calculations resulted in some diesel loads not being properly accounted for, such that further evaluation needed to be performed in order to demonstrate that the equipment could perform its safety function.
Although, by the end of the inspection, the licensee was able to demonstrate operability, at the time of discovery there was reasonable doubt on the operability of equipment. The inspectors screened the finding using IMC 0609, Appendix A. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
 
Significance:        Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation RWST Level Instruments Do Not Protect SI and RHR Pumps from Excessive Air Entrainment The inspectors identified a finding having very low significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to incorporate the results of design calculations with respect to minimum refueling water storage tank (RWST) level and transfer of suction sources into the appropriate emergency operating. Procedures allowed operators to transfer suction at 4 percent indicated level in the RWST; however, at this level, significant air entrainment may damage the pumps. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program because the licensee did not thoroughly evaluate problems such that the resolution addresses the extent of condition (P.1.c).
This issue was more than minor because the existing margin was already low and as a consequence, the large error associated with the level instrument resulted in eliminating the entire margin, and jeopardized the functionality of the pumps taking suction from the RWST due to excessive air entrainment. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations SDP Phase 1.
Inspection Report# : 2007006 (pdf)
Significance:        Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Non-Conservative Assumption Used in Service Water Flow Model Calculation The inspectors identified a finding having very low significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to appropriately account for service water strainer plugging in the service water system flow model. Upon discovery, the licensee placed this issue into their corrective action program and planned to formally revise the service water system flow model to reflect plugging of both strainers in a train.
The issue was more than minor because the error had more than a minimal effect on the outcome of the calculation, considerably impacting the available margin of the system such that further evaluation needed to be performed in order to demonstrate that the service water system could perform its safety function. The issue was of very low safety significance because the issue was a design issue confirmed to not result in a loss of operability.
Inspection Report# : 2007006 (pdf)
Significance:        Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Screen House Ventilation Damper Maintenance The inspectors identified a finding having very low significance and an associated NCV of 10 CFR Part 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, Paragraph (b)(2), for the licensees failure to scope the closing function of the screenhouse ventilation dampers into the monitoring program.
Specifically, the degraded screen-house dampers fail to close and maintain ambient temperatures > 60 &deg;F such that service water system would remain operable and available after a station blackout event with severely cold outside temperatures. Following discovery, the licensee entered the issue into its corrective action program for resolution.
This issue was more than minor because the licensee had not included the closing function of the screen-house ventilation dampers within the scope of its program for implementation of the Maintenance Rule. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
Significance:        Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation
 
Non-Conservative Assumption Used for "B" Component Cooling Water Pump Room Heat Gain Calculation The inspectors identified a finding having very low significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III. Specifically, the licensee failed to account for component cooling water (CCW) piping temperatures as high as 176&deg;F in the CCW B pump room and the impact upon the temperature in the CCW B pump room. As a result, the licensee used the non-conservative results in an operability evaluation for the auxiliary building fan coil unit (FCU). Upon discovery, the licensee placed this issue into their corrective action program, performed an immediate operability evaluation, and planned to perform a more thorough evaluation. This finding has a cross-cutting aspect in the area of human performance associated with decision making because the licensee did not use conservative assumptions. Specifically, the licensee failed to account for higher CCW piping temperatures because the licensee did not model the CCW room properly and did not use the maximum expected temperature under accident conditions when revising calculation C11156 (H.1.b).
The issue was more than minor because the error because, if left uncorrected, the finding would become a more safety significant concern. The use of a non-conservative value as a basis for operability could allow FCU performance to degrade to unacceptable levels without being detected and corrected. The issue was of very low safety significance because the issue was a design issue confirmed to not result in a loss of operability.
Inspection Report# : 2007006 (pdf)
Significance:        Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Safety Injection Pump Lube Oil Coolers Testing Deficiencies The inspectors identified a finding having very low significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control. Specifically, the licensee failed to establish a testing program capable of identifying an unacceptable condition of the safety injection (SI) lube oil coolers. Upon discovery, the licensee initiated a change to the test program methodology and performed back-flushing and inspection on the two SI lube oil coolers. The licensee also assessed that as a result of the very cold temperature of the water of Lake Michigan during the inspection, the cooler was considered operable. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with self- and independent assessments because during a 2005 audit of licensing commitments, the licensee failed to identify that the commitment to perform inspection and maintenance of the SI lube oil coolers in accordance with the licensee's response to Generic Letter 89-13 was not kept (P.3.a).
This issue was more than minor because when later assessed, the licensee realized that the coolers would have failed previous tests when reevaluated performance factors were less than the acceptance criterion of 0.9. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
Barrier Integrity Significance:        Mar 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow the Provisions of Corrective Action Procedure PI-KW-200 Following Surveillance Testing of Containment Isolation Valve LOCA-31 A finding of very low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors following surveillance testing of containment isolation valve LOCA 3A in accordance with plant procedure SP 55 167 4B, "Post LOCA Valves Timing Test (IST) from Local Panel-Train B." Specifically, the licensee failed to initiate a condition report in accordance with procedure PI-KW-200, Corrective Action, following a review of the test results by the inservice testing program engineer who subsequently identified a potential condition which called into question the operability of LOCA-3A.
The finding was more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 20, 2007, because the finding was associated with the structure, system and
 
component (SSC) and barrier performance attribute of the Barrier Integrity Cornerstone and affected the cornerstone objective to provide reasonable assurance that the physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the licensee failed to implement the provisions of Corrective Action Procedure, PI-KW-200, which resulted in a failure to ensure operability of containment isolation valve LOCA-3A. The inspectors also determined that the primary cause for this finding was related to the cross cutting area of human performance, work practices, because personnel have been trained in need for procedural use and adherence but did not follow applicable procedures.
Inspection Report# : 2008002 (pdf)
Significance:        Sep 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Preconditioning of Safety-Related Motor-Operated Valves Prior to Performance of Technical Specification Required Surveillance Testing The inspectors identified a finding of very low safety significance and an associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, during plant preparations to perform Surveillance Procedure SP-23-100B, Train B Containment Spray Pump and Valve Test - IST. Specifically, the inspectors noted on August 8, 2007, that shortly prior to performing the surveillance procedure, the plant had hung safety tags on the containment spray system in order to perform repair activities on IDS-102, a check valve in that system. These tags required that normally open motor- operated valves IDS-202 and IDS-2B be cycled closed and tagged in order to isolate the check valve. Because these motor-operated valves were required to be stroke and time-tested during the performance of the surveillance procedure, and the effects of preconditioning on these valves was not considered prior to implementation of the maintenance activity, the inspectors determined that plant procedures were inadequate to assess preconditioning implications associated with station activities. The licensee entered the issue into their corrective action program. Corrective actions included completion of the surveillance procedure with acceptable results and a evaluation of the test results, which determined that the surveillance test was acceptable.
The finding is greater than minor because it was associated with the configuration control attribute of the Barrier Integrity Cornerstone and 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. The inspectors evaluated the finding using IMC 0609, Appendix A, Significance Determination Process, and determined that this finding is of very low safety significance by answering No to all questions in the containment barriers cornerstone column. The inspectors also determined that the primary cause for this finding is related to the cross-cutting area of human performance. Specifically, under the component of resources, procedures to assess and prevent preconditioning of safety-related components were not complete, accurate, and up-to-date Inspection Report# : 2007004 (pdf)
Significance:        Jun 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedure for Surveillance Testing of Auxiliary Building Special Ventilation Zone The inspectors identified a finding having very low safety significance and an associated non-cited Violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, while reviewing surveillance testing procedures for the auxiliary building special ventilation zone (Zone SV). Specifically, the licensee procedure for tracking the amount of in-leakage into the Zone SV did not have adequate criteria to capture degraded conditions, nor ensure that the acceptance criteria reflected the design requirements of the system. The cause of this finding is related to the cross-cutting area of problem identification and resolution because the licensee failed to properly evaluate multiple condition reports for operability and extent of condition. (P1(c))
This finding was determined to be more than minor because, if left uncorrected, the failure to evaluate barrier breaches that did not have breach permits could become a more significant safety concern. Specifically, if left unmonitored the breaches without barrier permits could potentially exceed the allowable design limits. The finding was evaluated using IMC 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations. The answer to Question 1 in the Significance Determination Process Phase 1 Screening Worksheet in the Containment Barrier Cornerstone column was yes; therefore, this finding is of very low safety significance (Green). Corrective actions to date included revisions to procedure FPP-08-09, to track barrier breaches that result
 
from degraded conditions and provide conservative acceptance criteria.
Inspection Report# : 2007003 (pdf)
Significance:      Apr 17, 2007 Identified By: NRC Item Type: NCV NonCited Violation Loss of Coolant Environment Improperly Considered in Containment Fan Coil Unit Calculation The inspectors identified a finding having very low significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, "Design Control." Specifically, the licensee failed to use the correct data when determining the most limiting conditions on the safety related motors of the containment fan coil units (CFCUs). The engineers failed to use the combination of the greatest density of the air-steam mixture following a loss of coolant accident (LOCA) with the greatest flow rate attributed to the fans by testing. As a result, the licensee was not aware that post-LOCA, the motors will be operating at 113 percent of their design rating, and drawing 13 additional kilowatts from each diesel generator.
Upon discovery, the licensee recalculated the motors' horsepower, recalculated the service factor (percent above continuous design rating) at which the motors will be operating, and recalculated the elevated current that will be drawn by the motors, and the elevated current at degraded voltage. In addition, the licensee had to reevaluate whether the over-current trip setpoint of the motors will be exceeded.
This issue was more than minor because the assumed power drawn by the motors was significantly less, the existing margin was already low, and as a consequence, the error resulted in a significant reduction in margin. This issue also impacted the capability of the emergency diesel generators to supply the required power to the CFCU's motors. The issue was of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations.
Inspection Report# : 2007006 (pdf)
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.
Miscellaneous Last modified : June 05, 2008
 
Kewaunee 2Q/2008 Plant Inspection Findings Initiating Events Significance:      Jun 30, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow the Provisions of General Nuclear Procedure, GNP-12.06.01, "Hot and Cold Weather Operations."
Green. A finding of very low safety significance (Green) and an NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors following an inspection of licensee preparations for adverse weather protection. Specifically, the licensee failed to perform inspections for hot weather operations as required by plant procedure GNP-12.06.01, "Hot and Cold Weather Operations."
The finding was greater than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 20, 2007, because if left uncorrected would become a more significant safety concern. Specifically, the licensee failed to implement the provisions of GNP 12.06.01, "Hot and Cold Weather Operations," which resulted in a failure to ensure pre-summer readiness of numerous safety-related and risk-significant systems. The inspectors evaluated the finding using Attachment 0609.04, of IMC 0609, Significance Determination Process, dated January 10, 2008, and answered no to all of the questions in the Initiating Events column; therefore, the finding was determined to be of very low safety significance. The inspectors determined that the primary cause for this finding was related to the cross cutting area of human performance, work practices component, because personnel have been trained in the need for procedural use and adherence, but failed to follow applicable procedures. Specifically, the procedure which required the performance of plant inspections for hot weather operations, and the maintenance of QA documentation for these inspections, was not followed [H.4(b)]
Inspection Report# : 2008003 (pdf)
Significance:      Sep 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Implement Maintenance Rule (a)(1) in Corrective Actions on the "G" Instrument Air Compressor The inspectors identified a finding of very low safety significance and an associated non-cited violation of 10 CFR 50.65(a)(1), Requirements for monitoring the effectiveness of maintenance at nuclear power plants. Specifically, as of August 25, 2007, the licensee failed to implement the Maintenance Rule (a)(1) action plan which had been incorporated into plant procedure N-AS-01 to preclude a loss of the G air compressor. The licensee entered the issue into their corrective action program. Corrective actions have included implementation of the procedural requirements of N-AS-01 for both the G and F air compressors.
The finding is greater than minor because it relates to a licensee failure to implement prescribed significant compensatory measures to manage risk and implement the 10 CFR 50.65(a)(1) action plan. Additionally, the finding is 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. The inspectors evaluated the finding using Inspection Manual Chapter 0609, Appendix A, Significance Determination Process, and determined that this finding is of very low safety significance by answering No to all questions in the Initiating Events Cornerstone column.
Inspection Report# : 2007004 (pdf)
Mitigating Systems
 
Significance:      Mar 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Scaffolding in Close Proximity to Multiple Safety-Related Systems Affects Operability A finding of very low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the licensees failure to install scaffolding in accordance with station procedures. Specifically, more than ten examples where scaffolding was built within 2-inches of safety-related systems without an engineering evaluation, and six examples where non-seismic scaffolding was built in safety-related areas were identified. The licensee suspended all scaffold building pending the completion of their corrective actions. The corrective actions included training scaffold builders on proper scaffold building techniques and how to identify operational and seismic concerns, revising procedures for scaffold building to address operations and engineering involvement in the scaffold building process, and a complete plant walkdown of all scaffolding by engineering or operations.
This finding was more than minor because it was 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. Specifically, the improperly installed scaffolding could have impeded or prevented proper operation of the safety-related components.
Using Attachment 4 of IMC 0609, the inspectors answered no to all the screening questions in the SDP Phase 1 Screening Worksheet in the Mitigating Systems column; therefore, this finding is of very low safety significance (Green). The inspectors determined that this finding had a cross cutting aspect in the area of problem identification and resolution, corrective action program, because the licensee did not take appropriate corrective actions to address safety issues and adverse trends in a timely manner.
Inspection Report# : 2008002 (pdf)
Significance:      Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Characterize and Manage Risk in Accordance with Maintenance Rule With the Turbine-Driven Auxiliary Feedwater Pump in Pull-to-Lock A finding of very low safety significance and an associated Non Cited Violation (NCV) of 10 CFR 50.65(a)(4),
Requirements for monitoring the effectiveness of maintenance at nuclear power plants, was identified by the inspectors during startup of the reactor following a plant shutdown to replace leaking hydrogen coolers on the turbine generator. The licensee entered this issue into its corrective action program.
The finding is greater than minor because if left uncorrected the finding would become a more significant safety concern. Specifically, the licensee failed to correctly characterize the risk on October 12 13, 2007. The inspectors evaluated the finding using Appendix K of Inspection Manual Chapter 0609, Maintenance Risk Assessment and Risk Management Significance Determination Process, and determined that this finding was of very low safety significance in accordance with Flowchart 1, Assessment of Risk Deficit.
Inspection Report# : 2007005 (pdf)
Significance:      Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Solenoid Valve Not Installed Properly A finding of very low safety significance and an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed when the B turbine building service water header isolation valve failed to fully cycle on demand. Specifically, the licensee failed to provide adequate procedures to support installation and maintenance for certain designs of solenoid valves used in the instrument air system. A failure of an instrument air system solenoid valve caused the service water valve to fail. The licensee repaired the solenoid valve and entered the issue into its corrective action program.
The finding is greater than minor because it is 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). Specifically, the licensee failed to provide adequate procedures to support the installation, maintenance, and operation of safety-related solenoid valve, SV-33044. Using Appendix A of Inspection Manual Chapter 0609, the inspectors answered no to all the screening questions in the Significance Determination Process Phase 1 Screening Worksheet in the Mitigating Systems column; therefore, this finding is of very low safety significance. This finding has a cross-cutting aspect in the area of human performance because related installation and maintenance procedures (resources) were inadequate and not up to date to ensure safety-related equipment was protected (H.2(c)).
Inspection Report# : 2007005 (pdf)
Significance:        Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedures for Post-Maintenance Testing of Steam Traps in the Turbine-Driven Auxiliary Feedwater Pump System A finding of very low safety significance and an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors during observation of a post-maintenance test on a steam trap associated with the turbine-driven auxiliary feedwater pump (TDAFWP) in accordance with plant procedure CMP-13-01 TD-Turbine Room Traps and Drains-Trap Maintenance. Specifically, the licensee failed to provide adequate procedures to support the testing of the steam trap. The licensee has entered the issue into the corrective action program and will be revising the appropriate procedures.
The finding is greater than minor because the finding is 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).
Specifically, the licensee failed to provide adequate procedures to support the post maintenance testing of steam traps in the TDAFWP system. Using Appendix A of Inspection Manual Chapter 0609, the inspectors answered no to all the screening questions in the Significance Determination Process Phase 1 Screening Worksheet in the Mitigating Systems column; therefore, this finding is of very low safety significance. This finding has a cross-cutting aspect in the area of human performance because the licensee had ample opportunity (resources) available to update procedure CMP-13-01 during multiple prior maintenance activities but did not (H.2(c)).
Inspection Report# : 2007005 (pdf)
Significance:        Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Seismic Storage Procedure A finding of very low safety significance and an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors during a review of procedures related to the control and storage of material. Specifically, procedure GNP 01.31.01, Plant Cleanliness and Storage, permitted uncontrolled storage of materials next to a Seismic Class 1 system. Additionally, opportunities existed to correct it and/or place compensatory measures in place after the NRC issued an NCV related to this issue in the first quarter of 2007. The licensee has entered the issue into the corrective action program and will be revising the procedure.
The finding is greater than minor because it is associated with the procedure quality 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). Specifically, the procedure allowed uncontrolled storage of materials in the vicinity of Seismic Class 1 systems that could render the systems inoperable during a seismic event. Using Appendix A of Inspection Manual Chapter 0609, the inspectors answered no to all the screening questions in the Significance Determination Process Phase 1 Screening Worksheet in the Mitigating Systems column; therefore, this finding is of very low safety significance. This finding has a cross-cutting aspect in the area of problem identification and resolution because the licensee failed to take appropriate timely corrective actions or put compensatory actions in place after the NRC issued a Non-Cited Violation relating to this issue in the first quarter of 2007 (P.1(d)).
 
Inspection Report# : 2007005 (pdf)
Significance:        Dec 19, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Extent of Condition Review for Fuel Leak.
The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, of very low safety significance, for failure by the licensee to follow procedural requirements for performing an adequate extent of condition for a diesel fuel line failure in 2006. Specifically, the licensee failed to complete an extent of condition which would have evaluated different systems where a similar failure mechanism (cyclic fatigue) could occur. The licensee entered the item into their corrective action program.
The issue is greater than minor because it was associated with the equipment performance attribute of the Mitigating System Cornerstone and affected the cornerstone objective of ensuring the availability of systems that respond to initiating events to prevent undesirable consequences. Specifically, it affected the equipment performance attribute for availability and reliability. Using Inspection Manual Chapter 0609, Significance Determination Process, the inspectors screened this issue as of very low safety significance (Green) because no loss of safety function occurred.
Inspection Report# : 2007011 (pdf)
Significance:        Dec 19, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Initiate Corrective Action Documents for Multiple Leaks in the Plant The inspections identified a non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, of very low safety significance. Specifically, the licensee failed to initiate corrective action documents in accordance with plant procedures for multiple leaks found in the plant. The licensee entered this item into its corrective action program.
The finding is greater than minor because it was associated with the equipment performance attribute of the Mitigating System Cornerstone and affected the cornerstone objective of ensuring the availability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to identify and correct leakage on equipment important to safety could eventually lead to equipment unavailability during events that the equipment is designed to mitigate. The finding is of very low safety significance (Green), because the inspectors answered no to all five questions under the Mitigating Systems Cornerstone column of the Phase 1 worksheet in Inspection Manual Chapter 0609, Significance Determination Process. Specifically, at the time that the leakage was discovered, none of the leaks immediately impacted the functionality of the equipment affected. The finding has a cross-cutting aspect in the area of human performance because the licensee failed to effectively communicate expectations regarding procedural compliance for the corrective action program.
Inspection Report# : 2007011 (pdf)
Significance:        Dec 19, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Update the Updated Safety Analysis Report with Safety Analysis for Pressure Locking of Containment Sump Isolation Valves The inspectors identified a non cited violation of 10 CFR 50.71, of very low safety significance, for the licensees failure to update the Updated Safety Analysis Report (USAR). Specifically, the licensee failed to update the USAR to fully reflect the results of a safety analysis performed in response to Generic Letter 95 07, Pressure Locking and Thermal Binding of Safety Related Power Operated Gate Valves. The licensee entered this issue into its corrective action program.
Because this finding potentially impacted the NRCs ability to perform its regulatory function, it was evaluated using the traditional enforcement process. The finding is greater than minor because the failure to provide complete licensing and design basis information in the USAR could result in either the licensee making an inappropriate licensing interpretation or the NRC making an inappropriate regulatory decision based on incomplete information in the USAR. NRC management determined that this issue is of very low safety significance (Green) because it is a
 
design issue confirmed not to result in a loss of operability.
Inspection Report# : 2007011 (pdf)
Significance:        Dec 19, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Extent of Condition Review of BF-66 Relays The inspectors identified a non cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, of very low safety significance, for failure by the licensee to follow procedural requirements for performing an adequate extent of condition following relay failures that led to reactor trips in 2006 and 2007.
Specifically, the licensee failed to perform an extent of condition action to inspect Engineered Safety Feature (ESF) relays when sufficient causal evidence was present that the same style relay in the ESF system (BF 66 relays) were susceptible to sulfidation, installation deficiencies, or manufacturing defects. The licensee entered this issue into its corrective action program.
The issue is greater than minor because, if left uncorrected, the failure to assess the other systems would become a more significant safety concern. Using Inspection Manual Chapter 0609, Significance Determination Process, the inspectors screened this issue as being of very low safety significance (Green) because no loss of safety function occurred.
Inspection Report# : 2007011 (pdf)
Significance: SL-IV Sep 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Update the Updated Safety Evaluation Report The inspectors identified a finding of very low safety significance for the licensee's failure to adequately update the Updated Safety Analysis Report (USAR) in accordance to 10 CFR 50.71, Maintenance of records, making of reports. The licensee failed to update the USAR to fully reflect changes and analyses made in response to license amendment 184. Once identified, the licensee entered this issue into its corrective action program.
Because this issue potentially impacted the NRC's ability to perform its regulatory function, this finding was evaluated using the traditional enforcement process. The finding is greater than minor because of the failure to provide complete licensing and design basis information in the USAR could result in either the licensee making an inappropriate licensing interpretation or the NRC making an inappropriate regulatory decision based on incomplete information in the USAR. The issue is of very low safety significance based upon a Phase 2 significance determination analysis of the associated technical issue. The issue was a NCV of 10 CFR 50.71(e), which required that the USAR be updated to include the effects of all safely evaluations performed by the licensee in support of requested license amendments.
The primary cause of this violation is related to the cross-cutting area of problem identification and resolution because the extent of condition review performed for a recent and similar violation failed to identify the issue even though it was within the scope of the extent of condition review which had been performed Inspection Report# : 2007004 (pdf)
Significance: SL-IV Jun 30, 2007 Identified By: NRC Item Type: VIO Violation Failure to Perform a 10 CFR 50.59 Evaluation for Compensatory Measures Associated with a Procedure Change The inspectors identified a finding having very low safety significance and an associated Severity Level IV, Cited Violation of 10 CFR 50.59 while reviewing unresolved items URI 05000305/2006003-04, Adequacy of Compensatory Actions for Potential Turbine Missile Strike of Control Room Ventilation Cooling; and URI 05000305/2006016-01, Adequacy of 10 CFR 50.59 Screening for Procedure Change. Specifically, the licensee failed to properly interpret design and licensing basis requirements associated with protection against external events and as a result did not perform a 10 CFR 50.59 evaluation. The cause of this finding is related to the cross-cutting area of problem identification and resolution because the licensee had similar prior problems that, if effectively evaluated and resolved, could have prevented this issue. (P.1(c))
This finding was determined to be more than minor because the inspectors determined that the procedure change
 
would have ultimately required NRC approval. The procedure changes, in the form of compensatory operator actions, adversely impacted the operation of control room recirculation system following a tornado. A Phase 1 significance determination of this finding using IMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," using the Severe Weather Screening Criteria questions was completed. Since the loss of the control room recirculation system would not result in an initiating event or degrade two or more trains of a multi-train safety system, the issue screened as Green.
Inspection Report# : 2007003 (pdf)
Barrier Integrity Significance:      Mar 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow the Provisions of Corrective Action Procedure PI-KW-200 Following Surveillance Testing of Containment Isolation Valve LOCA-31 A finding of very low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors following surveillance testing of containment isolation valve LOCA 3A in accordance with plant procedure SP 55 167 4B, "Post LOCA Valves Timing Test (IST) from Local Panel-Train B." Specifically, the licensee failed to initiate a condition report in accordance with procedure PI-KW-200, Corrective Action, following a review of the test results by the inservice testing program engineer who subsequently identified a potential condition which called into question the operability of LOCA-3A.
The finding was more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 20, 2007, because the finding was associated with the structure, system and component (SSC) and barrier performance attribute of the Barrier Integrity Cornerstone and affected the cornerstone objective to provide reasonable assurance that the physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the licensee failed to implement the provisions of Corrective Action Procedure, PI-KW-200, which resulted in a failure to ensure operability of containment isolation valve LOCA-3A. The inspectors also determined that the primary cause for this finding was related to the cross cutting area of human performance, work practices, because personnel have been trained in need for procedural use and adherence but did not follow applicable procedures.
Inspection Report# : 2008002 (pdf)
Significance:      Sep 30, 2007 Identified By: NRC Item Type: NCV NonCited Violation Preconditioning of Safety-Related Motor-Operated Valves Prior to Performance of Technical Specification Required Surveillance Testing The inspectors identified a finding of very low safety significance and an associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, during plant preparations to perform Surveillance Procedure SP-23-100B, Train B Containment Spray Pump and Valve Test - IST. Specifically, the inspectors noted on August 8, 2007, that shortly prior to performing the surveillance procedure, the plant had hung safety tags on the containment spray system in order to perform repair activities on IDS-102, a check valve in that system. These tags required that normally open motor- operated valves IDS-202 and IDS-2B be cycled closed and tagged in order to isolate the check valve. Because these motor-operated valves were required to be stroke and time-tested during the performance of the surveillance procedure, and the effects of preconditioning on these valves was not considered prior to implementation of the maintenance activity, the inspectors determined that plant procedures were inadequate to assess preconditioning implications associated with station activities. The licensee entered the issue into their corrective action program. Corrective actions included completion of the surveillance procedure with acceptable results and a evaluation of the test results, which determined that the surveillance test was acceptable.
The finding is greater than minor because it was associated with the configuration control attribute of the Barrier Integrity Cornerstone and 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. The inspectors evaluated the finding using IMC 0609, Appendix A, Significance Determination Process, and determined that this finding is of very low safety significance by answering No to all questions in the containment barriers cornerstone column. The inspectors also determined that the primary cause for this finding is related to the cross-cutting area of human performance. Specifically, under the component of resources, procedures to assess and prevent preconditioning of safety-related components were not complete, accurate, and up-to-date Inspection Report# : 2007004 (pdf)
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.
Miscellaneous Last modified : August 29, 2008
 
Kewaunee 3Q/2008 Plant Inspection Findings Initiating Events Significance:        Jun 30, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow the Provisions of General Nuclear Procedure, GNP-12.06.01, "Hot and Cold Weather Operations."
Green. A finding of very low safety significance (Green) and an NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors following an inspection of licensee preparations for adverse weather protection. Specifically, the licensee failed to perform inspections for hot weather operations as required by plant procedure GNP-12.06.01, "Hot and Cold Weather Operations."
The finding was greater than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 20, 2007, because if left uncorrected would become a more significant safety concern. Specifically, the licensee failed to implement the provisions of GNP 12.06.01, "Hot and Cold Weather Operations," which resulted in a failure to ensure pre-summer readiness of numerous safety-related and risk-significant systems. The inspectors evaluated the finding using Attachment 0609.04, of IMC 0609, Significance Determination Process, dated January 10, 2008, and answered no to all of the questions in the Initiating Events column; therefore, the finding was determined to be of very low safety significance. The inspectors determined that the primary cause for this finding was related to the cross cutting area of human performance, work practices component, because personnel have been trained in the need for procedural use and adherence, but failed to follow applicable procedures. Specifically, the procedure which required the performance of plant inspections for hot weather operations, and the maintenance of QA documentation for these inspections, was not followed [H.4(b)]
Inspection Report# : 2008003 (pdf)
Mitigating Systems Significance:        Sep 30, 2008 Identified By: NRC Item Type: NCV NonCited Violation Operability Evaluation for Degraded Gauge Pedestals Failed to Adequately Evaluate Degraded Conditions Per Procedures A finding of very low safety significance and an associated NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors during a review of an operability evaluation for degraded concrete support pads under the discharge pressure gauge pedestals for safety-related service water pumps A1 and A2. Specifically, procedure OP AA 102, Operability Determination, required that when a potential degraded or nonconforming condition is identified, action must be taken to discover the facts and confirm the condition of the systems, structures, and components. The licensees operability evaluation failed to adequately evaluate the degraded condition and failed to confirm that the compensatory actions used as a basis for operability for the pumps were effective.
Corrective actions included the engineering department providing a more thorough evaluation of the potential for damage to the gauge isolation valve and associated piping from a falling gauge support including field measurements and piping configuration information.
The finding is greater than minor because the failure to perform an adequate operability evaluation, if left uncorrected, would become a more significant failure to comply with the technical specifications or the licensing basis. The significance of the finding was determined to be of very low safety significance because the inspectors answered no to all of the questions for the Mitigation Systems Cornerstone column of 609.04, of IMC 0609, Significance Determination Process. Additionally, the inspectors attributed this issue to the cross cutting area of problem identification, corrective action program, because the operability evaluation and associated problems were not thoroughly evaluated.
Inspection Report# : 2008004 (pdf)
Significance:        Sep 30, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform a 10 CFR Part 50.59 Screening for Alteration During Maintenance that Existed for More Than 90 Days A finding of very low safety significance and an associated Severity Level IV, NCV of 10 CFR 50.59 was identified by the inspectors for a
 
failure to perform a 50.59 screening for an alteration during maintenance that existed for more than 90 days. Specifically, the licensee failed to perform a 50.59 screening when spare breakers were removed from safety related motor control centers (MCCs) and the cubicle were left in an altered state for more than 90 days. Proposed corrective actions include changes to the station housekeeping and work control/planning procedures to better evaluate job site and environmental conditions.
The finding is greater than minor because, if left uncorrected, the failure to perform a 10 CFR 50.59 screening on an alteration/change to the facility would become more significant. 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 2 for the Mitigation Systems Cornerstone. Using information provided by the licensee relative to the affected MCCs, the inspectors answered no to all of the questions in this cornerstone column; therefore, the finding was determined to be of very low safety significance. Additionally, the inspectors determined that the finding has a cross-cutting aspect in the area of human performance, work control because the licensee failed to appropriately plan work activities by incorporating risk insights gained from operating experience and factor in environmental conditions during planning contingencies for systems, structures, and components anticipated to be in a maintenance condition for extensive periods of time.
Inspection Report# : 2008004 (pdf)
Significance:        Mar 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Scaffolding in Close Proximity to Multiple Safety-Related Systems Affects Operability A finding of very low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the licensees failure to install scaffolding in accordance with station procedures. Specifically, more than ten examples where scaffolding was built within 2-inches of safety-related systems without an engineering evaluation, and six examples where non-seismic scaffolding was built in safety-related areas were identified. The licensee suspended all scaffold building pending the completion of their corrective actions. The corrective actions included training scaffold builders on proper scaffold building techniques and how to identify operational and seismic concerns, revising procedures for scaffold building to address operations and engineering involvement in the scaffold building process, and a complete plant walkdown of all scaffolding by engineering or operations.
This finding was more than minor because it was 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. Specifically, the improperly installed scaffolding could have impeded or prevented proper operation of the safety-related components. Using Attachment 4 of IMC 0609, the inspectors answered no to all the screening questions in the SDP Phase 1 Screening Worksheet in the Mitigating Systems column; therefore, this finding is of very low safety significance (Green). The inspectors determined that this finding had a cross cutting aspect in the area of problem identification and resolution, corrective action program, because the licensee did not take appropriate corrective actions to address safety issues and adverse trends in a timely manner.
Inspection Report# : 2008002 (pdf)
Significance:        Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Characterize and Manage Risk in Accordance with Maintenance Rule With the Turbine-Driven Auxiliary Feedwater Pump in Pull-to-Lock A finding of very low safety significance and an associated Non Cited Violation (NCV) of 10 CFR 50.65(a)(4), Requirements for monitoring the effectiveness of maintenance at nuclear power plants, was identified by the inspectors during startup of the reactor following a plant shutdown to replace leaking hydrogen coolers on the turbine generator. The licensee entered this issue into its corrective action program.
The finding is greater than minor because if left uncorrected the finding would become a more significant safety concern. Specifically, the licensee failed to correctly characterize the risk on October 12 13, 2007. The inspectors evaluated the finding using Appendix K of Inspection Manual Chapter 0609, Maintenance Risk Assessment and Risk Management Significance Determination Process, and determined that this finding was of very low safety significance in accordance with Flowchart 1, Assessment of Risk Deficit.
Inspection Report# : 2007005 (pdf)
Significance:        Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Solenoid Valve Not Installed Properly A finding of very low safety significance and an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed when the B turbine building service water header isolation valve failed to fully cycle on demand. Specifically, the licensee failed to provide adequate procedures to support installation and maintenance for certain designs of
 
solenoid valves used in the instrument air system. A failure of an instrument air system solenoid valve caused the service water valve to fail.
The licensee repaired the solenoid valve and entered the issue into its corrective action program.
The finding is greater than minor because it is 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). Specifically, the licensee failed to provide adequate procedures to support the installation, maintenance, and operation of safety-related solenoid valve, SV-33044. Using Appendix A of Inspection Manual Chapter 0609, the inspectors answered no to all the screening questions in the Significance Determination Process Phase 1 Screening Worksheet in the Mitigating Systems column; therefore, this finding is of very low safety significance. This finding has a cross-cutting aspect in the area of human performance because related installation and maintenance procedures (resources) were inadequate and not up to date to ensure safety-related equipment was protected (H.2(c)).
Inspection Report# : 2007005 (pdf)
Significance:          Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedures for Post-Maintenance Testing of Steam Traps in the Turbine-Driven Auxiliary Feedwater Pump System A finding of very low safety significance and an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors during observation of a post-maintenance test on a steam trap associated with the turbine-driven auxiliary feedwater pump (TDAFWP) in accordance with plant procedure CMP-13-01 TD-Turbine Room Traps and Drains-Trap Maintenance. Specifically, the licensee failed to provide adequate procedures to support the testing of the steam trap. The licensee has entered the issue into the corrective action program and will be revising the appropriate procedures.
The finding is greater than minor because the finding is 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). Specifically, the licensee failed to provide adequate procedures to support the post maintenance testing of steam traps in the TDAFWP system. Using Appendix A of Inspection Manual Chapter 0609, the inspectors answered no to all the screening questions in the Significance Determination Process Phase 1 Screening Worksheet in the Mitigating Systems column; therefore, this finding is of very low safety significance. This finding has a cross-cutting aspect in the area of human performance because the licensee had ample opportunity (resources) available to update procedure CMP-13-01 during multiple prior maintenance activities but did not (H.2(c)).
Inspection Report# : 2007005 (pdf)
Significance:          Dec 31, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Seismic Storage Procedure A finding of very low safety significance and an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors during a review of procedures related to the control and storage of material.
Specifically, procedure GNP 01.31.01, Plant Cleanliness and Storage, permitted uncontrolled storage of materials next to a Seismic Class 1 system. Additionally, opportunities existed to correct it and/or place compensatory measures in place after the NRC issued an NCV related to this issue in the first quarter of 2007. The licensee has entered the issue into the corrective action program and will be revising the procedure.
The finding is greater than minor because it is associated with the procedure quality 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). Specifically, the procedure allowed uncontrolled storage of materials in the vicinity of Seismic Class 1 systems that could render the systems inoperable during a seismic event. Using Appendix A of Inspection Manual Chapter 0609, the inspectors answered no to all the screening questions in the Significance Determination Process Phase 1 Screening Worksheet in the Mitigating Systems column; therefore, this finding is of very low safety significance. This finding has a cross-cutting aspect in the area of problem identification and resolution because the licensee failed to take appropriate timely corrective actions or put compensatory actions in place after the NRC issued a Non-Cited Violation relating to this issue in the first quarter of 2007 (P.1(d)).
Inspection Report# : 2007005 (pdf)
Significance:          Dec 19, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Extent of Condition Review for Fuel Leak.
The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, of very low safety significance, for failure by the licensee to follow procedural requirements for performing an adequate extent of condition for a diesel fuel line failure in 2006. Specifically, the licensee failed to complete an extent of condition which would have evaluated different systems
 
where a similar failure mechanism (cyclic fatigue) could occur. The licensee entered the item into their corrective action program.
The issue is greater than minor because it was associated with the equipment performance attribute of the Mitigating System Cornerstone and affected the cornerstone objective of ensuring the availability of systems that respond to initiating events to prevent undesirable consequences. Specifically, it affected the equipment performance attribute for availability and reliability. Using Inspection Manual Chapter 0609, Significance Determination Process, the inspectors screened this issue as of very low safety significance (Green) because no loss of safety function occurred.
Inspection Report# : 2007011 (pdf)
Significance:        Dec 19, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Initiate Corrective Action Documents for Multiple Leaks in the Plant The inspections identified a non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, of very low safety significance. Specifically, the licensee failed to initiate corrective action documents in accordance with plant procedures for multiple leaks found in the plant. The licensee entered this item into its corrective action program.
The finding is greater than minor because it was associated with the equipment performance attribute of the Mitigating System Cornerstone and affected the cornerstone objective of ensuring the availability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to identify and correct leakage on equipment important to safety could eventually lead to equipment unavailability during events that the equipment is designed to mitigate. The finding is of very low safety significance (Green), because the inspectors answered no to all five questions under the Mitigating Systems Cornerstone column of the Phase 1 worksheet in Inspection Manual Chapter 0609, Significance Determination Process. Specifically, at the time that the leakage was discovered, none of the leaks immediately impacted the functionality of the equipment affected. The finding has a cross-cutting aspect in the area of human performance because the licensee failed to effectively communicate expectations regarding procedural compliance for the corrective action program.
Inspection Report# : 2007011 (pdf)
Significance:        Dec 19, 2007 Identified By: NRC Item Type: NCV NonCited Violation Failure to Update the Updated Safety Analysis Report with Safety Analysis for Pressure Locking of Containment Sump Isolation Valves The inspectors identified a non cited violation of 10 CFR 50.71, of very low safety significance, for the licensees failure to update the Updated Safety Analysis Report (USAR). Specifically, the licensee failed to update the USAR to fully reflect the results of a safety analysis performed in response to Generic Letter 95 07, Pressure Locking and Thermal Binding of Safety Related Power Operated Gate Valves. The licensee entered this issue into its corrective action program.
Because this finding potentially impacted the NRCs ability to perform its regulatory function, it was evaluated using the traditional enforcement process. The finding is greater than minor because the failure to provide complete licensing and design basis information in the USAR could result in either the licensee making an inappropriate licensing interpretation or the NRC making an inappropriate regulatory decision based on incomplete information in the USAR. NRC management determined that this issue is of very low safety significance (Green) because it is a design issue confirmed not to result in a loss of operability.
Inspection Report# : 2007011 (pdf)
Significance:        Dec 19, 2007 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Extent of Condition Review of BF-66 Relays The inspectors identified a non cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, of very low safety significance, for failure by the licensee to follow procedural requirements for performing an adequate extent of condition following relay failures that led to reactor trips in 2006 and 2007. Specifically, the licensee failed to perform an extent of condition action to inspect Engineered Safety Feature (ESF) relays when sufficient causal evidence was present that the same style relay in the ESF system (BF 66 relays) were susceptible to sulfidation, installation deficiencies, or manufacturing defects. The licensee entered this issue into its corrective action program.
The issue is greater than minor because, if left uncorrected, the failure to assess the other systems would become a more significant safety concern. Using Inspection Manual Chapter 0609, Significance Determination Process, the inspectors screened this issue as being of very low safety significance (Green) because no loss of safety function occurred.
Inspection Report# : 2007011 (pdf)
Significance: SL-IV Jun 30, 2007 Identified By: NRC
 
Item Type: VIO Violation Failure to Perform a 10 CFR 50.59 Evaluation for Compensatory Measures Associated with a Procedure Change The inspectors identified a finding having very low safety significance and an associated Severity Level IV, Cited Violation of 10 CFR 50.59 while reviewing unresolved items URI 05000305/2006003-04, Adequacy of Compensatory Actions for Potential Turbine Missile Strike of Control Room Ventilation Cooling; and URI 05000305/2006016-01, Adequacy of 10 CFR 50.59 Screening for Procedure Change.
Specifically, the licensee failed to properly interpret design and licensing basis requirements associated with protection against external events and as a result did not perform a 10 CFR 50.59 evaluation. The cause of this finding is related to the cross-cutting area of problem identification and resolution because the licensee had similar prior problems that, if effectively evaluated and resolved, could have prevented this issue. (P.1(c))
This finding was determined to be more than minor because the inspectors determined that the procedure change would have ultimately required NRC approval. The procedure changes, in the form of compensatory operator actions, adversely impacted the operation of control room recirculation system following a tornado. A Phase 1 significance determination of this finding using IMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," using the Severe Weather Screening Criteria questions was completed. Since the loss of the control room recirculation system would not result in an initiating event or degrade two or more trains of a multi-train safety system, the issue screened as Green.
Inspection Report# : 2007003 (pdf)
Barrier Integrity Significance:        Sep 30, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedure Results in Unplanned Control Rod Motion A finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed when control rods automatically stepped inward unexpectedly. Ultimately, it was determined that procedures for operation of the power range nuclear instrument were found to be inadequate for the circumstances. Specifically, procedures for bypassing nuclear instrument N 43 did not contain steps to place control rods in manual when placing a failed instrument in bypass. Corrective actions were taken to replace the inappropriately deleted steps from the associated procedures.
The finding is greater than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 20, 2007, because the finding affected the procedure quality attribute of the Barrier Integrity Cornerstone of Reactor Safety.
Specifically, the failure to either leave the step for placing rods in manual in multiple alarm response procedures, or transferring the step to the common procedure OP KW AOP MISC 001, resulted in a preventable condition which resulted in an unexpected reactivity transient. The inspectors evaluated the finding using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings, Table 2 for the Barriers Cornerstone. The inspectors answered no to all of the questions in this cornerstone column; therefore, the finding was determined to be of very low safety significance. The inspectors concluded that the finding had a cross-cutting aspect in the area of human performance, decision-making, because interdisciplinary reviews performed by station personnel, including the on site safety review committee, failed to make changes to the various procedures using a systematic process. Additionally, the inspectors reviewed the licensee evaluation of the cause of the issue and found that it agreed with their understanding of the issue.
Inspection Report# : 2008004 (pdf)
Significance:        Mar 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow the Provisions of Corrective Action Procedure PI-KW-200 Following Surveillance Testing of Containment Isolation Valve LOCA-31 A finding of very low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors following surveillance testing of containment isolation valve LOCA 3A in accordance with plant procedure SP 55 167 4B, "Post LOCA Valves Timing Test (IST) from Local Panel-Train B." Specifically, the licensee failed to initiate a condition report in accordance with procedure PI-KW-200, Corrective Action, following a review of the test results by the inservice testing program engineer who subsequently identified a potential condition which called into question the operability of LOCA-3A.
The finding was more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 20, 2007, because the finding was associated with the structure, system and component (SSC) and barrier performance attribute of the Barrier Integrity Cornerstone and affected the cornerstone objective to provide reasonable assurance that the physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the licensee failed to implement the provisions of Corrective Action Procedure, PI-KW-200, which resulted in a failure to ensure operability of containment isolation valve LOCA-3A. The inspectors also determined that the primary cause for this finding was related to the cross
 
cutting area of human performance, work practices, because personnel have been trained in need for procedural use and adherence but did not follow applicable procedures.
Inspection Report# : 2008002 (pdf)
Emergency Preparedness Significance: TBD Aug 29, 2008 Identified By: NRC Item Type: AV Apparent Violation Failure to Maintain a Standard EAL Scheme An AV was identified by the inspector for failure to follow and maintain in effect emergency plans which use a standard emergency classification and action level scheme. Specifically, the licensee's emergency plan Alert emergency action levels (EALs) RA1.1 and RA1.2 specified instrument setpoints that were beyond the limits of the effluent radiation monitors capabilities.
This finding was considered more than minor because the licensee is required to be capable to implement adequate measures to protect public health and safety in the event of a radiological emergency. Regulations require a standard emergency classification and action level scheme, the bases which included facility system and effluent parameters, in use by the licensee and State and local response plans call for reliance on information provided by the licensee for determination of minimum initial offsite response measures. As a result of having Alert EAL threshold values that were beyond the range of the associated effluent radiation monitors, Kewaunee personnel would not have been able to classify an emergency based upon an effluent radioactive material release in a timely manner. Emergency response actions directed by the State and local emergency response plans, which rely on information provided by the licensee, could have potentially been delayed.
The cause of the finding is related to the human performance cross-cutting element of H.2(c) for ensuring that personnel, equipment, and procedures are available and adequate to assure nuclear safety. Specifically, those necessary for complete, accurate, and up-to-date design documentation, procedures, and work packages.
Inspection Report# : 2008503 (pdf)
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.
Miscellaneous Last modified : November 26, 2008
 
Kewaunee 4Q/2008 Plant Inspection Findings Initiating Events Significance:      Dec 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Transfer Turbine Valve Testing Requirements into the USAR
. A finding of very low safety significance and associated Severity Level IV NCV of 10 CFR 50.71, Maintenance of records, making of reports, was identified by the inspectors for the licensees failure to adequately update the Kewaunee Power Station Updated Safety Analysis Report (USAR). Specifically, the inspectors identified that the licensee had not updated the USAR completely when they relocated the turbine valve testing requirements from technical specifications to the USAR in License Amendment No. 121. Proposed corrective actions include performing an apparent cause evaluation and USAR changes as appropriate.
This finding was more than minor because it had a material impact on licensed activities in that the incorrect USAR allowed the licensee to schedule periodic testing of the reheat and interceptor valves at an interval beyond one year.
The inspectors evaluated the finding using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings, Table 3b, for the Initiating Events Cornerstone, dated January 10, 2008. Using information provided by the licensee, the inspectors answered no to the transient initiator contributor questions in this cornerstone column; therefore, the finding was determined to be of very low safety significance (Green). Additionally, the inspectors determined that the finding has a cross-cutting aspect in the area of problem identification and resolution, corrective action program, because the licensee failed to accurately identify the issue when conducting corrective actions for Condition Report CR040457, Discrepancy in Turbine Valve Testing Requirements and Acceptance Criteria, [P.1(a)].
Inspection Report# : 2008005 (pdf)
Significance:      Jun 30, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow the Provisions of General Nuclear Procedure, GNP-12.06.01, "Hot and Cold Weather Operations" A finding of very low safety significance (Green) and a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors following an inspection of licensee preparations for adverse weather protection. Specifically, the licensee failed to perform inspections for hot weather operations as required by plant procedure GNP-12.06.01, "Hot and Cold Weather Operations."
The finding was greater than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 20, 2007, because if left uncorrected would become a more significant safety concern. Specifically, the licensee failed to implement the provisions of GNP 12.06.01, "Hot and Cold Weather Operations," which resulted in a failure to ensure pre-summer readiness of numerous safety-related and risk-significant systems. The inspectors evaluated the finding using Attachment 0609.04, of IMC 0609, Significance Determination Process, dated January 10, 2008, and answered no to all of the questions in the Initiating Events column; therefore, the finding was determined to be of very low safety significance. The inspectors determined that the primary cause for this finding was related to the cross cutting area of human performance, work practices component, because personnel have been trained in the need for procedural use and adherence, but failed to follow applicable procedures. Specifically, the procedure which required the performance of plant inspections for hot weather operations, and the maintenance of QA documentation for these inspections, was not followed [H.4(b)]
 
Inspection Report# : 2008003 (pdf)
Mitigating Systems Significance:        Dec 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Emergency Diesel Generator Room Cooling Fan Testing Deficienceies
. A finding of very low safety significance and an associated NCV of 10 CFR, Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors for the licensees failure to verify the ventilation flow rate for the emergency diesel generator (EDG) rooms and for using an incorrect EDG heat load in the design basis calculation of record. As part of corrective actions, the licensee remeasured flow rates and duct dimensions and recalculated post-accident temperature values.
This finding was more than minor because a revision to the design calculation was necessary to demonstrate EDG cubicle temperatures would remain under the design basis 120 degrees Fahrenheit (&#xba;F) equipment qualification limit under any accident conditions. The finding was of very low safety significance (Green) based on a Phase 1 screening in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, dated January 10, 2008, because the revised calculation showed that the diesel generators had remained operable in all circumstances. There was no cross-cutting aspect associated with this finding.
Inspection Report# : 2008005 (pdf)
Significance:        Sep 30, 2008 Identified By: NRC Item Type: NCV NonCited Violation Operability Evaluation for Degraded Gauge Pedestals Failed to Adequately Evaluate Degraded Conditions Per Procedures A finding of very low safety significance and an associated NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors during a review of an operability evaluation for degraded concrete support pads under the discharge pressure gauge pedestals for safety-related service water pumps A1 and A2. Specifically, procedure OP AA 102, Operability Determination, required that when a potential degraded or nonconforming condition is identified, action must be taken to discover the facts and confirm the condition of the systems, structures, and components. The licensees operability evaluation failed to adequately evaluate the degraded condition and failed to confirm that the compensatory actions used as a basis for operability for the pumps were effective. Corrective actions included the engineering department providing a more thorough evaluation of the potential for damage to the gauge isolation valve and associated piping from a falling gauge support including field measurements and piping configuration information.
The finding is greater than minor because the failure to perform an adequate operability evaluation, if left uncorrected, would become a more significant failure to comply with the technical specifications or the licensing basis. The significance of the finding was determined to be of very low safety significance because the inspectors answered no to all of the questions for the Mitigation Systems Cornerstone column of Attachment 0609.04, of IMC 0609, Significance Determination Process. Additionally, the inspectors attributed this issue to the cross cutting area of problem identification, corrective action program, because the operability evaluation and associated problems were not thoroughly evaluated.
Inspection Report# : 2008004 (pdf)
Significance:        Sep 30, 2008
 
Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform a 10 CFR Part 50.59 Screening for Alteration During Maintenance that Existed for More Than 90 Days A finding of very low safety significance and an associated Severity Level IV, Non-Cited Violation of 10 CFR 50.59 was identified by the inspectors for a failure to perform a 50.59 screening for an alteration during maintenance that existed for more than 90 days. Specifically, the licensee failed to perform a 50.59 screening when spare breakers were removed from safety-related motor control centers (MCCs) and the cubicle were left in an altered state for more than 90 days. Proposed corrective actions include changes to the station housekeeping and work control/planning procedures to better evaluate job site and environmental conditions.
The finding is greater than minor because, if left uncorrected, the failure to perform a 10 CFR 50.59 screening on an alteration/change to the facility would become more significant. The inspectors determined the finding could be evaluated using the Significance Determination Process in accordance with Inspection Manual Chapter 0609, Significance Determination Process, Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings, Table 2 for the Mitigation Systems Cornerstone. Using information provided by the licensee relative to the affected MCCs, the inspectors answered no to all of the questions in this cornerstone column; therefore, the finding was determined to be of very low safety significance. Additionally, the inspectors determined that the finding has a cross-cutting aspect in the area of human performance, work control because the licensee failed to appropriately plan work activities by incorporating risk insights gained from operating experience and factor in environmental conditions during planning contingencies for systems, structures, and components anticipated to be in a maintenance condition for extensive periods of time.
Inspection Report# : 2008004 (pdf)
Significance:        Mar 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Scaffolding in Close Proximity to Multiple Safety-Related Systems Affects Operability A finding of very low safety significance (Green) and an associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the licensees failure to install scaffolding in accordance with station procedures. Specifically, more than ten examples where scaffolding was built within 2-inches of safety-related systems without an engineering evaluation, and six examples where non-seismic scaffolding was built in safety-related areas were identified. The licensee suspended all scaffold building pending the completion of their corrective actions. The corrective actions included training scaffold builders on proper scaffold building techniques and how to identify operational and seismic concerns, revising procedures for scaffold building to address operations and engineering involvement in the scaffold building process, and a complete plant walkdown of all scaffolding by engineering or operations.
This finding was more than minor because it was 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. Specifically, the improperly installed scaffolding could have impeded or prevented proper operation of the safety-related components.
Using Attachment 4 of Inspection Manual Chapter 0609, the inspectors answered no to all the screening questions in the Significance Determination Process Phase 1 Screening Worksheet in the Mitigating Systems column; therefore, this finding is of very low safety significance (Green). The inspectors determined that this finding had a cross cutting aspect in the area of problem identification and resolution, corrective action program, because the licensee did not take appropriate corrective actions to address safety issues and adverse trends in a timely manner.
Inspection Report# : 2008002 (pdf)
Significance: SL-IV Jun 30, 2007 Identified By: NRC Item Type: VIO Violation Failure to Perform a 10 CFR 50.59 Evaluation for Compensatory Measures Associated with a Procedure
 
Change The inspectors identified a finding having very low safety significance and an associated Severity Level IV, Cited Violation of 10 CFR 50.59 while reviewing unresolved items URI 05000305/2006003-04, Adequacy of Compensatory Actions for Potential Turbine Missile Strike of Control Room Ventilation Cooling; and URI 05000305/2006016-01, Adequacy of 10 CFR 50.59 Screening for Procedure Change. Specifically, the licensee failed to properly interpret design and licensing basis requirements associated with protection against external events and as a result did not perform a 10 CFR 50.59 evaluation. The cause of this finding is related to the cross-cutting area of problem identification and resolution because the licensee had similar prior problems that, if effectively evaluated and resolved, could have prevented this issue. (P.1(c))
This finding was determined to be more than minor because the inspectors determined that the procedure change would have ultimately required NRC approval. The procedure changes, in the form of compensatory operator actions, adversely impacted the operation of control room recirculation system following a tornado. A Phase 1 significance determination of this finding using IMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," using the Severe Weather Screening Criteria questions was completed. Since the loss of the control room recirculation system would not result in an initiating event or degrade two or more trains of a multi-train safety system, the issue screened as Green.
Inspection Report# : 2007003 (pdf)
Barrier Integrity Significance:      Sep 30, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedure Results in Unplanned Control Rod Motion A finding of very low safety significance and an associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed when control rods automatically stepped inward unexpectedly. Ultimately, it was determined that procedures for operation of the power range nuclear instrument were found to be inadequate for the circumstances. Specifically, procedures for bypassing nuclear instrument N 43 did not contain steps to place control rods in manual when placing a failed instrument in bypass.
Corrective actions were taken to replace the inappropriately deleted steps from the associated procedures.
The finding is greater than minor in accordance with Inspection Manual Chapter 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 20, 2007, because the finding affected the procedure quality attribute of the Barrier Integrity Cornerstone of Reactor Safety. Specifically, the failure to either leave the step for placing rods in manual in multiple alarm response procedures, or transferring the step to the common procedure OP KW AOP MISC 001, resulted in a preventable condition which resulted in an unexpected reactivity transient. The inspectors evaluated the finding using the Significance Determination Process in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings, Table 2 for the Barriers Cornerstone. The inspectors answered no to all of the questions in this cornerstone column; therefore, the finding was determined to be of very low safety significance. The inspectors concluded that the finding had a cross-cutting aspect in the area of human performance, decision-making, because interdisciplinary reviews performed by station personnel, including the on site safety review committee, failed to make changes to the various procedures using a systematic process. Additionally, the inspectors reviewed the licensee evaluation of the cause of the issue and found that it agreed with their understanding of the issue.
Inspection Report# : 2008004 (pdf)
Significance:      Mar 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow the Provisions of Corrective Action Procedure PI-KW-200 Following Surveillance Testing of Containment Isolation Valve LOCA-31
 
A finding of very low safety significance (Green) and an associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors following surveillance testing of containment isolation valve LOCA 3A in accordance with plant procedure SP 55 167 4B, "Post LOCA Valves Timing Test (IST) from Local Panel-Train B." Specifically, the licensee failed to initiate a condition report in accordance with procedure PI-KW-200, Corrective Action, following a review of the test results by the inservice testing program engineer who subsequently identified a potential condition which called into question the operability of LOCA-3A.
The finding was more than minor in accordance with Inspection Manual Chapter 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 20, 2007, because the finding was associated with the structure, system and component (SSC) and barrier performance attribute of the Barrier Integrity Cornerstone and affected the cornerstone objective to provide reasonable assurance that the physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events.
Specifically, the licensee failed to implement the provisions of Corrective Action Procedure, PI-KW-200, which resulted in a failure to ensure operability of containment isolation valve LOCA-3A. The inspectors also determined that the primary cause for this finding was related to the cross cutting area of human performance, work practices, because personnel have been trained in need for procedural use and adherence but did not follow applicable procedures.
Inspection Report# : 2008002 (pdf)
Emergency Preparedness Significance:        Aug 29, 2008 Identified By: NRC Item Type: VIO Violation Failure to Maintain a Standard EAL Scheme An apparent violation was identified by the inspector for failure to follow and maintain in effect emergency plans which use a standard emergency classification and action level scheme. Specifically, the licensee's emergency plan Alert emergency action levels (EALs) RA1.1 and RA1.2 specified instrument setpoints that were beyond the limits of the effluent radiation monitors capabilities.
This finding was considered more than minor because the licensee is required to be capable to implement adequate measures to protect public health and safety in the event of a radiological emergency. Regulations require a standard emergency classification and action level scheme, the bases which included facility system and effluent parameters, in use by the licensee and State and local response plans call for reliance on information provided by the licensee for determination of minimum initial offsite response measures. As a result of having Alert EAL threshold values that were beyond the range of the associated effluent radiation monitors, Kewaunee personnel would not have been able to classify an emergency based upon an effluent radioactive material release in a timely manner. Emergency response actions directed by the State and local emergency response plans, which rely on information provided by the licensee, could have potentially been delayed.
The cause of the finding is related to the human performance cross-cutting element of H.2(c) for ensuring that personnel, equipment, and procedures are available and adequate to assure nuclear safety. Specifically, those necessary for complete, accurate, and up-to-date design documentation, procedures, and work packages.
Final significance determination letter was issued on October 29, 2008, and no response was required to the white violation.
Inspection Report# : 2008503 (pdf)
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.
Miscellaneous Significance: N/A Jun 06, 2008 Identified By: NRC Item Type: FIN Finding PI&R Inspection Report Summary On the basis of the sample selected for review, the team concluded that implementation of the corrective action program (CAP) at Kewaunee Power Station was generally good. The licensee had a low threshold for identifying problems and entering them in the CAP. Items entered into the CAP were screened and prioritized in a timely manner using established criteria; were properly evaluated commensurate with their safety significance; and corrective actions were generally implemented in a timely manner, commensurate with the safety significance. The team noted that the licensee reviewed Operating Experience (OE) for applicability to station activities. Audits and self-assessments were determined to be performed at an appropriate level to identify deficiencies. In interviews conducted during the inspection, workers at the site expressed freedom to enter safety concerns into the CAP.
Inspection Report# : 2008007 (pdf)
Last modified : April 07, 2009
 
Kewaunee 1Q/2009 Plant Inspection Findings Initiating Events Significance:      Mar 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inappropriate Application Of A Dedicated Operator During A Component Cooling Water Surveillance A finding of very low safety significance (Green) and associated Non-Cited Violation of 10 CFR 50.65(a)(4) was identified by the inspectors for the failure to properly assess risk that resulted from risk significant maintenance being performed on the component cooling water (CCW) system, when the licensee inappropriately applied criteria for the use of a dedicated operator to meet availability requirements. As part of its corrective actions, the licensee stopped work that required the use of a dedicated operator pending further evaluation.
The issue was more than minor because the licensees risk assessment for March 11, 2009, failed to consider the CCW unavailable during maintenance. Specifically, the failure to account for the unavailability of CCW resulted in an inadequate daily risk assessment and could affect the unavailability time of this system in related performance and maintenance rule indicators. The inspectors evaluated the finding using the Significance Determination Process in accordance with Inspection Manual Chapter 0609, Significance Determination Process, Attachment K, Maintenance Risk Assessment and Risk Management Significance Determination Process, dated May 19, 2005, and determined the issue screened as having very low safety significance (Green), because the incremental conditional core damage probability was less than 1E 6 due to the test condition lasting only four hours. The inspectors determined that the finding had a cross cutting aspect in the corrective action program component of problem identification and resolution, because the licensee failed to thoroughly evaluate a prior problem such that the resolution addressed the causes and extent of conditions necessary to preclude this event (P.1(c)).
Inspection Report# : 2009002 (pdf)
Significance:      Dec 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Transfer Turbine Valve Testing Requirements into the USAR A finding of very low safety significance and associated Severity Level IV Non-Cited Violation of 10 CFR 50.71, Maintenance of records, making of reports, was identified by the inspectors for the licensees failure to adequately update the Kewaunee Power Station Updated Safety Analysis Report (USAR). Specifically, the inspectors identified that the licensee had not updated the USAR completely when they relocated the turbine valve testing requirements from technical specifications to the USAR in License Amendment No. 121. Proposed corrective actions include performing an apparent cause evaluation and USAR changes as appropriate.
This finding was more than minor because it had a material impact on licensed activities in that the incorrect USAR allowed the licensee to schedule periodic testing of the reheat and interceptor valves at an interval beyond one year.
The inspectors evaluated the finding using the Significance Determination Process in accordance with Inspection Manual Chapter 0609, Significance Determination Process, Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings, Table 3b, for the Initiating Events Cornerstone, dated January 10, 2008. Using information provided by the licensee, the inspectors answered no to the transient initiator contributor questions in this cornerstone column; therefore, the finding was determined to be of very low safety significance (Green).
Additionally, the inspectors determined that the finding has a cross-cutting aspect in the area of problem identification and resolution, corrective action program, because the licensee failed to accurately identify the issue when conducting corrective actions for Condition Report CR040457, Discrepancy in Turbine Valve Testing Requirements and Acceptance Criteria, [P.1(a)].
 
Inspection Report# : 2008005 (pdf)
Significance:        Jun 30, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Follow the Provisions of General Nuclear Procedure, GNP-12.06.01, "Hot and Cold Weather Operations" A finding of very low safety significance (Green) and a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors following an inspection of licensee preparations for adverse weather protection. Specifically, the licensee failed to perform inspections for hot weather operations as required by plant procedure GNP-12.06.01, "Hot and Cold Weather Operations."
The finding was greater than minor in accordance with Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 20, 2007, because if left uncorrected would become a more significant safety concern. Specifically, the licensee failed to implement the provisions of GNP 12.06.01, "Hot and Cold Weather Operations," which resulted in a failure to ensure pre-summer readiness of numerous safety-related and risk-significant systems. The inspectors evaluated the finding using Attachment 0609.04, of IMC 0609, Significance Determination Process, dated January 10, 2008, and answered no to all of the questions in the Initiating Events column; therefore, the finding was determined to be of very low safety significance.
The inspectors determined that the primary cause for this finding was related to the cross cutting area of human performance, work practices component, because personnel have been trained in the need for procedural use and adherence, but failed to follow applicable procedures. Specifically, the procedure which required the performance of plant inspections for hot weather operations, and the maintenance of QA documentation for these inspections, was not followed [H.4(b)] .
Inspection Report# : 2008003 (pdf)
Mitigating Systems Significance:        Mar 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Potential Debris Sources Could Clog A Drain Credited During Internal Floods A finding of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors for the failure to translate the flooding design basis into specifications, procedures, and instructions. Specifically, the licensee failed to control the storage of material in the steam generator blowdown tank room that could potentially clog a floor drain, in an adjoining room, that was credited in a flood analysis. As part of its corrective actions, the licensee removed or secured the material of concern.
The finding was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of Protection Against External Events and adversely 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). Specifically, the licensee did not put adequate controls in place to ensure that the drain would performed its credited function to be open and free flowing during an internal flood scenario involving a break in a 4-inch condensate line. The inspectors evaluated the finding using the Significance Determination Process in accordance with Inspection Manual Chapter 0609, Significance Determination Process, Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings, Table 4a, for the Mitigating Systems Cornerstone, dated January 10, 2008. The significance of the finding was determined to be of very low safety significance (Green) because the inspectors answered no to the questions in the Mitigation Systems Cornerstone column. The inspectors did not identify a cross-cutting aspect associated with this finding because the controls over material that could plug the drain should have been implemented when calculation 2005-05708 was completed and incorporated in the flooding design
 
basis in 2005; therefore, this issue was not reflective of current performance.
Inspection Report# : 2009002 (pdf)
Significance:      Mar 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Siphon Line Which Interconnected Two Diesel Generator Emergency Fuel Oil Storage Tanks Was Not Functioning as Designed A finding of very low safety significance (Green) and associated Severity Level IV, Non-Cited Violation of 10 CFR 50.59 was identified by the inspectors while reviewing Unresolved Item 05000305/2008003-03, Siphon Line Which Interconnected Two Diesel Generator Emergency Fuel Oil Storage Tanks Was Not Functioning as Designed.
Specifically, while performing Updated Safety Analysis Report change request, UCR 93 031, the licensee inappropriately screened the removal of the Updated Safety Analysis Report reference to the siphon line when plant staff incorrectly answered no to all of the 10 CFR 50.59 evaluation questions. The licensee entered this issue into its corrective action program for evaluation and development of corrective actions, as appropriate.
Because violations of 10 CFR 50.59 are considered to be violations that potentially impede or impact the regulatory process, they are dispositioned using the traditional enforcement process. As described in Supplement I of the Enforcement Policy, to determine the severity of a 10 CFR 50.59 violation, the underlying technical issue was evaluated under the Significance Determination Process. The inspectors evaluated the finding using the Significance Determination Process in accordance with Inspection Manual Chapter 0609, Significance Determination Process, 609.04, Phase 1- Initial Screening and Characterization of Findings, Table 4a, for the Mitigating Systems Cornerstone, dated January 10, 2008. The inspectors answered yes to Question 2 in the Mitigation System Cornerstone column which required the issue to be evaluated in accordance with Appendix A, of Inspection Manual Chapter 0609. Using Appendix A, the inspectors screened the issue as very low safety significance (Green) because the quantity of fuel to the diesel generators that was historically available always exceeded that needed for 24 hours of operation, thereby, resulting in the probabilistic risk assessment function for the diesels being met. The inspectors determined that the issue had a cross-cutting aspect in problem identification and resolution, corrective action program, because the licensee had identified similar deficiencies with accurately applying or interpreting the current licensing basis, and failed to take timely action to complete corrective actions, or establish barriers to prevent recurrence of this deficiency (P.1(d)).
Inspection Report# : 2009002 (pdf)
Significance:      Dec 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Emergency Diesel Generator Room Cooling Fan Testing Deficienceies
. A finding of very low safety significance and an associated NCV of 10 CFR, Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors for the licensees failure to verify the ventilation flow rate for the emergency diesel generator (EDG) rooms and for using an incorrect EDG heat load in the design basis calculation of record. As part of corrective actions, the licensee remeasured flow rates and duct dimensions and recalculated post-accident temperature values.
This finding was more than minor because a revision to the design calculation was necessary to demonstrate EDG cubicle temperatures would remain under the design basis 120 degrees Fahrenheit (&#xba;F) equipment qualification limit under any accident conditions. The finding was of very low safety significance (Green) based on a Phase 1 screening in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, dated January 10, 2008, because the revised calculation showed that the diesel generators had remained operable in all circumstances. There was no cross-cutting aspect associated with this finding.
Inspection Report# : 2008005 (pdf)
 
Significance:        Sep 30, 2008 Identified By: NRC Item Type: NCV NonCited Violation Operability Evaluation for Degraded Gauge Pedestals Failed to Adequately Evaluate Degraded Conditions Per Procedures A finding of very low safety significance and an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors during a review of an operability evaluation for degraded concrete support pads under the discharge pressure gauge pedestals for safety-related service water pumps A1 and A2. Specifically, procedure OP AA 102, Operability Determination, required that when a potential degraded or nonconforming condition is identified, action must be taken to discover the facts and confirm the condition of the systems, structures, and components. The licensees operability evaluation failed to adequately evaluate the degraded condition and failed to confirm that the compensatory actions used as a basis for operability for the pumps were effective. Corrective actions included the engineering department providing a more thorough evaluation of the potential for damage to the gauge isolation valve and associated piping from a falling gauge support including field measurements and piping configuration information.
The finding is greater than minor because the failure to perform an adequate operability evaluation, if left uncorrected, would become a more significant failure to comply with the technical specifications or the licensing basis. The significance of the finding was determined to be of very low safety significance because the inspectors answered no to all of the questions for the Mitigation Systems Cornerstone column of Attachment 0609.04, of Inspection Manual Chapter 0609, Significance Determination Process. Additionally, the inspectors attributed this issue to the cross cutting area of problem identification, corrective action program, because the operability evaluation and associated problems were not thoroughly evaluated.
Inspection Report# : 2008004 (pdf)
Significance:        Sep 30, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform a 10 CFR Part 50.59 Screening for Alteration During Maintenance that Existed for More Than 90 Days A finding of very low safety significance and an associated Severity Level IV, Non-Cited Violation of 10 CFR 50.59 was identified by the inspectors for a failure to perform a 50.59 screening for an alteration during maintenance that existed for more than 90 days. Specifically, the licensee failed to perform a 50.59 screening when spare breakers were removed from safety-related motor control centers (MCCs) and the cubicle were left in an altered state for more than 90 days. Proposed corrective actions include changes to the station housekeeping and work control/planning procedures to better evaluate job site and environmental conditions.
The finding is greater than minor because, if left uncorrected, the failure to perform a 10 CFR 50.59 screening on an alteration/change to the facility would become more significant. The inspectors determined the finding could be evaluated using the Significance Determination Process in accordance with Inspection Manual Chapter 0609, Significance Determination Process, Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings, Table 2 for the Mitigation Systems Cornerstone. Using information provided by the licensee relative to the affected MCCs, the inspectors answered no to all of the questions in this cornerstone column; therefore, the finding was determined to be of very low safety significance. Additionally, the inspectors determined that the finding has a cross-cutting aspect in the area of human performance, work control because the licensee failed to appropriately plan work activities by incorporating risk insights gained from operating experience and factor in environmental conditions during planning contingencies for systems, structures, and components anticipated to be in a maintenance condition for extensive periods of time.
Inspection Report# : 2008004 (pdf)
Significance: SL-IV Jun 30, 2007 Identified By: NRC
 
Item Type: VIO Violation Failure to Perform a 10 CFR 50.59 Evaluation for Compensatory Measures Associated with a Procedure Change The inspectors identified a finding having very low safety significance and an associated Severity Level IV, Cited Violation of 10 CFR 50.59 while reviewing unresolved items URI 05000305/2006003-04, Adequacy of Compensatory Actions for Potential Turbine Missile Strike of Control Room Ventilation Cooling; and URI 05000305/2006016-01, Adequacy of 10 CFR 50.59 Screening for Procedure Change. Specifically, the licensee failed to properly interpret design and licensing basis requirements associated with protection against external events and as a result did not perform a 10 CFR 50.59 evaluation. The cause of this finding is related to the cross-cutting area of problem identification and resolution because the licensee had similar prior problems that, if effectively evaluated and resolved, could have prevented this issue. (P.1(c))
This finding was determined to be more than minor because the inspectors determined that the procedure change would have ultimately required NRC approval. The procedure changes, in the form of compensatory operator actions, adversely impacted the operation of control room recirculation system following a tornado. A Phase 1 significance determination of this finding using IMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," using the Severe Weather Screening Criteria questions was completed. Since the loss of the control room recirculation system would not result in an initiating event or degrade two or more trains of a multi-train safety system, the issue screened as Green.
Inspection Report# : 2007003 (pdf)
Barrier Integrity Significance:        Mar 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Steam Exclusion Door Failure Results In Multiple Safety Systems Being Declared Inoperable A finding of very low safety significance (Green) and associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self revealed for the licensees failure to follow the corrective action program procedure to implement corrective actions that could have prevented a December 30, 2008, door seal failure, which rendered both trains of control room ventilation inoperable. The licensee entered this issue into its corrective action program and, as partially corrective action, has increased its monitoring of doors for potential failure mechanisms.
The finding was determined to be more than minor because it was associated with the Barrier Integrity Cornerstone attribute of Configuration Control 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. The inspectors evaluated the finding using the Significance Determination Process (SDP) in accordance with Inspection Manual Chapter (IMC) 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, dated January 10, 2008, and determined the finding represented a degradation of the barrier function to protect against radiological hazards, toxic gas, and smoke that required a Phase 3 analysis. A Region III Senior Reactor Analyst completed a qualitative Phase 3 analysis and determined that because the duration of the event was small, 44 minutes, the issue screened as having very low safety significance (Green).
The inspectors determined that the finding had a cross cutting aspect in the corrective action program component element of problem identification and resolution because the licensee failed to take appropriate corrective actions to address safety issues and adverse trends in a timely manner (P.1(d)).
Inspection Report# : 2009002 (pdf)
Significance:        Sep 30, 2008 Identified By: NRC Item Type: NCV NonCited Violation
 
Inadequate Procedure Results in Unplanned Control Rod Motion A finding of very low safety significance and an associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed when control rods automatically stepped inward unexpectedly. Ultimately, it was determined that procedures for operation of the power range nuclear instrument were found to be inadequate for the circumstances. Specifically, procedures for bypassing nuclear instrument N 43 did not contain steps to place control rods in manual when placing a failed instrument in bypass.
Corrective actions were taken to replace the inappropriately deleted steps from the associated procedures.
The finding is greater than minor in accordance with Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 20, 2007, because the finding affected the procedure quality attribute of the Barrier Integrity Cornerstone of Reactor Safety. Specifically, the failure to either leave the step for placing rods in manual in multiple alarm response procedures, or transferring the step to the common procedure OP KW AOP MISC 001, resulted in a preventable condition which resulted in an unexpected reactivity transient. The inspectors evaluated the finding using the Significance Determination Process in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings, Table 2 for the Barriers Cornerstone. The inspectors answered no to all of the questions in this cornerstone column; therefore, the finding was determined to be of very low safety significance. The inspectors concluded that the finding had a cross-cutting aspect in the area of human performance, decision-making, because interdisciplinary reviews performed by station personnel, including the on site safety review committee, failed to make changes to the various procedures using a systematic process. Additionally, the inspectors reviewed the licensee evaluation of the cause of the issue and found that it agreed with their understanding of the issue.
Inspection Report# : 2008004 (pdf)
Emergency Preparedness Significance:      Aug 29, 2008 Identified By: NRC Item Type: VIO Violation Failure to Maintain a Standard EAL Scheme An apparent violation was identified by the inspector for failure to follow and maintain in effect emergency plans which use a standard emergency classification and action level scheme. Specifically, the licensee's emergency plan Alert emergency action levels (EALs) RA1.1 and RA1.2 specified instrument setpoints that were beyond the limits of the effluent radiation monitors capabilities.
This finding was considered more than minor because the licensee is required to be capable to implement adequate measures to protect public health and safety in the event of a radiological emergency. Regulations require a standard emergency classification and action level scheme, the bases which included facility system and effluent parameters, in use by the licensee and State and local response plans call for reliance on information provided by the licensee for determination of minimum initial offsite response measures. As a result of having Alert EAL threshold values that were beyond the range of the associated effluent radiation monitors, Kewaunee personnel would not have been able to classify an emergency based upon an effluent radioactive material release in a timely manner. Emergency response actions directed by the State and local emergency response plans, which rely on information provided by the licensee, could have potentially been delayed.
The cause of the finding is related to the human performance cross-cutting element of H.2(c) for ensuring that personnel, equipment, and procedures are available and adequate to assure nuclear safety. Specifically, those necessary for complete, accurate, and up-to-date design documentation, procedures, and work packages.
Final significance determination letter was issued on October 29, 2008, and no response was required to the white violation.
Inspection Report# : 2008503 (pdf)
 
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.
Miscellaneous Significance: N/A Jun 06, 2008 Identified By: NRC Item Type: FIN Finding PI&R Inspection Report Summary On the basis of the sample selected for review, the team concluded that implementation of the corrective action program (CAP) at Kewaunee Power Station was generally good. The licensee had a low threshold for identifying problems and entering them in the CAP. Items entered into the CAP were screened and prioritized in a timely manner using established criteria; were properly evaluated commensurate with their safety significance; and corrective actions were generally implemented in a timely manner, commensurate with the safety significance. The team noted that the licensee reviewed Operating Experience (OE) for applicability to station activities. Audits and self-assessments were determined to be performed at an appropriate level to identify deficiencies. In interviews conducted during the inspection, workers at the site expressed freedom to enter safety concerns into the CAP.
Inspection Report# : 2008007 (pdf)
Last modified : May 28, 2009
 
Kewaunee 2Q/2009 Plant Inspection Findings Initiating Events Significance:      Mar 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inappropriate Application Of A Dedicated Operator During A Component Cooling Water Surveillance A finding of very low safety significance (Green) and associated Non-Cited Violation of 10 CFR 50.65(a)(4) was identified by the inspectors for the failure to properly assess risk that resulted from risk significant maintenance being performed on the component cooling water (CCW) system, when the licensee inappropriately applied criteria for the use of a dedicated operator to meet availability requirements. As part of its corrective actions, the licensee stopped work that required the use of a dedicated operator pending further evaluation.
The issue was more than minor because the licensees risk assessment for March 11, 2009, failed to consider the CCW unavailable during maintenance. Specifically, the failure to account for the unavailability of CCW resulted in an inadequate daily risk assessment and could affect the unavailability time of this system in related performance and maintenance rule indicators. The inspectors evaluated the finding using the Significance Determination Process in accordance with Inspection Manual Chapter 0609, Significance Determination Process, Attachment K, Maintenance Risk Assessment and Risk Management Significance Determination Process, dated May 19, 2005, and determined the issue screened as having very low safety significance (Green), because the incremental conditional core damage probability was less than 1E 6 due to the test condition lasting only four hours. The inspectors determined that the finding had a cross cutting aspect in the corrective action program component of problem identification and resolution, because the licensee failed to thoroughly evaluate a prior problem such that the resolution addressed the causes and extent of conditions necessary to preclude this event (P.1(c)).
Inspection Report# : 2009002 (pdf)
Significance:      Dec 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Transfer Turbine Valve Testing Requirements into the USAR A finding of very low safety significance and associated Severity Level IV Non-Cited Violation of 10 CFR 50.71, Maintenance of records, making of reports, was identified by the inspectors for the licensees failure to adequately update the Kewaunee Power Station Updated Safety Analysis Report (USAR). Specifically, the inspectors identified that the licensee had not updated the USAR completely when they relocated the turbine valve testing requirements from technical specifications to the USAR in License Amendment No. 121. Proposed corrective actions include performing an apparent cause evaluation and USAR changes as appropriate.
This finding was more than minor because it had a material impact on licensed activities in that the incorrect USAR allowed the licensee to schedule periodic testing of the reheat and interceptor valves at an interval beyond one year.
The inspectors evaluated the finding using the Significance Determination Process in accordance with Inspection Manual Chapter 0609, Significance Determination Process, Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings, Table 3b, for the Initiating Events Cornerstone, dated January 10, 2008. Using information provided by the licensee, the inspectors answered no to the transient initiator contributor questions in this cornerstone column; therefore, the finding was determined to be of very low safety significance (Green).
Additionally, the inspectors determined that the finding has a cross-cutting aspect in the area of problem identification and resolution, corrective action program, because the licensee failed to accurately identify the issue when conducting corrective actions for Condition Report CR040457, Discrepancy in Turbine Valve Testing Requirements and Acceptance Criteria, [P.1(a)].
 
Inspection Report# : 2008005 (pdf)
Mitigating Systems Significance: SL-IV Jun 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Emergency Diesel Generator Air System May Not Be Appropriately Qualified A finding of very low safety significance (Green) and an associated Non-Cited Violation of 10 CFR 50.59 were identified by the inspectors for the licensees failure to obtain a license amendment when it failed to properly assess a quality assurance typing change to the emergency diesel generator starting air compressors. Violations of 10 CFR 50.59 are considered to be violations that potentially impede or impact the regulatory process, and are dispositioned using the traditional enforcement process. The licensee entered this issue into its corrective action program as Condition Report (CR) 326432 for evaluation and development of corrective actions, as appropriate.
Supplement I of the Enforcement Policy was used to determine the severity of the underlying technical issue evaluated under the Significance Determination Process (SDP). The issue was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of design control and adversely 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 inspectors evaluated the finding using the SDP and the inspectors screened the issue as very low safety significance (Green) because the quantity of air available to supply air for five start attempts of the diesels and to supply support systems for the emergency diesel generators always exceeded that needed for 24 hours of operation, thereby, resulting in the probabilistic risk assessment function for the diesels being met. The inspectors determined this issue had a cross-cutting aspect in the area of Problem Identification and Resolution, trending and assessment, because the licensee failed to perform aggregate assessments that could have identified and prevented this and related issues earlier (P.1(b)).
Inspection Report# : 2009003 (pdf)
Significance:      Jun 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure To Ensure That Motor Control Circuit Control Circuits Have Adequate Voltage To Operate During Design Basis Accident Conditions
. A finding of very low safety significance (Green) and an associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, were identified by the inspectors for the failure to verify that motor control center (MCC) control circuits for some ventilation fans and safety injection system isolation valves would have adequate voltage to operate and, therefore, could result in a loss of function of the circuits during a design basis accident. To address this issue, the licensee modified several MCC starter circuits, which entailed replacement of some inadequately-sized control power transformers, starters, and fuses, and implemented procedures changes to reduce MCC loads.
The finding was determined to be more than minor because the calculation errors resulted in four inoperable components and a condition where there was reasonable doubt on the operability of several other safety-related loads.
The inspectors assessed the significance of this finding for each affected component and determined that the finding did not either relate to a containment structure, system, or component or containment status that had an impact on large early release frequency, or did not result in loss of operability or functionality of the safety injection system because the discharge isolation valves were aligned in their required accident positions and de-powered. In addition, the inspectors assessed the impact on the components powered from the MCCs and determined that the overall failure to ensure adequate voltage at the MCCs as having very low safety significance. Therefore, the finding screened as of very low safety significance (Green). The inspectors determined that there was no cross-cutting aspect to this finding because the cause of the problem occurred many years and was not indicative of current performance.
 
Inspection Report# : 2009003 (pdf)
Significance:        Mar 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Potential Debris Sources Could Clog A Drain Credited During Internal Floods A finding of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors for the failure to translate the flooding design basis into specifications, procedures, and instructions. Specifically, the licensee failed to control the storage of material in the steam generator blowdown tank room that could potentially clog a floor drain, in an adjoining room, that was credited in a flood analysis. As part of its corrective actions, the licensee removed or secured the material of concern.
The finding was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of Protection Against External Events and adversely 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). Specifically, the licensee did not put adequate controls in place to ensure that the drain would performed its credited function to be open and free flowing during an internal flood scenario involving a break in a 4-inch condensate line. The inspectors evaluated the finding using the Significance Determination Process in accordance with Inspection Manual Chapter 0609, Significance Determination Process, Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings, Table 4a, for the Mitigating Systems Cornerstone, dated January 10, 2008. The significance of the finding was determined to be of very low safety significance (Green) because the inspectors answered no to the questions in the Mitigation Systems Cornerstone column. The inspectors did not identify a cross-cutting aspect associated with this finding because the controls over material that could plug the drain should have been implemented when calculation 2005-05708 was completed and incorporated in the flooding design basis in 2005; therefore, this issue was not reflective of current performance.
Inspection Report# : 2009002 (pdf)
Significance:        Mar 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Siphon Line Which Interconnected Two Diesel Generator Emergency Fuel Oil Storage Tanks Was Not Functioning as Designed A finding of very low safety significance (Green) and associated Severity Level IV, Non-Cited Violation of 10 CFR 50.59 was identified by the inspectors while reviewing Unresolved Item 05000305/2008003-03, Siphon Line Which Interconnected Two Diesel Generator Emergency Fuel Oil Storage Tanks Was Not Functioning as Designed.
Specifically, while performing Updated Safety Analysis Report change request, UCR 93 031, the licensee inappropriately screened the removal of the Updated Safety Analysis Report reference to the siphon line when plant staff incorrectly answered no to all of the 10 CFR 50.59 evaluation questions. The licensee entered this issue into its corrective action program for evaluation and development of corrective actions, as appropriate.
Because violations of 10 CFR 50.59 are considered to be violations that potentially impede or impact the regulatory process, they are dispositioned using the traditional enforcement process. As described in Supplement I of the Enforcement Policy, to determine the severity of a 10 CFR 50.59 violation, the underlying technical issue was evaluated under the Significance Determination Process. The inspectors evaluated the finding using the Significance Determination Process in accordance with Inspection Manual Chapter 0609, Significance Determination Process, 609.04, Phase 1- Initial Screening and Characterization of Findings, Table 4a, for the Mitigating Systems Cornerstone, dated January 10, 2008. The inspectors answered yes to Question 2 in the Mitigation System Cornerstone column which required the issue to be evaluated in accordance with Appendix A, of Inspection Manual Chapter 0609. Using Appendix A, the inspectors screened the issue as very low safety significance (Green) because the quantity of fuel to the diesel generators that was historically available always exceeded that needed for 24 hours of operation, thereby, resulting in the probabilistic risk assessment function for the diesels being met. The inspectors determined that the issue had a cross-cutting aspect in problem identification and resolution, corrective action
 
program, because the licensee had identified similar deficiencies with accurately applying or interpreting the current licensing basis, and failed to take timely action to complete corrective actions, or establish barriers to prevent recurrence of this deficiency (P.1(d)).
Inspection Report# : 2009002 (pdf)
Significance:        Dec 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Emergency Diesel Generator Room Cooling Fan Testing Deficienceies A finding of very low safety significance and an associated NCV of 10 CFR, Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors for the licensees failure to verify the ventilation flow rate for the emergency diesel generator (EDG) rooms and for using an incorrect EDG heat load in the design basis calculation of record. As part of corrective actions, the licensee remeasured flow rates and duct dimensions and recalculated post-accident temperature values.
This finding was more than minor because a revision to the design calculation was necessary to demonstrate EDG cubicle temperatures would remain under the design basis 120 degrees Fahrenheit (&#xba;F) equipment qualification limit under any accident conditions. The finding was of very low safety significance (Green) based on a Phase 1 screening in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, dated January 10, 2008, because the revised calculation showed that the diesel generators had remained operable in all circumstances. There was no cross-cutting aspect associated with this finding.
Inspection Report# : 2008005 (pdf)
Significance:        Sep 30, 2008 Identified By: NRC Item Type: NCV NonCited Violation Operability Evaluation for Degraded Gauge Pedestals Failed to Adequately Evaluate Degraded Conditions Per Procedures A finding of very low safety significance and an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors during a review of an operability evaluation for degraded concrete support pads under the discharge pressure gauge pedestals for safety-related service water pumps A1 and A2. Specifically, procedure OP AA 102, Operability Determination, required that when a potential degraded or nonconforming condition is identified, action must be taken to discover the facts and confirm the condition of the systems, structures, and components. The licensees operability evaluation failed to adequately evaluate the degraded condition and failed to confirm that the compensatory actions used as a basis for operability for the pumps were effective. Corrective actions included the engineering department providing a more thorough evaluation of the potential for damage to the gauge isolation valve and associated piping from a falling gauge support including field measurements and piping configuration information.
The finding is greater than minor because the failure to perform an adequate operability evaluation, if left uncorrected, would become a more significant failure to comply with the technical specifications or the licensing basis. The significance of the finding was determined to be of very low safety significance because the inspectors answered no to all of the questions for the Mitigation Systems Cornerstone column of Attachment 0609.04, of Inspection Manual Chapter 0609, Significance Determination Process. Additionally, the inspectors attributed this issue to the cross cutting area of problem identification, corrective action program, because the operability evaluation and associated problems were not thoroughly evaluated.
Inspection Report# : 2008004 (pdf)
Significance:        Sep 30, 2008
 
Identified By: NRC Item Type: NCV NonCited Violation Failure to Perform a 10 CFR Part 50.59 Screening for Alteration During Maintenance that Existed for More Than 90 Days A finding of very low safety significance and an associated Severity Level IV, Non-Cited Violation of 10 CFR 50.59 was identified by the inspectors for a failure to perform a 50.59 screening for an alteration during maintenance that existed for more than 90 days. Specifically, the licensee failed to perform a 50.59 screening when spare breakers were removed from safety-related motor control centers (MCCs) and the cubicle were left in an altered state for more than 90 days. Proposed corrective actions include changes to the station housekeeping and work control/planning procedures to better evaluate job site and environmental conditions.
The finding is greater than minor because, if left uncorrected, the failure to perform a 10 CFR 50.59 screening on an alteration/change to the facility would become more significant. The inspectors determined the finding could be evaluated using the Significance Determination Process in accordance with Inspection Manual Chapter 0609, Significance Determination Process, Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings, Table 2 for the Mitigation Systems Cornerstone. Using information provided by the licensee relative to the affected MCCs, the inspectors answered no to all of the questions in this cornerstone column; therefore, the finding was determined to be of very low safety significance. Additionally, the inspectors determined that the finding has a cross-cutting aspect in the area of human performance, work control because the licensee failed to appropriately plan work activities by incorporating risk insights gained from operating experience and factor in environmental conditions during planning contingencies for systems, structures, and components anticipated to be in a maintenance condition for extensive periods of time.
Inspection Report# : 2008004 (pdf)
Barrier Integrity Significance:      Jun 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failed Backdraft Damper Renders Containment Fan Coil Unit Inoperable
. A finding of very low safety significance (Green) and an associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, were self revealed for the licensees failure to maintain adequate procedures for the inspection and verification-of-operation for the A containment fan coil unit backdraft dampers. The licensee entered this issue into the licensees corrective action program as Condition Report (CR) 328191; immediate corrective actions were accomplished to repair the affected components.
The finding was determined to be more than minor because the finding was associated with the Barrier Integrity Cornerstone attribute of design control and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers, specifically containment, protect the public from radionuclide releases caused by accidents or events. The inspectors determined the finding could be evaluated using the Significance Determination Process (SDP) and screened the finding as having very low safety significance (Green). This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, operating experience, because the licensee did not use operating experience to support plant safety (P.2(b)).
Inspection Report# : 2009003 (pdf)
Significance:      Jun 30, 2009 Identified By: NRC Item Type: FIN Finding Failure To Update Procedures As Required By Commitments A finding of very low safety significance (Green) was identified by the inspectors for the licensees failure to update
 
procedures as required by NRC commitments. Specifically, a procedure for fuel oil sampling and a procedure for steam generator tube inspections were not maintained as required by the referenced commitments. The inspectors determined that the issues constituted a finding relating to management of commitments as required by Nuclear Energy Institute 99-04, Guidelines for Managing NRC Commitment Changes. The licensee has entered this issue into its corrective action program as Condition Report (CR) 340864 to assess the failure to the effects of revisions to reference and end-use documents on each other.
The inspectors concluded that the issue was more than minor because the integration of vendor/industry guidance was related to a commitment to the NRC for steam generator tube inspections, and the failure to appropriately manage the commitments impacted the regulatory process. The issue was administrative in nature and did not impact any safety or risk significant systems, therefore, the issue was determined to be of very low safety significance (Green). The inspectors determined that the issue had a cross cutting aspect related in the area of Human Performance, resources, because the licensee failed to maintain the related procedures complete, accurate, and up to date (H.2(c)).
Inspection Report# : 2009003 (pdf)
Significance:        Mar 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Steam Exclusion Door Failure Results In Multiple Safety Systems Being Declared Inoperable A finding of very low safety significance (Green) and associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed for the licensees failure to follow the corrective action program procedure to implement corrective actions that could have prevented a December 30, 2008, door seal failure, which rendered both trains of control room ventilation inoperable. The licensee entered this issue into its corrective action program and, as partially corrective action, has increased its monitoring of doors for potential failure mechanisms.
The finding was determined to be more than minor because it was associated with the Barrier Integrity Cornerstone attribute of Configuration Control 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. The inspectors evaluated the finding using the Significance Determination Process (SDP) in accordance with Inspection Manual Chapter (IMC) 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, dated January 10, 2008, and determined the finding represented a degradation of the barrier function to protect against radiological hazards, toxic gas, and smoke that required a Phase 3 analysis. A Region III Senior Reactor Analyst completed a qualitative Phase 3 analysis and determined that because the duration of the event was small, 44 minutes, the issue screened as having very low safety significance (Green).
The inspectors determined that the finding had a cross-cutting aspect in the corrective action program component element of problem identification and resolution because the licensee failed to take appropriate corrective actions to address safety issues and adverse trends in a timely manner (P.1(d)).
Inspection Report# : 2009002 (pdf)
Significance:        Sep 30, 2008 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedure Results in Unplanned Control Rod Motion A finding of very low safety significance and an associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed when control rods automatically stepped inward unexpectedly. Ultimately, it was determined that procedures for operation of the power range nuclear instrument were found to be inadequate for the circumstances. Specifically, procedures for bypassing nuclear instrument N 43 did not contain steps to place control rods in manual when placing a failed instrument in bypass.
Corrective actions were taken to replace the inappropriately deleted steps from the associated procedures.
The finding is greater than minor in accordance with Inspection Manual Chapter (IMC) 0612, Power Reactor
 
Inspection Reports, Appendix B, Issue Screening, dated September 20, 2007, because the finding affected the procedure quality attribute of the Barrier Integrity Cornerstone of Reactor Safety. Specifically, the failure to either leave the step for placing rods in manual in multiple alarm response procedures, or transferring the step to the common procedure OP KW AOP MISC 001, resulted in a preventable condition which resulted in an unexpected reactivity transient. The inspectors evaluated the finding using the Significance Determination Process in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings, Table 2 for the Barriers Cornerstone. The inspectors answered no to all of the questions in this cornerstone column; therefore, the finding was determined to be of very low safety significance. The inspectors concluded that the finding had a cross-cutting aspect in the area of human performance, decision-making, because interdisciplinary reviews performed by station personnel, including the on site safety review committee, failed to make changes to the various procedures using a systematic process. Additionally, the inspectors reviewed the licensee evaluation of the cause of the issue and found that it agreed with their understanding of the issue.
Inspection Report# : 2008004 (pdf)
Emergency Preparedness Significance:      Jun 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Seismic Monitoring System Repeatedly Fails Surveillance A finding of very low safety significance (Green) and an associated Non-Cited Violation were identified by the inspectors for the licensees failure to maintain radiation monitoring instrumentation operable that was required by its emergency plans to meet the standards set forth in 10 CFR 50.47(b). Specifically, seismic instrumentation needed for two Emergency Action Levels, HU1.1 and HA1.1, was not maintained operable such that a related Unusual Event notification and an Alert declaration could have been made under certain conditions. Corrective actions were taken for this issue and included revising Emergency Action Level (EAL) requirements to values within the range of the instrumentation.
The inspectors determined that the issue was more than minor because the finding, if left uncorrected, would become a more significant safety concern. Specifically, for a seismic event, the deficiency could lead to the failure to declare an Unusual Event for a Natural and Destructive Phenomena Affecting the Plant Protected Area, HU1.1, and an Alert for a Natural and Destructive Phenomena Affecting the Plant Vital Area, HA1.1. The inspectors determined the finding could be evaluated using the Significance Determincation Process (SDP) and concluded that the risk significant planning standard problem was not a functional failure, nor did it represent a degraded function and, therefore, screened as an issue of very low safety significance (Green). The inspectors determined this was a Green risk significant planning standard problem, rather than degraded or failed risk significant planning standard function, because the process failure affected only one Unusual Event and one Alert emergency classification. The inspectors determined this issue had a cross-cutting aspect in the area of Problem Identification and Resolution, trending and assessment, because the licensee failed to perform aggregate assessments that could have identified and prevented this and related issues earlier (P.1(b)).
Inspection Report# : 2009003 (pdf)
Significance:      Aug 29, 2008 Identified By: NRC Item Type: VIO Violation Failure to Maintain a Standard EAL Scheme An apparent violation was identified by the inspector for failure to follow and maintain in effect emergency plans which use a standard emergency classification and action level scheme. Specifically, the licensee's emergency plan Alert emergency action levels (EALs) RA1.1 and RA1.2 specified instrument setpoints that were beyond the limits of the effluent radiation monitors capabilities.
 
This finding was considered more than minor because the licensee is required to be capable to implement adequate measures to protect public health and safety in the event of a radiological emergency. Regulations require a standard emergency classification and action level scheme, the bases which included facility system and effluent parameters, in use by the licensee and State and local response plans call for reliance on information provided by the licensee for determination of minimum initial offsite response measures. As a result of having Alert EAL threshold values that were beyond the range of the associated effluent radiation monitors, Kewaunee personnel would not have been able to classify an emergency based upon an effluent radioactive material release in a timely manner. Emergency response actions directed by the State and local emergency response plans, which rely on information provided by the licensee, could have potentially been delayed.
The cause of the finding is related to the human performance cross-cutting element of H.2(c) for ensuring that personnel, equipment, and procedures are available and adequate to assure nuclear safety. Specifically, those necessary for complete, accurate, and up-to-date design documentation, procedures, and work packages.
Final significance determination letter was issued on October 29, 2008, and no response was required to the white violation.
Inspection Report# : 2008503 (pdf)
Inspection Report# : 2009503 (pdf)
Occupational Radiation Safety Public Radiation Safety Significance:      Jun 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Calibration Of Radiation Monitor R-19 A finding of very low safety significance (Green) and an associated Non-Cited Violation of Technical Specification 4.1 were identified by the inspectors for the licensees calibration practices for process radiation instrument R-19 that did not qualify as a Channel Calibration as required by technical specifications. Specifically, the sources for calibration of R-19 were not of sufficient strength to test the instrument in the range where alarms were required to be operable. Corrective actions were taken for this issue and included revising Emergency Action Level (EAL) requirements to values within the range of the instrumentation.
The inspectors determined that the issue was more than minor because it was associated with the Public Radiation Safety Cornerstone attribute of equipment and instrumentation and adversely affected the cornerstone objective to ensure protection of public health and safety from exposure to radioactive materials released into the public domain.
The inspectors used Inspection Manual Chapter (IMC) 0609, Significance Determination Process, Attachment D, Public Radiation Safety Significance Determination Process, dated February 12, 2008, and determined that the finding was in the licensees radiological effluent monitoring program and was contrary to a technical specification requirement. However, the finding was not related to a spill or release of radioactive material to the environment and, therefore, screened as an issue of very low safety significance (Green). The inspectors reviewed this issue for a cross-cutting aspect and determined that no cross-cutting aspect was applicable.
Inspection Report# : 2009003 (pdf)
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.
Miscellaneous Last modified : August 31, 2009
 
Kewaunee 3Q/2009 Plant Inspection Findings Initiating Events Significance:        Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure To Adequately Analyze The Automatic Fast Transfer Feature That Allowed Operation With Both 4.16-kiloVolt Safety-Related Buses 1-5 And 1-6 Connected To The Reserve Auxiliary Transformer A finding of very low safety-significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," was identified by the inspectors for the failure to perform a power system analysis calculation that would have identified that the fast transfer design feature/scheme was deficient, in that, it allowed an unanalyzed electrical power system alignment where both redundant 4.16-kiloVolt safety-related buses were being supplied by an offsite source via the same transformer. Use of this electrical configuration could have resulted in an out-of-phase transfer, loss of available offsite power to the buses and potential damaging effects on redundant safety related equipment, during a design basis event such as initiation of safety injection signal. When identified, the licensee entered this issue into their corrective action program and implemented interim actions to prohibit use of the fast transfer feature or manually aligning two safety-related buses to be fed from the same transformer during plant operation.
This performance deficiency was more than minor because the failure to perform the required calculation resulted in a condition where the plant was being operated in an unanalyzed configuration where there was reasonable doubt as to the operability of redundant safeguard loads; this concern resulted in issuance of a Licensee Event Report 2007-007-00 on May 21, 2007. Consequently, the potential for damage or loss of power to safety-related loads during an event could have led to unacceptable consequences. The finding screened as being of very low safety-significance (Green) for the Initiating Events Cornerstone because the finding did not contribute to both the likelihood of a reactor trip and the likelihood that mitigating equipment or functions will not be available. The inspectors did not identify a cross cutting aspect associated with this finding because the cause of the performance deficiency was related to a historical design issue and not indicative of current licensee performance.
Inspection Report# : 2009004 (pdf)
Significance:        Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Design Analysis For 105-Ton Transfer Cask Lifting Beam A finding of very low safety significance and associated Non-Cited Violation of Title 10 Code of Federal Regulations Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors for the licensees failure to provide an adequate single failure proof design basis analysis for the 105-ton transfer cask-lifting beam. The licensee entered this issue into their corrective action program as condition report CR339267. The licensee revised the design calculation for the 105-ton transfer cask-lifting beam and demonstrated compliance with single failure proof acceptance criteria.
The finding was determined to be more than minor because the finding was associated with the Initiating Events Cornerstone attribute of equipment performance 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. The finding was determined to be of very low safety significance by the NRCs significance determination process because the transfer cask-lifting beam had not been previously used at the Kewaunee Power Station. This finding has a cross-cutting aspect in the area of human performance, work practices, because the licensee did not ensure supervisory and management oversight of work activities, including contractors, such that nuclear safety is supported, in that, the licensee failed to perform an effective owners review to assure that appropriate design
 
methods are used in calculations that demonstrate nuclear safety (H.4(c)).
Inspection Report# : 2009004 (pdf)
Significance:      Mar 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inappropriate Application Of A Dedicated Operator During A Component Cooling Water Surveillance A finding of very low safety significance (Green) and associated Non-Cited Violation of 10 CFR 50.65(a)(4) was identified by the inspectors for the failure to properly assess risk that resulted from risk significant maintenance being performed on the component cooling water (CCW) system, when the licensee inappropriately applied criteria for the use of a dedicated operator to meet availability requirements. As part of its corrective actions, the licensee stopped work that required the use of a dedicated operator pending further evaluation.
The issue was more than minor because the licensees risk assessment for March 11, 2009, failed to consider the CCW unavailable during maintenance. Specifically, the failure to account for the unavailability of CCW resulted in an inadequate daily risk assessment and could affect the unavailability time of this system in related performance and maintenance rule indicators. The inspectors evaluated the finding using the Significance Determination Process in accordance with Inspection Manual Chapter 0609, Significance Determination Process, Attachment K, Maintenance Risk Assessment and Risk Management Significance Determination Process, dated May 19, 2005, and determined the issue screened as having very low safety significance (Green), because the incremental conditional core damage probability was less than 1E 6 due to the test condition lasting only four hours. The inspectors determined that the finding had a cross cutting aspect in the corrective action program component of problem identification and resolution, because the licensee failed to thoroughly evaluate a prior problem such that the resolution addressed the causes and extent of conditions necessary to preclude this event (P.1(c)).
Inspection Report# : 2009002 (pdf)
Significance:      Dec 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Failure to Transfer Turbine Valve Testing Requirements into the USAR A finding of very low safety significance and associated Severity Level IV Non-Cited Violation of 10 CFR 50.71, Maintenance of records, making of reports, was identified by the inspectors for the licensees failure to adequately update the Kewaunee Power Station Updated Safety Analysis Report (USAR). Specifically, the inspectors identified that the licensee had not updated the USAR completely when they relocated the turbine valve testing requirements from technical specifications to the USAR in License Amendment No. 121. Proposed corrective actions include performing an apparent cause evaluation and USAR changes as appropriate.
This finding was more than minor because it had a material impact on licensed activities in that the incorrect USAR allowed the licensee to schedule periodic testing of the reheat and interceptor valves at an interval beyond one year.
The inspectors evaluated the finding using the Significance Determination Process in accordance with Inspection Manual Chapter 0609, Significance Determination Process, Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings, Table 3b, for the Initiating Events Cornerstone, dated January 10, 2008. Using information provided by the licensee, the inspectors answered no to the transient initiator contributor questions in this cornerstone column; therefore, the finding was determined to be of very low safety significance (Green).
Additionally, the inspectors determined that the finding has a cross-cutting aspect in the area of problem identification and resolution, corrective action program, because the licensee failed to accurately identify the issue when conducting corrective actions for Condition Report CR040457, Discrepancy in Turbine Valve Testing Requirements and Acceptance Criteria, [P.1(a)].
Inspection Report# : 2008005 (pdf)
 
Mitigating Systems Significance:      Aug 20, 2009 Identified By: NRC Item Type: NCV NonCited Violation Improper Application of 440Vac Rated Motors The inspectors identified a finding of very low safety-significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to ensure the proper application of safety-related 440Vac motors.
Specifically, eight 440Vac safety-related motors were not suitable for operation at analyzed voltages. This finding was entered into the licensees corrective action program.
The finding was more than minor because if left uncorrected it could result in the loss of safety-related 440Vac motors by overstressing of the motor windings through exposure to higher than design rated voltages, and in the failure of motor drive components caused by increased torque produced at the higher voltages. The finding was determined to be of very low safety-significance (Green) because it did not result in a loss of operability. The cause of this finding is related to the cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, because the licensee did not identify this issue completely, accurately, and in a timely manner. The values were produced in a calculation but the licensee did not identify that they exceeded the acceptance criteria. (P.1(a)) (Section 1R21.3)
Inspection Report# : 2009006 (pdf)
Significance:      Aug 20, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inaccurate Minimum Low Head Safety Injection Flow Specified in Emergency Operating Procedure The inspectors identified a finding of very low safety-significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to specify the appropriate quantitative acceptance criterion to assure that adequate Emergency Core Cooling System flow would be delivered to the core following switchover to containment sump recirculation. This finding was entered into the licensees corrective action program.
The finding was determined to be more than minor because the licensee failed to include the appropriate quantitative set-point value for the minimum low-head safety injection train flow following switchover to containment sump recirculation to assure sufficient reactor coolant was available. This finding is of very low safety-significance (Green) because it did not result in a loss of operability, did not represent an actual loss of safety function, and is not potentially risk-significant due to external events. The cause of this finding is related to the cross-cutting aspect in the area of Human Performance, Work Practices, because the licensee did not ensure proper supervisory and management oversight of contractor work activities. Vendor calculations were used as the basis for an EOP set-point without taking into account specific plant design information such as instrument uncertainties, flow instrument calibration effects, and RHR minimum flow. (H.4(c)) (Section 1R21.3)
Inspection Report# : 2009006 (pdf)
Significance:      Aug 20, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate procedure for a Battery Room Flooding Event.
A finding of very low safety significance and associated Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the failure to provide adequate procedural direction to respond to a rupture of the service water piping in the battery rooms. As part of its corrective actions, the licensee revised OP-KW-AOP-MDS-001, Abnormal Operation of Miscellaneous Drains and Sumps, to correct the inadequate operator actions.
The finding was determined to be more than minor because the licensee failed to provide adequate procedural
 
direction for a battery room A or B flood caused by a rupture of the SW piping to/from the battery room fan coil unit in the affected battery room, which ensured the protection of the battery in the unaffected room not associated with the initial flooding event. This finding is of very low safety significance (Green) because it did not result in a loss of operability, did not represent an actual loss of safety function, and is not potentially risk-significant due to external events. The cause of this finding is related to the cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, because the licensee did not fully evaluate the battery room flooding event (an issue potentially impacting nuclear safety) such that the resolution addressed causes, and extent of condition as necessary, to assure nuclear safety. (P.1(c)). (Section 1R21.6).
Inspection Report# : 2009006 (pdf)
Significance: SL-IV Jun 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Emergency Diesel Generator Air System May Not Be Appropriately Qualified A finding of very low safety significance (Green) and an associated Non-Cited Violation of 10 CFR 50.59 were identified by the inspectors for the licensees failure to obtain a license amendment when it failed to properly assess a quality assurance typing change to the emergency diesel generator starting air compressors. Violations of 10 CFR 50.59 are considered to be violations that potentially impede or impact the regulatory process, and are dispositioned using the traditional enforcement process. The licensee entered this issue into its corrective action program as Condition Report (CR) 326432 for evaluation and development of corrective actions, as appropriate.
Supplement I of the Enforcement Policy was used to determine the severity of the underlying technical issue evaluated under the Significance Determination Process (SDP). The issue was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of design control and adversely 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 inspectors evaluated the finding using the SDP and the inspectors screened the issue as very low safety significance (Green) because the quantity of air available to supply air for five start attempts of the diesels and to supply support systems for the emergency diesel generators always exceeded that needed for 24 hours of operation, thereby, resulting in the probabilistic risk assessment function for the diesels being met. The inspectors determined this issue had a cross-cutting aspect in the area of Problem Identification and Resolution, trending and assessment, because the licensee failed to perform aggregate assessments that could have identified and prevented this and related issues earlier (P.1(b)).
Inspection Report# : 2009003 (pdf)
Significance:      Jun 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure To Ensure That Motor Control Circuit Control Circuits Have Adequate Voltage To Operate During Design Basis Accident Conditions
. A finding of very low safety significance (Green) and an associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, were identified by the inspectors for the failure to verify that motor control center (MCC) control circuits for some ventilation fans and safety injection system isolation valves would have adequate voltage to operate and, therefore, could result in a loss of function of the circuits during a design basis accident. To address this issue, the licensee modified several MCC starter circuits, which entailed replacement of some inadequately-sized control power transformers, starters, and fuses, and implemented procedures changes to reduce MCC loads.
The finding was determined to be more than minor because the calculation errors resulted in four inoperable components and a condition where there was reasonable doubt on the operability of several other safety-related loads.
The inspectors assessed the significance of this finding for each affected component and determined that the finding did not either relate to a containment structure, system, or component or containment status that had an impact on large early release frequency, or did not result in loss of operability or functionality of the safety injection system because the discharge isolation valves were aligned in their required accident positions and de-powered. In addition, the inspectors assessed the impact on the components powered from the MCCs and determined that the overall failure to ensure adequate voltage at the MCCs as having very low safety significance. Therefore, the finding screened as of
 
very low safety significance (Green). The inspectors determined that there was no cross-cutting aspect to this finding because the cause of the problem occurred many years and was not indicative of current performance.
Inspection Report# : 2009003 (pdf)
Significance:        Mar 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Potential Debris Sources Could Clog A Drain Credited During Internal Floods A finding of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors for the failure to translate the flooding design basis into specifications, procedures, and instructions. Specifically, the licensee failed to control the storage of material in the steam generator blowdown tank room that could potentially clog a floor drain, in an adjoining room, that was credited in a flood analysis. As part of its corrective actions, the licensee removed or secured the material of concern.
The finding was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of Protection Against External Events and adversely 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). Specifically, the licensee did not put adequate controls in place to ensure that the drain would performed its credited function to be open and free flowing during an internal flood scenario involving a break in a 4-inch condensate line. The inspectors evaluated the finding using the Significance Determination Process in accordance with Inspection Manual Chapter 0609, Significance Determination Process, Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings, Table 4a, for the Mitigating Systems Cornerstone, dated January 10, 2008. The significance of the finding was determined to be of very low safety significance (Green) because the inspectors answered no to the questions in the Mitigation Systems Cornerstone column. The inspectors did not identify a cross-cutting aspect associated with this finding because the controls over material that could plug the drain should have been implemented when calculation 2005-05708 was completed and incorporated in the flooding design basis in 2005; therefore, this issue was not reflective of current performance.
Inspection Report# : 2009002 (pdf)
Significance:        Mar 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Siphon Line Which Interconnected Two Diesel Generator Emergency Fuel Oil Storage Tanks Was Not Functioning as Designed A finding of very low safety significance (Green) and associated Severity Level IV, Non-Cited Violation of 10 CFR 50.59 was identified by the inspectors while reviewing Unresolved Item 05000305/2008003-03, Siphon Line Which Interconnected Two Diesel Generator Emergency Fuel Oil Storage Tanks Was Not Functioning as Designed.
Specifically, while performing Updated Safety Analysis Report change request, UCR 93 031, the licensee inappropriately screened the removal of the Updated Safety Analysis Report reference to the siphon line when plant staff incorrectly answered no to all of the 10 CFR 50.59 evaluation questions. The licensee entered this issue into its corrective action program for evaluation and development of corrective actions, as appropriate.
Because violations of 10 CFR 50.59 are considered to be violations that potentially impede or impact the regulatory process, they are dispositioned using the traditional enforcement process. As described in Supplement I of the Enforcement Policy, to determine the severity of a 10 CFR 50.59 violation, the underlying technical issue was evaluated under the Significance Determination Process. The inspectors evaluated the finding using the Significance Determination Process in accordance with Inspection Manual Chapter 0609, Significance Determination Process, 609.04, Phase 1- Initial Screening and Characterization of Findings, Table 4a, for the Mitigating Systems Cornerstone, dated January 10, 2008. The inspectors answered yes to Question 2 in the Mitigation System Cornerstone column which required the issue to be evaluated in accordance with Appendix A, of Inspection Manual Chapter 0609. Using Appendix A, the inspectors screened the issue as very low safety significance (Green) because the quantity of fuel to the diesel generators that was historically available always exceeded that needed for 24 hours of operation, thereby, resulting in the probabilistic risk assessment function for the diesels being met. The inspectors
 
determined that the issue had a cross-cutting aspect in problem identification and resolution, corrective action program, because the licensee had identified similar deficiencies with accurately applying or interpreting the current licensing basis, and failed to take timely action to complete corrective actions, or establish barriers to prevent recurrence of this deficiency (P.1(d)).
Inspection Report# : 2009002 (pdf)
Significance:      Dec 31, 2008 Identified By: NRC Item Type: NCV NonCited Violation Emergency Diesel Generator Room Cooling Fan Testing Deficienceies A finding of very low safety significance and an associated NCV of 10 CFR, Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors for the licensees failure to verify the ventilation flow rate for the emergency diesel generator (EDG) rooms and for using an incorrect EDG heat load in the design basis calculation of record. As part of corrective actions, the licensee remeasured flow rates and duct dimensions and recalculated post-accident temperature values.
This finding was more than minor because a revision to the design calculation was necessary to demonstrate EDG cubicle temperatures would remain under the design basis 120 degrees Fahrenheit (&#xba;F) equipment qualification limit under any accident conditions. The finding was of very low safety significance (Green) based on a Phase 1 screening in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, dated January 10, 2008, because the revised calculation showed that the diesel generators had remained operable in all circumstances. There was no cross-cutting aspect associated with this finding.
Inspection Report# : 2008005 (pdf)
Barrier Integrity Significance:      Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Containment Isolation Valve Inoperable With No Technical Specification Action requirement Entry A finding of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the licensees failure to have adequate procedures that ensured technical specifications were entered and followed for containment isolation valves. The licensee entered the issue into their corrective action program as Condition Report 344856 and Condition Report 350526A, and provided additional guidance to operations personnel. At the end of the inspection period, the licensee continued to perform a causal analysis.
The inspectors determined that the issue was more than minor because the finding, if left uncorrected, would become a more significant safety concern. Specifically, not entering the appropriate technical specification action requirements, when necessary, would lead to more significant safety concerns. The inspectors determined the finding could be evaluated using the Significance Determination Process in accordance with Inspection Manual Chapter 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of findings, Table 4a for the Barrier Integrity Cornerstone. The inspectors answered no to the Barrier Integrity questions and screened the finding as having very low safety significance (Green). The finding has a cross-cutting aspect in the area of human performance, resources, because the licensee did not have complete, accurate and up-to-date design documentation, procedures and work packages (H.2(c)).
Inspection Report# : 2009004 (pdf)
Significance:      Aug 20, 2009
 
Identified By: NRC Item Type: NCV NonCited Violation Non-Conservative Main Steam Line Break Analysis The inspectors identified a finding of very low safety significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to correctly translate the design bases for the maximum steam generator narrow range level into procedures and instructions. This finding was entered into the licensees corrective action program.
The finding was determined to be more than minor because an evaluation was required to ensure that accident analysis requirements for peak containment pressure were met. The finding also impacted the Barrier Integrity cornerstone attribute of procedure quality, and affected the cornerstone objective of maintaining the functionality of containment to protect the public from radionuclide releases caused by accidents or events. Procedural guidance was not adequate to maintain the plant within the parameters specified in the analysis for containment operability. The finding screened as having very low safety significance (Green) because there was no actual barrier degradation. The inspectors determined there was no cross-cutting aspect associated with this finding. (Section 1R21.4)
Inspection Report# : 2009006 (pdf)
Significance: SL-IV Aug 20, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate 50.59 Evaluation of Shutdown Loss of Coolant Accident Procedure.
The inspectors identified a finding of very low safety significance and associated Non-Cited Violation of 10 CFR 50.59(d)(1) for the licensees failure to perform an adequate review of an abnormal operating procedure associated with a shutdown loss of coolant accident. As part of its corrective actions, the licensee revised procedure OP-KW-AOP-RHR-002 to remove the procedure applicability to the Cold Shutdown mode and Refueling mode with reactor vessel head on.
The inspectors determined that the finding was more than minor because it could not reasonably be determined that the activity would not ultimately have required NRC approval. Operation in accordance with the procedure may have challenged the reactor coolant system barrier. The inspectors determined that the finding did not require a quantitative assessment per IMC 0609, Appendix G. Therefore, the finding screened as having very low safety significance (Green) and was determined to be a Severity Level IV violation. The cause of this finding is related to the cross-cutting aspect in the area of Human Performance, Decision Making, because the licensee failed to use conservative assumptions in decision making to demonstrate that the proposed action to include additional modes of applicability for the Shutdown LOCA procedure was safe in order to proceed. (H.1(b)) (Section 1R21.6)
Inspection Report# : 2009006 (pdf)
Significance:      Jun 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failed Backdraft Damper Renders Containment Fan Coil Unit Inoperable
. A finding of very low safety significance (Green) and an associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, were self revealed for the licensees failure to maintain adequate procedures for the inspection and verification-of-operation for the A containment fan coil unit backdraft dampers. The licensee entered this issue into the licensees corrective action program as Condition Report (CR) 328191; immediate corrective actions were accomplished to repair the affected components.
The finding was determined to be more than minor because the finding was associated with the Barrier Integrity Cornerstone attribute of design control and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers, specifically containment, protect the public from radionuclide releases caused by accidents or events. The inspectors determined the finding could be evaluated using the Significance Determination Process (SDP) and screened the finding as having very low safety significance (Green). This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, operating experience, because the licensee did not use operating experience to support plant safety (P.2(b)).
Inspection Report# : 2009003 (pdf)
 
Significance:        Jun 30, 2009 Identified By: NRC Item Type: FIN Finding Failure To Update Procedures As Required By Commitments A finding of very low safety significance (Green) was identified by the inspectors for the licensees failure to update procedures as required by NRC commitments. Specifically, a procedure for fuel oil sampling and a procedure for steam generator tube inspections were not maintained as required by the referenced commitments. The inspectors determined that the issues constituted a finding relating to management of commitments as required by Nuclear Energy Institute 99-04, Guidelines for Managing NRC Commitment Changes. The licensee has entered this issue into its corrective action program as Condition Report (CR) 340864 to assess the failure to the effects of revisions to reference and end-use documents on each other.
The inspectors concluded that the issue was more than minor because the integration of vendor/industry guidance was related to a commitment to the NRC for steam generator tube inspections, and the failure to appropriately manage the commitments impacted the regulatory process. The issue was administrative in nature and did not impact any safety or risk significant systems, therefore, the issue was determined to be of very low safety significance (Green). The inspectors determined that the issue had a cross cutting aspect related in the area of Human Performance, resources, because the licensee failed to maintain the related procedures complete, accurate, and up to date (H.2(c)).
Inspection Report# : 2009003 (pdf)
Significance:        Mar 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Steam Exclusion Door Failure Results In Multiple Safety Systems Being Declared Inoperable A finding of very low safety significance (Green) and associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed for the licensees failure to follow the corrective action program procedure to implement corrective actions that could have prevented a December 30, 2008, door seal failure, which rendered both trains of control room ventilation inoperable. The licensee entered this issue into its corrective action program and, as partially corrective action, has increased its monitoring of doors for potential failure mechanisms.
The finding was determined to be more than minor because it was associated with the Barrier Integrity Cornerstone attribute of Configuration Control 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. The inspectors evaluated the finding using the Significance Determination Process (SDP) in accordance with Inspection Manual Chapter (IMC) 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, dated January 10, 2008, and determined the finding represented a degradation of the barrier function to protect against radiological hazards, toxic gas, and smoke that required a Phase 3 analysis. A Region III Senior Reactor Analyst completed a qualitative Phase 3 analysis and determined that because the duration of the event was small, 44 minutes, the issue screened as having very low safety significance (Green).
The inspectors determined that the finding had a cross-cutting aspect in the corrective action program component element of problem identification and resolution because the licensee failed to take appropriate corrective actions to address safety issues and adverse trends in a timely manner (P.1(d)).
Inspection Report# : 2009002 (pdf)
Emergency Preparedness Significance:        Jun 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation
 
Seismic Monitoring System Repeatedly Fails Surveillance A finding of very low safety significance (Green) and an associated Non-Cited Violation were identified by the inspectors for the licensees failure to maintain radiation monitoring instrumentation operable that was required by its emergency plans to meet the standards set forth in 10 CFR 50.47(b). Specifically, seismic instrumentation needed for two Emergency Action Levels, HU1.1 and HA1.1, was not maintained operable such that a related Unusual Event notification and an Alert declaration could have been made under certain conditions. Corrective actions were taken for this issue and included revising Emergency Action Level (EAL) requirements to values within the range of the instrumentation.
The inspectors determined that the issue was more than minor because the finding, if left uncorrected, would become a more significant safety concern. Specifically, for a seismic event, the deficiency could lead to the failure to declare an Unusual Event for a Natural and Destructive Phenomena Affecting the Plant Protected Area, HU1.1, and an Alert for a Natural and Destructive Phenomena Affecting the Plant Vital Area, HA1.1. The inspectors determined the finding could be evaluated using the Significance Determincation Process (SDP) and concluded that the risk significant planning standard problem was not a functional failure, nor did it represent a degraded function and, therefore, screened as an issue of very low safety significance (Green). The inspectors determined this was a Green risk significant planning standard problem, rather than degraded or failed risk significant planning standard function, because the process failure affected only one Unusual Event and one Alert emergency classification. The inspectors determined this issue had a cross-cutting aspect in the area of Problem Identification and Resolution, trending and assessment, because the licensee failed to perform aggregate assessments that could have identified and prevented this and related issues earlier (P.1(b)).
Inspection Report# : 2009003 (pdf)
Occupational Radiation Safety Public Radiation Safety Significance:      Jun 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Calibration Of Radiation Monitor R-19 A finding of very low safety significance (Green) and an associated Non-Cited Violation of Technical Specification 4.1 were identified by the inspectors for the licensees calibration practices for process radiation instrument R-19 that did not qualify as a Channel Calibration as required by technical specifications. Specifically, the sources for calibration of R-19 were not of sufficient strength to test the instrument in the range where alarms were required to be operable. Corrective actions were taken for this issue and included revising Emergency Action Level (EAL) requirements to values within the range of the instrumentation.
The inspectors determined that the issue was more than minor because it was associated with the Public Radiation Safety Cornerstone attribute of equipment and instrumentation and adversely affected the cornerstone objective to ensure protection of public health and safety from exposure to radioactive materials released into the public domain.
The inspectors used Inspection Manual Chapter (IMC) 0609, Significance Determination Process, Attachment D, Public Radiation Safety Significance Determination Process, dated February 12, 2008, and determined that the finding was in the licensees radiological effluent monitoring program and was contrary to a technical specification requirement. However, the finding was not related to a spill or release of radioactive material to the environment and, therefore, screened as an issue of very low safety significance (Green). The inspectors reviewed this issue for a cross-cutting aspect and determined that no cross-cutting aspect was applicable.
Inspection Report# : 2009003 (pdf)
 
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.
Miscellaneous Significance: SL-IV Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure To Follow Independent Spent Fuel Storage Installation Loading Procedure Step The inspectors identified a Severity Level IV Non-Cited Violation of 10 CFR 72.150, Instructions, Procedures, and Drawings, during the Independent Spent Fuel Storage Installation loading campaign. The licensee failed to follow procedure OP KW NOP ISF 001, Dry Shielded Canister Loading. The inspectors determined that the licensees failure to follow step 5.2.6 of Procedure OP-KW-NOP-ISF-001 to perform a crane brake check was contrary to 10 CFR 72.150. The licensee immediately evaluated the situation and discussed the need to check the crane brakes when lifting loads approaching the rated loads with the refueling crew to prevent missing this step in the future.
The inspectors determined that the violation had more than minor safety significance because the failure to check the crane brakes, results in not knowing if the brakes are functioning properly, which may lead to a failure of the brakes while lifting a loaded spent fuel canister. The issue was addressed by traditional enforcement since 10 CFR Part 72 is not risk based and is not covered under the reactor oversight process. Because this violation was of very low safety significance, was non-repetitive and non-willful, and was entered into the corrective action program, this violation is being treated as a Non-Cited Violation of 10 CFR 72.150 consistent with Section VI.A.1 of the Enforcement Policy.
The inspectors determined that there was no cross-cutting aspect associated with this finding.
Inspection Report# : 2009004 (pdf)
Last modified : December 10, 2009
 
Kewaunee 4Q/2009 Plant Inspection Findings Initiating Events Significance:      Dec 31, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate Work Instructions Lead to Component Cooling Water Relief Valve Lift And Surge Tank Level Drop A finding of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was self-revealed for the failure to have adequate work instructions in place during the isolation of component cooling water (CCW) flow in the reactor coolant pump vaults.
Specifically, the inadequate valve isolation sequence and the speed at which the outlet valves were closed caused CCW system relief valves to lift and rapidly drain the component cooling water surge tank while the CCW system was supporting the residual heat removal system for decay heat removal. In response to the issue, the licensee implemented compensatory corrective actions to modify the tagout and hang tags on the appropriate CCW isolation valves.
The inspectors determined that the finding was more than minor because it was associated with the Initiating Events Cornerstone attribute of configuration control and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations.
The inspectors determined that the finding could be evaluated in accordance with Inspection Manual Chapter 0609, Appendix G, "Shutdown Operations Significance Determination Process." The inspectors used Checklist 3 contained in Attachment 1 and determined that the finding required a Phase 2 analysis since the finding increased the likelihood that a loss of decay heat removal would occur. The Region III senior reactor analyst performed the assessment using Appendix G, Attachment 2, "Phase 2 Significance Determination Process Template for PWR [Pressurized Water Reactor] During Shutdown," and determined that this issue is best characterized as a finding of very low safety significance (Green). This finding has a cross-cutting aspect in the area of human performance, resources component, because the licensee did not maintain long-term plant safety by maintenance of design margins. Specifically, the work instruction did not adequately account for the low design margin that existed between the system operating pressure and the relief valve setpoints when both CCW pumps were running (H.2(a)).
Inspection Report# : 2009005 (pdf)
Significance:      Dec 31, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation Procedure Inadequacy Results In The Tertiary Auxiliary Transformer Breaker Reopening After Alignment To The Bus A finding of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was self-revealed for the licensee's failure to have adequate procedures to ensure that steps were sequenced such that unplanned transients were not initiated. Specifically, the procedure for performing emergency diesel generator train "A" automatic testing allowed steps to be sequenced in a manner such that a jumper used to simulate a station blackout signal was left installed during the restoration of offsite power. Because of the installed jumpers, a transient was initiated on the associated bus and attached equipment during the restoration from testing. In response to the issue, the licensee implemented compensatory corrective actions and corrected the procedure deficiency prior to conducting the same test on the opposite train.
The inspectors determined that the finding was more than minor because it was associated with the Initiating Events Cornerstone attribute of procedure quality 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. The
 
inspectors evaluated the significance of the issue using Inspection Manual Chapter 0609, Appendix G, Checklist 3, and determined that the power availability guidelines were met. Because the finding did not increase the likelihood of a loss of offsite power or degrade the licensee's ability to cope with a loss of offsite power, the finding screened as having very low safety significance (Green). The finding has a cross-cutting aspect in the area of human performance, work practices component, because the procedure was not adequately verified when steps were changed from being sequence-dependent to allow for completion in any order. Specifically, personnel proceeded to change procedure without implementing peer-checking during the validation process to ensure that the change was applicable to all plant conditions (H.4(a)).
Inspection Report# : 2009005 (pdf)
Significance:        Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure To Adequately Analyze The Automatic Fast Transfer Feature That Allowed Operation With Both 4.16-kiloVolt Safety-Related Buses 1-5 And 1-6 Connected To The Reserve Auxiliary Transformer A finding of very low safety-significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," was identified by the inspectors for the failure to perform a power system analysis calculation that would have identified that the fast transfer design feature/scheme was deficient, in that, it allowed an unanalyzed electrical power system alignment where both redundant 4.16-kiloVolt safety-related buses were being supplied by an offsite source via the same transformer. Use of this electrical configuration could have resulted in an out-of-phase transfer, loss of available offsite power to the buses and potential damaging effects on redundant safety related equipment, during a design basis event such as initiation of safety injection signal. When identified, the licensee entered this issue into their corrective action program and implemented interim actions to prohibit use of the fast transfer feature or manually aligning two safety-related buses to be fed from the same transformer during plant operation.
This performance deficiency was more than minor because the failure to perform the required calculation resulted in a condition where the plant was being operated in an unanalyzed configuration where there was reasonable doubt as to the operability of redundant safeguard loads; this concern resulted in issuance of a Licensee Event Report 2007-007-00 on May 21, 2007. Consequently, the potential for damage or loss of power to safety-related loads during an event could have led to unacceptable consequences. The finding screened as being of very low safety-significance (Green) for the Initiating Events Cornerstone because the finding did not contribute to both the likelihood of a reactor trip and the likelihood that mitigating equipment or functions will not be available. The inspectors did not identify a cross cutting aspect associated with this finding because the cause of the performance deficiency was related to a historical design issue and not indicative of current licensee performance.
Inspection Report# : 2009004 (pdf)
Significance:        Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Design Analysis For 105-Ton Transfer Cask Lifting Beam A finding of very low safety significance and associated Non-Cited Violation of Title 10 Code of Federal Regulations Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors for the licensees failure to provide an adequate single failure proof design basis analysis for the 105-ton transfer cask-lifting beam. The licensee entered this issue into their corrective action program as condition report CR339267. The licensee revised the design calculation for the 105-ton transfer cask-lifting beam and demonstrated compliance with single failure proof acceptance criteria.
The finding was determined to be more than minor because the finding was associated with the Initiating Events Cornerstone attribute of equipment performance 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. The finding was determined to be of very low safety significance by the NRCs significance determination process because the transfer cask-lifting beam had not been previously used at the Kewaunee Power
 
Station. This finding has a cross-cutting aspect in the area of human performance, work practices, because the licensee did not ensure supervisory and management oversight of work activities, including contractors, such that nuclear safety is supported, in that, the licensee failed to perform an effective owners review to assure that appropriate design methods are used in calculations that demonstrate nuclear safety (H.4(c)).
Inspection Report# : 2009004 (pdf)
Significance:      Mar 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inappropriate Application Of A Dedicated Operator During A Component Cooling Water Surveillance A finding of very low safety significance (Green) and associated Non-Cited Violation of 10 CFR 50.65(a)(4) was identified by the inspectors for the failure to properly assess risk that resulted from risk significant maintenance being performed on the component cooling water (CCW) system, when the licensee inappropriately applied criteria for the use of a dedicated operator to meet availability requirements. As part of its corrective actions, the licensee stopped work that required the use of a dedicated operator pending further evaluation.
The issue was more than minor because the licensees risk assessment for March 11, 2009, failed to consider the CCW unavailable during maintenance. Specifically, the failure to account for the unavailability of CCW resulted in an inadequate daily risk assessment and could affect the unavailability time of this system in related performance and maintenance rule indicators. The inspectors evaluated the finding using the Significance Determination Process in accordance with Inspection Manual Chapter 0609, Significance Determination Process, Attachment K, Maintenance Risk Assessment and Risk Management Significance Determination Process, dated May 19, 2005, and determined the issue screened as having very low safety significance (Green), because the incremental conditional core damage probability was less than 1E 6 due to the test condition lasting only four hours. The inspectors determined that the finding had a cross cutting aspect in the corrective action program component of problem identification and resolution, because the licensee failed to thoroughly evaluate a prior problem such that the resolution addressed the causes and extent of conditions necessary to preclude this event (P.1(c)).
Inspection Report# : 2009002 (pdf)
Mitigating Systems Significance:      Dec 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure To Perform Dye Penetrant Examinations Of The Full Code Required Exam Surfaces The inspectors identified a finding of very low safety significance and an associated Non-Cited Violation of 10 CFR 50.55a(g)(4) for the failure to perform dye penetrant examinations of the full required exam surface on safety injection (SI) gas collection chamber welds (SI-W603, SI-W604, and SI-H109) in accordance with the American Society of Mechanical Engineers Section XI Code. Specifically, the examiner proceeded with the examination without anticipating the effects of the increased dwell and drying times of the developer due to cooler ambient temperature than those he had been working under previously. The developer, which would normally dry to a white residue shortly after application to a warm surface and aid in determining the extent of application, remained somewhat translucent when applied to the cooler surface, masking the extent of coverage. This resulted in the examiner's failure to coat the full required Code areas of the welds he was examining and his failure to recognize the lack of coverage. The licensee subsequently re-performed the dye penetrant examinations and entered this issue into their corrective action program.
The inspectors determined that the finding was more than minor because it was associated with the Mitigating System Cornerstone attribute of equipment performance and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Absent NRC intervention, the licensee would not have performed the full Code required examination of welds SI-W603, SI W604, and SI-H109 for an indefinite period of service, which would have placed the reactor coolant pressure
 
boundary at increased risk for unanalyzed cracking, leakage, or component failure. This finding was of very low safety significance because a qualified examination was subsequently performed with no relevant indications detected.
In particular, it did not result in the loss of function of the mitigating system. The inspectors determined that the finding had a cross-cutting aspect in the area of human performance, work practices component, because the licensee proceeded in the face of uncertainty or unexpected circumstances (H.4(a)).
Inspection Report# : 2009005 (pdf)
Significance:        Dec 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Latching Pawl On Safety-Related Bus Tie Breakers Fails To Engage Due To Grease Hardening The inspectors identified a finding of very low safety significance and an associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," for the failure to promptly identify and correct deficiencies that had caused 4160-Volt alternating current breaker failures, which, if corrected, may have prevented subsequent similar failures. Specifically, the licensee did not evaluate other safety-related breakers after hardened grease was identified in the safety-related bus 5 to bus 6 crosstie breakers. In response to this finding, the licensee entered the issue into its corrective action program as Condition Report (CR) 360677.
The inspectors determined that the finding was more than minor because it was associated with the Mitigating Systems Cornerstone attribute of procedure quality 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 inspectors determined the finding could be evaluated using the Significance Determination Process in accordance with Inspection Manual Chapter 0609, "Significance Determination Process," Attachment 0609.04, "Phase 1, Initial Screening and Characterization of Findings," Table 4a, for the Mitigating Systems Cornerstone, dated January 10, 2008. The significance of the finding was determined to be of very low safety significance because the inspectors answered "no" to all of the questions in the Mitigating Systems Cornerstone column. The inspectors determined that the issue had a cross-cutting aspect in human performance, work practices component, because licensee staff did not comply with the timeliness aspects of completing an apparent cause evaluation in accordance with procedure guidance (H.4(b)).
Inspection Report# : 2009005 (pdf)
Significance:        Aug 20, 2009 Identified By: NRC Item Type: NCV NonCited Violation Improper Application of 440Vac Rated Motors The inspectors identified a finding of very low safety-significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to ensure the proper application of safety-related 440Vac motors.
Specifically, eight 440Vac safety-related motors were not suitable for operation at analyzed voltages. This finding was entered into the licensees corrective action program.
The finding was more than minor because if left uncorrected it could result in the loss of safety-related 440Vac motors by overstressing of the motor windings through exposure to higher than design rated voltages, and in the failure of motor drive components caused by increased torque produced at the higher voltages. The finding was determined to be of very low safety-significance (Green) because it did not result in a loss of operability. The cause of this finding is related to the cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, because the licensee did not identify this issue completely, accurately, and in a timely manner. The values were produced in a calculation but the licensee did not identify that they exceeded the acceptance criteria. (P.1(a)) (Section 1R21.3)
Inspection Report# : 2009006 (pdf)
Significance:        Aug 20, 2009 Identified By: NRC
 
Item Type: NCV NonCited Violation Inaccurate Minimum Low Head Safety Injection Flow Specified in Emergency Operating Procedure The inspectors identified a finding of very low safety-significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to specify the appropriate quantitative acceptance criterion to assure that adequate Emergency Core Cooling System flow would be delivered to the core following switchover to containment sump recirculation. This finding was entered into the licensees corrective action program.
The finding was determined to be more than minor because the licensee failed to include the appropriate quantitative set-point value for the minimum low-head safety injection train flow following switchover to containment sump recirculation to assure sufficient reactor coolant was available. This finding is of very low safety-significance (Green) because it did not result in a loss of operability, did not represent an actual loss of safety function, and is not potentially risk-significant due to external events. The cause of this finding is related to the cross-cutting aspect in the area of Human Performance, Work Practices, because the licensee did not ensure proper supervisory and management oversight of contractor work activities. Vendor calculations were used as the basis for an EOP set-point without taking into account specific plant design information such as instrument uncertainties, flow instrument calibration effects, and RHR minimum flow. (H.4(c)) (Section 1R21.3)
Inspection Report# : 2009006 (pdf)
Significance:      Aug 20, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate procedure for a Battery Room Flooding Event.
A finding of very low safety significance and associated Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the failure to provide adequate procedural direction to respond to a rupture of the service water piping in the battery rooms. As part of its corrective actions, the licensee revised OP-KW-AOP-MDS-001, Abnormal Operation of Miscellaneous Drains and Sumps, to correct the inadequate operator actions.
The finding was determined to be more than minor because the licensee failed to provide adequate procedural direction for a battery room A or B flood caused by a rupture of the SW piping to/from the battery room fan coil unit in the affected battery room, which ensured the protection of the battery in the unaffected room not associated with the initial flooding event. This finding is of very low safety significance (Green) because it did not result in a loss of operability, did not represent an actual loss of safety function, and is not potentially risk-significant due to external events. The cause of this finding is related to the cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, because the licensee did not fully evaluate the battery room flooding event (an issue potentially impacting nuclear safety) such that the resolution addressed causes, and extent of condition as necessary, to assure nuclear safety. (P.1(c)). (Section 1R21.6).
Inspection Report# : 2009006 (pdf)
Significance: SL-IV Jun 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Emergency Diesel Generator Air System May Not Be Appropriately Qualified A finding of very low safety significance (Green) and an associated Non-Cited Violation of 10 CFR 50.59 were identified by the inspectors for the licensees failure to obtain a license amendment when it failed to properly assess a quality assurance typing change to the emergency diesel generator starting air compressors. Violations of 10 CFR 50.59 are considered to be violations that potentially impede or impact the regulatory process, and are dispositioned using the traditional enforcement process. The licensee entered this issue into its corrective action program as Condition Report (CR) 326432 for evaluation and development of corrective actions, as appropriate.
Supplement I of the Enforcement Policy was used to determine the severity of the underlying technical issue evaluated under the Significance Determination Process (SDP). The issue was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of design control and adversely 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 inspectors evaluated the finding using the SDP
 
and the inspectors screened the issue as very low safety significance (Green) because the quantity of air available to supply air for five start attempts of the diesels and to supply support systems for the emergency diesel generators always exceeded that needed for 24 hours of operation, thereby, resulting in the probabilistic risk assessment function for the diesels being met. The inspectors determined this issue had a cross-cutting aspect in the area of Problem Identification and Resolution, trending and assessment, because the licensee failed to perform aggregate assessments that could have identified and prevented this and related issues earlier (P.1(b)).
Inspection Report# : 2009003 (pdf)
Significance:        Jun 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure To Ensure That Motor Control Circuit Control Circuits Have Adequate Voltage To Operate During Design Basis Accident Conditions
. A finding of very low safety significance (Green) and an associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, were identified by the inspectors for the failure to verify that motor control center (MCC) control circuits for some ventilation fans and safety injection system isolation valves would have adequate voltage to operate and, therefore, could result in a loss of function of the circuits during a design basis accident. To address this issue, the licensee modified several MCC starter circuits, which entailed replacement of some inadequately-sized control power transformers, starters, and fuses, and implemented procedures changes to reduce MCC loads.
The finding was determined to be more than minor because the calculation errors resulted in four inoperable components and a condition where there was reasonable doubt on the operability of several other safety-related loads.
The inspectors assessed the significance of this finding for each affected component and determined that the finding did not either relate to a containment structure, system, or component or containment status that had an impact on large early release frequency, or did not result in loss of operability or functionality of the safety injection system because the discharge isolation valves were aligned in their required accident positions and de-powered. In addition, the inspectors assessed the impact on the components powered from the MCCs and determined that the overall failure to ensure adequate voltage at the MCCs as having very low safety significance. Therefore, the finding screened as of very low safety significance (Green). The inspectors determined that there was no cross-cutting aspect to this finding because the cause of the problem occurred many years and was not indicative of current performance.
Inspection Report# : 2009003 (pdf)
Significance:        Mar 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Potential Debris Sources Could Clog A Drain Credited During Internal Floods A finding of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors for the failure to translate the flooding design basis into specifications, procedures, and instructions. Specifically, the licensee failed to control the storage of material in the steam generator blowdown tank room that could potentially clog a floor drain, in an adjoining room, that was credited in a flood analysis. As part of its corrective actions, the licensee removed or secured the material of concern.
The finding was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of Protection Against External Events and adversely 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). Specifically, the licensee did not put adequate controls in place to ensure that the drain would performed its credited function to be open and free flowing during an internal flood scenario involving a break in a 4-inch condensate line. The inspectors evaluated the finding using the Significance Determination Process in accordance with Inspection Manual Chapter 0609, Significance Determination Process, Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings, Table 4a, for the Mitigating Systems Cornerstone, dated January 10, 2008. The significance of the finding was determined to be of very low safety significance (Green) because the inspectors answered no to the questions in the Mitigation Systems Cornerstone column. The inspectors did not
 
identify a cross-cutting aspect associated with this finding because the controls over material that could plug the drain should have been implemented when calculation 2005-05708 was completed and incorporated in the flooding design basis in 2005; therefore, this issue was not reflective of current performance.
Inspection Report# : 2009002 (pdf)
Significance:      Mar 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Siphon Line Which Interconnected Two Diesel Generator Emergency Fuel Oil Storage Tanks Was Not Functioning as Designed A finding of very low safety significance (Green) and associated Severity Level IV, Non-Cited Violation of 10 CFR 50.59 was identified by the inspectors while reviewing Unresolved Item 05000305/2008003-03, Siphon Line Which Interconnected Two Diesel Generator Emergency Fuel Oil Storage Tanks Was Not Functioning as Designed.
Specifically, while performing Updated Safety Analysis Report change request, UCR 93 031, the licensee inappropriately screened the removal of the Updated Safety Analysis Report reference to the siphon line when plant staff incorrectly answered no to all of the 10 CFR 50.59 evaluation questions. The licensee entered this issue into its corrective action program for evaluation and development of corrective actions, as appropriate.
Because violations of 10 CFR 50.59 are considered to be violations that potentially impede or impact the regulatory process, they are dispositioned using the traditional enforcement process. As described in Supplement I of the Enforcement Policy, to determine the severity of a 10 CFR 50.59 violation, the underlying technical issue was evaluated under the Significance Determination Process. The inspectors evaluated the finding using the Significance Determination Process in accordance with Inspection Manual Chapter 0609, Significance Determination Process, 609.04, Phase 1- Initial Screening and Characterization of Findings, Table 4a, for the Mitigating Systems Cornerstone, dated January 10, 2008. The inspectors answered yes to Question 2 in the Mitigation System Cornerstone column which required the issue to be evaluated in accordance with Appendix A, of Inspection Manual Chapter 0609. Using Appendix A, the inspectors screened the issue as very low safety significance (Green) because the quantity of fuel to the diesel generators that was historically available always exceeded that needed for 24 hours of operation, thereby, resulting in the probabilistic risk assessment function for the diesels being met. The inspectors determined that the issue had a cross-cutting aspect in problem identification and resolution, corrective action program, because the licensee had identified similar deficiencies with accurately applying or interpreting the current licensing basis, and failed to take timely action to complete corrective actions, or establish barriers to prevent recurrence of this deficiency (P.1(d)).
Inspection Report# : 2009002 (pdf)
Barrier Integrity Significance:      Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Containment Isolation Valve Inoperable With No Technical Specification Action requirement Entry A finding of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the licensees failure to have adequate procedures that ensured technical specifications were entered and followed for containment isolation valves. The licensee entered the issue into their corrective action program as Condition Report 344856 and Condition Report 350526A, and provided additional guidance to operations personnel. At the end of the inspection period, the licensee continued to perform a causal analysis.
The inspectors determined that the issue was more than minor because the finding, if left uncorrected, would become a more significant safety concern. Specifically, not entering the appropriate technical specification action requirements, when necessary, would lead to more significant safety concerns. The inspectors determined the finding
 
could be evaluated using the Significance Determination Process in accordance with Inspection Manual Chapter 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of findings, Table 4a for the Barrier Integrity Cornerstone. The inspectors answered no to the Barrier Integrity questions and screened the finding as having very low safety significance (Green). The finding has a cross-cutting aspect in the area of human performance, resources, because the licensee did not have complete, accurate and up-to-date design documentation, procedures and work packages (H.2(c)).
Inspection Report# : 2009004 (pdf)
Significance:      Aug 20, 2009 Identified By: NRC Item Type: NCV NonCited Violation Non-Conservative Main Steam Line Break Analysis The inspectors identified a finding of very low safety significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to correctly translate the design bases for the maximum steam generator narrow range level into procedures and instructions. This finding was entered into the licensees corrective action program.
The finding was determined to be more than minor because an evaluation was required to ensure that accident analysis requirements for peak containment pressure were met. The finding also impacted the Barrier Integrity cornerstone attribute of procedure quality, and affected the cornerstone objective of maintaining the functionality of containment to protect the public from radionuclide releases caused by accidents or events. Procedural guidance was not adequate to maintain the plant within the parameters specified in the analysis for containment operability. The finding screened as having very low safety significance (Green) because there was no actual barrier degradation. The inspectors determined there was no cross-cutting aspect associated with this finding. (Section 1R21.4)
Inspection Report# : 2009006 (pdf)
Significance: SL-IV Aug 20, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate 50.59 Evaluation of Shutdown Loss of Coolant Accident Procedure.
The inspectors identified a finding of very low safety significance and associated Non-Cited Violation of 10 CFR 50.59(d)(1) for the licensees failure to perform an adequate review of an abnormal operating procedure associated with a shutdown loss of coolant accident. As part of its corrective actions, the licensee revised procedure OP-KW-AOP-RHR-002 to remove the procedure applicability to the Cold Shutdown mode and Refueling mode with reactor vessel head on.
The inspectors determined that the finding was more than minor because it could not reasonably be determined that the activity would not ultimately have required NRC approval. Operation in accordance with the procedure may have challenged the reactor coolant system barrier. The inspectors determined that the finding did not require a quantitative assessment per IMC 0609, Appendix G. Therefore, the finding screened as having very low safety significance (Green) and was determined to be a Severity Level IV violation. The cause of this finding is related to the cross-cutting aspect in the area of Human Performance, Decision Making, because the licensee failed to use conservative assumptions in decision making to demonstrate that the proposed action to include additional modes of applicability for the Shutdown LOCA procedure was safe in order to proceed. (H.1(b)) (Section 1R21.6)
Inspection Report# : 2009006 (pdf)
Significance:      Jun 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failed Backdraft Damper Renders Containment Fan Coil Unit Inoperable
. A finding of very low safety significance (Green) and an associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, were self revealed for the licensees failure to maintain adequate procedures for the inspection and verification-of-operation for the A containment fan coil unit backdraft dampers. The licensee entered this issue into the licensees corrective action program as Condition Report (CR) 328191; immediate corrective actions were accomplished to repair the affected components.
 
The finding was determined to be more than minor because the finding was associated with the Barrier Integrity Cornerstone attribute of design control and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers, specifically containment, protect the public from radionuclide releases caused by accidents or events. The inspectors determined the finding could be evaluated using the Significance Determination Process (SDP) and screened the finding as having very low safety significance (Green). This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, operating experience, because the licensee did not use operating experience to support plant safety (P.2(b)).
Inspection Report# : 2009003 (pdf)
Significance:        Jun 30, 2009 Identified By: NRC Item Type: FIN Finding Failure To Update Procedures As Required By Commitments A finding of very low safety significance (Green) was identified by the inspectors for the licensees failure to update procedures as required by NRC commitments. Specifically, a procedure for fuel oil sampling and a procedure for steam generator tube inspections were not maintained as required by the referenced commitments. The inspectors determined that the issues constituted a finding relating to management of commitments as required by Nuclear Energy Institute 99-04, Guidelines for Managing NRC Commitment Changes. The licensee has entered this issue into its corrective action program as Condition Report (CR) 340864 to assess the failure to the effects of revisions to reference and end-use documents on each other.
The inspectors concluded that the issue was more than minor because the integration of vendor/industry guidance was related to a commitment to the NRC for steam generator tube inspections, and the failure to appropriately manage the commitments impacted the regulatory process. The issue was administrative in nature and did not impact any safety or risk significant systems, therefore, the issue was determined to be of very low safety significance (Green). The inspectors determined that the issue had a cross cutting aspect related in the area of Human Performance, resources, because the licensee failed to maintain the related procedures complete, accurate, and up to date (H.2(c)).
Inspection Report# : 2009003 (pdf)
Significance:        Mar 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Steam Exclusion Door Failure Results In Multiple Safety Systems Being Declared Inoperable A finding of very low safety significance (Green) and associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed for the licensees failure to follow the corrective action program procedure to implement corrective actions that could have prevented a December 30, 2008, door seal failure, which rendered both trains of control room ventilation inoperable. The licensee entered this issue into its corrective action program and, as partially corrective action, has increased its monitoring of doors for potential failure mechanisms.
The finding was determined to be more than minor because it was associated with the Barrier Integrity Cornerstone attribute of Configuration Control 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. The inspectors evaluated the finding using the Significance Determination Process (SDP) in accordance with Inspection Manual Chapter (IMC) 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, dated January 10, 2008, and determined the finding represented a degradation of the barrier function to protect against radiological hazards, toxic gas, and smoke that required a Phase 3 analysis. A Region III Senior Reactor Analyst completed a qualitative Phase 3 analysis and determined that because the duration of the event was small, 44 minutes, the issue screened as having very low safety significance (Green).
The inspectors determined that the finding had a cross-cutting aspect in the corrective action program component element of problem identification and resolution because the licensee failed to take appropriate corrective actions to address safety issues and adverse trends in a timely manner (P.1(d)).
 
Inspection Report# : 2009002 (pdf)
Emergency Preparedness Significance:      Jun 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Seismic Monitoring System Repeatedly Fails Surveillance A finding of very low safety significance (Green) and an associated Non-Cited Violation were identified by the inspectors for the licensees failure to maintain radiation monitoring instrumentation operable that was required by its emergency plans to meet the standards set forth in 10 CFR 50.47(b). Specifically, seismic instrumentation needed for two Emergency Action Levels, HU1.1 and HA1.1, was not maintained operable such that a related Unusual Event notification and an Alert declaration could have been made under certain conditions. Corrective actions were taken for this issue and included revising Emergency Action Level (EAL) requirements to values within the range of the instrumentation.
The inspectors determined that the issue was more than minor because the finding, if left uncorrected, would become a more significant safety concern. Specifically, for a seismic event, the deficiency could lead to the failure to declare an Unusual Event for a Natural and Destructive Phenomena Affecting the Plant Protected Area, HU1.1, and an Alert for a Natural and Destructive Phenomena Affecting the Plant Vital Area, HA1.1. The inspectors determined the finding could be evaluated using the Significance Determincation Process (SDP) and concluded that the risk significant planning standard problem was not a functional failure, nor did it represent a degraded function and, therefore, screened as an issue of very low safety significance (Green). The inspectors determined this was a Green risk significant planning standard problem, rather than degraded or failed risk significant planning standard function, because the process failure affected only one Unusual Event and one Alert emergency classification. The inspectors determined this issue had a cross-cutting aspect in the area of Problem Identification and Resolution, trending and assessment, because the licensee failed to perform aggregate assessments that could have identified and prevented this and related issues earlier (P.1(b)).
Inspection Report# : 2009003 (pdf)
Occupational Radiation Safety Public Radiation Safety Significance:      Jun 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Calibration Of Radiation Monitor R-19 A finding of very low safety significance (Green) and an associated Non-Cited Violation of Technical Specification 4.1 were identified by the inspectors for the licensees calibration practices for process radiation instrument R-19 that did not qualify as a Channel Calibration as required by technical specifications. Specifically, the sources for calibration of R-19 were not of sufficient strength to test the instrument in the range where alarms were required to be operable. Corrective actions were taken for this issue and included revising Emergency Action Level (EAL) requirements to values within the range of the instrumentation.
The inspectors determined that the issue was more than minor because it was associated with the Public Radiation
 
Safety Cornerstone attribute of equipment and instrumentation and adversely affected the cornerstone objective to ensure protection of public health and safety from exposure to radioactive materials released into the public domain.
The inspectors used Inspection Manual Chapter (IMC) 0609, Significance Determination Process, Attachment D, Public Radiation Safety Significance Determination Process, dated February 12, 2008, and determined that the finding was in the licensees radiological effluent monitoring program and was contrary to a technical specification requirement. However, the finding was not related to a spill or release of radioactive material to the environment and, therefore, screened as an issue of very low safety significance (Green). The inspectors reviewed this issue for a cross-cutting aspect and determined that no cross-cutting aspect was applicable.
Inspection Report# : 2009003 (pdf)
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.
Miscellaneous Significance: SL-IV Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure To Follow Independent Spent Fuel Storage Installation Loading Procedure Step The inspectors identified a Severity Level IV Non-Cited Violation of 10 CFR 72.150, Instructions, Procedures, and Drawings, during the Independent Spent Fuel Storage Installation loading campaign. The licensee failed to follow procedure OP KW NOP ISF 001, Dry Shielded Canister Loading. The inspectors determined that the licensees failure to follow step 5.2.6 of Procedure OP-KW-NOP-ISF-001 to perform a crane brake check was contrary to 10 CFR 72.150. The licensee immediately evaluated the situation and discussed the need to check the crane brakes when lifting loads approaching the rated loads with the refueling crew to prevent missing this step in the future.
The inspectors determined that the violation had more than minor safety significance because the failure to check the crane brakes, results in not knowing if the brakes are functioning properly, which may lead to a failure of the brakes while lifting a loaded spent fuel canister. The issue was addressed by traditional enforcement since 10 CFR Part 72 is not risk based and is not covered under the reactor oversight process. Because this violation was of very low safety significance, was non-repetitive and non-willful, and was entered into the corrective action program, this violation is being treated as a Non-Cited Violation of 10 CFR 72.150 consistent with Section VI.A.1 of the Enforcement Policy.
The inspectors determined that there was no cross-cutting aspect associated with this finding.
Inspection Report# : 2009004 (pdf)
Last modified : March 01, 2010
 
Kewaunee 1Q/2010 Plant Inspection Findings Initiating Events Significance:        Mar 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Fuel Loading Occurs With Boron Concentration Below Required Minimum A finding of very low safety significance and associated Non-Cited Violation of Technical Specification 3.8.a.5 was self-revealed when the licensee loaded fuel into the reactor with reactor coolant system boron sample results less than the minimum boron concentration as specified in the core operating limits report. Once the licensee believed the boron concentration samples were accurate and that boron concentration was below the required minimum, operators stopped moving fuel until the boron concentration was restored to acceptable limits. The licensee entered the issue into the corrective action program as Condition Report 351923. The licensee conducted an apparent cause evaluation and proposed long-term corrective actions, including procedure enhancements, operator training on the event, and conservative decision making training.
This finding was determined to be more than minor because it was associated with the Initiating Events Cornerstone attribute of human performance and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown operations. Specifically, the licensee did not believe the initial boron sample results and continued to move fuel with actual boron concentrations below the minimum value specified in the core operating limits report. The inspectors determined that the finding could be evaluated in accordance with Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process. The inspectors used Checklist 4 contained in Attachment 1 and determined that the finding did not require a phase 2 or phase 3 analysis and screened as very low safety significance (Green).
This finding has a cross-cutting aspect in the area of human performance, decision-making, because the licensee failed to use conservative assumptions when making decisions and did not demonstrate that nuclear safety was an overriding priority (H.1(b)).
Inspection Report# : 2010002 (pdf)
Significance:        Mar 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Incorrect Settings On Differential Relay Results In Loss Of Tertiary Auxiliary Transformer A finding of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was self-revealed for the failure to establish adequate measures to identify and control design interfaces and coordinate among participating design organizations. Specifically, the licensee failed to adequately control all required tertiary auxiliary transformer relay inputs/settings that interfaced with the existing plant design. This adversely impacted associated equipment and caused an unanticipated system response. The licensee promptly cleared tags on the reserve auxiliary transformer to restore a normal offsite power source to one of the two 4160-volt safeguards buses. The licensee performed a root cause evaluation and implemented corrective actions, some of which included: modifying the design change process to ensure that all programmable digital device setpoints and inputs were
 
identified; documenting the basis for each setpoint or input in the design change documentation; and providing programmable digital device training for design engineering and maintenance personnel. The licensee entered the issue into its corrective action program as CR 352878.
The finding was determined to be more than minor because the finding was associated with the Initiating Events Cornerstone attribute of design control and adversely affected the cornerstone objective of limiting 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 adequately control all required tertiary auxiliary transformer relay inputs/settings adversely impacted the associated equipment, which caused an unanticipated system response and challenged core shutdown cooling.
The inspectors determined that the finding could be evaluated in accordance with Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process. The inspectors used Checklist 4, contained in Attachment 1, and determined that the finding required a Phase 2 analysis because it degraded the ability to recover the decay heat removal system. The Region III senior reactor analyst performed a phase 2 and subsequently a phase 3 analysis and determined the finding was of very low safety significance (Green).
This finding has a cross-cutting aspect in the area of human performance, resources, because the licensee did not maintain complete, accurate, and up-to-date design documentation (H.2(c)).
Inspection Report# : 2010002 (pdf)
Significance:      Dec 31, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate Work Instructions Lead to Component Cooling Water Relief Valve Lift And Surge Tank Level Drop A finding of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was self-revealed for the failure to have adequate work instructions in place during the isolation of component cooling water (CCW) flow in the reactor coolant pump vaults.
Specifically, the inadequate valve isolation sequence and the speed at which the outlet valves were closed caused CCW system relief valves to lift and rapidly drain the component cooling water surge tank while the CCW system was supporting the residual heat removal system for decay heat removal. In response to the issue, the licensee implemented compensatory corrective actions to modify the tagout and hang tags on the appropriate CCW isolation valves.
The inspectors determined that the finding was more than minor because it was associated with the Initiating Events Cornerstone attribute of configuration control and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations.
The inspectors determined that the finding could be evaluated in accordance with Inspection Manual Chapter 0609, Appendix G, "Shutdown Operations Significance Determination Process." The inspectors used Checklist 3 contained in Attachment 1 and determined that the finding required a Phase 2 analysis since the finding increased the likelihood that a loss of decay heat removal would occur. The Region III senior reactor analyst performed the assessment using Appendix G, Attachment 2, "Phase 2 Significance Determination Process Template for PWR [Pressurized Water Reactor] During Shutdown," and determined that this issue is best characterized as a finding of very low safety significance (Green). This finding has a cross-cutting aspect in the area of human performance, resources component, because the licensee did not maintain long-term plant safety by maintenance of design margins. Specifically, the work instruction did not adequately account for the low design margin that existed between the system operating pressure and the relief valve setpoints when both CCW pumps were running (H.2(a)).
Inspection Report# : 2009005 (pdf)
Significance:      Dec 31, 2009 Identified By: Self-Revealing
 
Item Type: NCV NonCited Violation Procedure Inadequacy Results In The Tertiary Auxiliary Transformer Breaker Reopening After Alignment To The Bus A finding of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was self-revealed for the licensee's failure to have adequate procedures to ensure that steps were sequenced such that unplanned transients were not initiated. Specifically, the procedure for performing emergency diesel generator train "A" automatic testing allowed steps to be sequenced in a manner such that a jumper used to simulate a station blackout signal was left installed during the restoration of offsite power. Because of the installed jumpers, a transient was initiated on the associated bus and attached equipment during the restoration from testing. In response to the issue, the licensee implemented compensatory corrective actions and corrected the procedure deficiency prior to conducting the same test on the opposite train.
The inspectors determined that the finding was more than minor because it was associated with the Initiating Events Cornerstone attribute of procedure quality 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. The inspectors evaluated the significance of the issue using Inspection Manual Chapter 0609, Appendix G, Checklist 3, and determined that the power availability guidelines were met. Because the finding did not increase the likelihood of a loss of offsite power or degrade the licensee's ability to cope with a loss of offsite power, the finding screened as having very low safety significance (Green). The finding has a cross-cutting aspect in the area of human performance, work practices component, because the procedure was not adequately verified when steps were changed from being sequence-dependent to allow for completion in any order. Specifically, personnel proceeded to change procedure without implementing peer-checking during the validation process to ensure that the change was applicable to all plant conditions (H.4(a)).
Inspection Report# : 2009005 (pdf)
Significance:        Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure To Adequately Analyze The Automatic Fast Transfer Feature That Allowed Operation With Both 4.16-kiloVolt Safety-Related Buses 1-5 And 1-6 Connected To The Reserve Auxiliary Transformer A finding of very low safety-significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," was identified by the inspectors for the failure to perform a power system analysis calculation that would have identified that the fast transfer design feature/scheme was deficient, in that, it allowed an unanalyzed electrical power system alignment where both redundant 4.16-kiloVolt safety-related buses were being supplied by an offsite source via the same transformer. Use of this electrical configuration could have resulted in an out-of-phase transfer, loss of available offsite power to the buses and potential damaging effects on redundant safety related equipment, during a design basis event such as initiation of safety injection signal. When identified, the licensee entered this issue into their corrective action program and implemented interim actions to prohibit use of the fast transfer feature or manually aligning two safety-related buses to be fed from the same transformer during plant operation.
This performance deficiency was more than minor because the failure to perform the required calculation resulted in a condition where the plant was being operated in an unanalyzed configuration where there was reasonable doubt as to the operability of redundant safeguard loads; this concern resulted in issuance of a Licensee Event Report 2007-007-00 on May 21, 2007. Consequently, the potential for damage or loss of power to safety-related loads during an event could have led to unacceptable consequences. The finding screened as being of very low safety-significance (Green) for the Initiating Events Cornerstone because the finding did not contribute to both the likelihood of a reactor trip and the likelihood that mitigating equipment or functions will not be available. The inspectors did not identify a cross cutting aspect associated with this finding because the cause of the performance deficiency was related to a historical design issue and not indicative of current licensee performance.
Inspection Report# : 2009004 (pdf)
Significance:        Sep 30, 2009
 
Identified By: NRC Item Type: NCV NonCited Violation Inadequate Design Analysis For 105-Ton Transfer Cask Lifting Beam A finding of very low safety significance and associated Non-Cited Violation of Title 10 Code of Federal Regulations Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors for the licensees failure to provide an adequate single failure proof design basis analysis for the 105-ton transfer cask-lifting beam. The licensee entered this issue into their corrective action program as condition report CR339267. The licensee revised the design calculation for the 105-ton transfer cask-lifting beam and demonstrated compliance with single failure proof acceptance criteria.
The finding was determined to be more than minor because the finding was associated with the Initiating Events Cornerstone attribute of equipment performance 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. The finding was determined to be of very low safety significance by the NRCs significance determination process because the transfer cask-lifting beam had not been previously used at the Kewaunee Power Station. This finding has a cross-cutting aspect in the area of human performance, work practices, because the licensee did not ensure supervisory and management oversight of work activities, including contractors, such that nuclear safety is supported, in that, the licensee failed to perform an effective owners review to assure that appropriate design methods are used in calculations that demonstrate nuclear safety (H.4(c)).
Inspection Report# : 2009004 (pdf)
Mitigating Systems Significance:      Mar 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Incorrect Curve Was Incorporated Into Calibration Surveillance Procedures A finding of very low safety-significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for inadequate surveillance calibration procedures.
Specifically, calibration surveillance procedure SP-06-034B-1, Steam Generator Flow Mismatch and Steam Pressure Instrument Channel 1, failed to have the correct negative ramp curve. The curve was required to ensure that the low steam line pressure safety injection lag circuitry unit did not exceed the Technical Specification setpoint value. This condition also existed in calibration procedures for channels 2, 3, and 4.
The licensee subsequently entered the issue into its corrective action program as CR 367826 and CR 367932. The licensee conducted an apparent cause evaluation and corrective actions were in progress at the conclusion of the inspection period.
The finding was determined to be more than minor because it was associated with the Mitigating System Cornerstone attribute of procedure quality and adversely affected the cornerstone objective of ensuring the reliability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to ensure that the low steam line pressure safety injection lag circuitry units did not exceed the Technical Specification value of less than or equal to 2 seconds. The finding was of very low safety-significance (Green) based on a phase 1 screening in accordance with Inspection Manual Chapter 0609, Appendix A, ASignificance Determination of Reactor Inspection Findings for At-Power Situations." The finding has a cross-cutting aspect in the areas of human performance, work practices, because the licensee failed to ensure that the calculation upon which the surveillance procedure was based, was approved prior to issuance of the procedure (H.4(b)).
Inspection Report# : 2010002 (pdf)
 
Significance:        Feb 12, 2010 Identified By: NRC Item Type: NCV NonCited Violation Calculation Methodology Did Not Represent Actual Plant Equipment Configuration A finding of very low safety-significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors for the licensees failure to assure that the calculation methodology represented the actual plant equipment configuration and that adequate design reviews were performed for verifying or checking the adequacy of design. Specifically, the licensee failed to assure that the methodology used in calculation C11716, MCC [Motor Control Center] Control Circuit Voltage Drop, Revision 1, correctly represented the sequence of operation for the various devices contained within the plant equipments control circuitry, such that the minimum required MCC voltage was available for proper circuit operation. Upon discovery of this condition, the licensee performed a preliminary evaluation and entered the finding into their corrective action program (CR366627 and CR366865).
This finding was more than minor in accordance with IMC 0612, Appendix B because the finding was associated with the design control attribute of the mitigating systems cornerstone and affected the cornerstones objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the inadequate MCC voltages could render the safety-related loads required to mitigate the consequences of a design basis accident inoperable and not available. In addition, as a result of the calculation errors, the inspectors were concerned that unsubstantiated MCC voltage values could be used in future calculations and modifications to plant equipment. To resolve the inspectors concerns, the licensee completed an interim evaluation, which evaluated the calculations other circuit models and associated cases. Although, by the end of the inspection, the licensee was able to demonstrate operability; at the time of discovery there was reasonable doubt on the operability of the control circuits modeled in the calculation. The finding was of very low safety-significance based on a Phase 1 screening in accordance with IMC 0609, Significance Determination Process, 609.04, Phase 1 - Initial Screening and Characterization of Findings, Table 4a.
This finding has a cross-cutting aspect in the area of human performance, work practices because the licensee did not ensure supervisory and management oversight of work activities, including contractors, such that nuclear safety was supported. Specifically, the licensee failed to assure that the calculation methodology represented the actual plant equipment configuration and that adequate design reviews were performed for verifying or checking the adequacy of design. (H.4(c)) (Section 1R17.2b)
Inspection Report# : 2010007 (pdf)
Significance:        Dec 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure To Perform Dye Penetrant Examinations Of The Full Code Required Exam Surfaces The inspectors identified a finding of very low safety significance and an associated Non-Cited Violation of 10 CFR 50.55a(g)(4) for the failure to perform dye penetrant examinations of the full required exam surface on safety injection (SI) gas collection chamber welds (SI-W603, SI-W604, and SI-H109) in accordance with the American Society of Mechanical Engineers Section XI Code. Specifically, the examiner proceeded with the examination without anticipating the effects of the increased dwell and drying times of the developer due to cooler ambient temperature than those he had been working under previously. The developer, which would normally dry to a white residue shortly after application to a warm surface and aid in determining the extent of application, remained somewhat translucent when applied to the cooler surface, masking the extent of coverage. This resulted in the examiner's failure to coat the full required Code areas of the welds he was examining and his failure to recognize the lack of coverage. The licensee subsequently re-performed the dye penetrant examinations and entered this issue into their corrective action program.
The inspectors determined that the finding was more than minor because it was associated with the Mitigating System Cornerstone attribute of equipment performance and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Absent NRC intervention, the licensee would not have performed the full Code required examination of welds SI-W603, SI W604, and SI-H109 for an indefinite period of service, which would have placed the reactor coolant pressure boundary at increased risk for unanalyzed cracking, leakage, or component failure. This finding was of very low
 
safety significance because a qualified examination was subsequently performed with no relevant indications detected.
In particular, it did not result in the loss of function of the mitigating system. The inspectors determined that the finding had a cross-cutting aspect in the area of human performance, work practices component, because the licensee proceeded in the face of uncertainty or unexpected circumstances (H.4(a)).
Inspection Report# : 2009005 (pdf)
Significance:        Dec 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Latching Pawl On Safety-Related Bus Tie Breakers Fails To Engage Due To Grease Hardening The inspectors identified a finding of very low safety significance and an associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," for the failure to promptly identify and correct deficiencies that had caused 4160-Volt alternating current breaker failures, which, if corrected, may have prevented subsequent similar failures. Specifically, the licensee did not evaluate other safety-related breakers after hardened grease was identified in the safety-related bus 5 to bus 6 crosstie breakers. In response to this finding, the licensee entered the issue into its corrective action program as Condition Report (CR) 360677.
The inspectors determined that the finding was more than minor because it was associated with the Mitigating Systems Cornerstone attribute of procedure quality 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 inspectors determined the finding could be evaluated using the Significance Determination Process in accordance with Inspection Manual Chapter 0609, "Significance Determination Process," Attachment 0609.04, "Phase 1, Initial Screening and Characterization of Findings," Table 4a, for the Mitigating Systems Cornerstone, dated January 10, 2008. The significance of the finding was determined to be of very low safety significance because the inspectors answered "no" to all of the questions in the Mitigating Systems Cornerstone column. The inspectors determined that the issue had a cross-cutting aspect in human performance, work practices component, because licensee staff did not comply with the timeliness aspects of completing an apparent cause evaluation in accordance with procedure guidance (H.4(b)).
Inspection Report# : 2009005 (pdf)
Significance:        Aug 20, 2009 Identified By: NRC Item Type: NCV NonCited Violation Improper Application of 440Vac Rated Motors The inspectors identified a finding of very low safety-significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to ensure the proper application of safety-related 440Vac motors.
Specifically, eight 440Vac safety-related motors were not suitable for operation at analyzed voltages. This finding was entered into the licensees corrective action program.
The finding was more than minor because if left uncorrected it could result in the loss of safety-related 440Vac motors by overstressing of the motor windings through exposure to higher than design rated voltages, and in the failure of motor drive components caused by increased torque produced at the higher voltages. The finding was determined to be of very low safety-significance (Green) because it did not result in a loss of operability. The cause of this finding is related to the cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, because the licensee did not identify this issue completely, accurately, and in a timely manner. The values were produced in a calculation but the licensee did not identify that they exceeded the acceptance criteria. (P.1(a)) (Section 1R21.3)
Inspection Report# : 2009006 (pdf)
Significance:        Aug 20, 2009 Identified By: NRC Item Type: NCV NonCited Violation
 
Inaccurate Minimum Low Head Safety Injection Flow Specified in Emergency Operating Procedure The inspectors identified a finding of very low safety-significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to specify the appropriate quantitative acceptance criterion to assure that adequate Emergency Core Cooling System flow would be delivered to the core following switchover to containment sump recirculation. This finding was entered into the licensees corrective action program.
The finding was determined to be more than minor because the licensee failed to include the appropriate quantitative set-point value for the minimum low-head safety injection train flow following switchover to containment sump recirculation to assure sufficient reactor coolant was available. This finding is of very low safety-significance (Green) because it did not result in a loss of operability, did not represent an actual loss of safety function, and is not potentially risk-significant due to external events. The cause of this finding is related to the cross-cutting aspect in the area of Human Performance, Work Practices, because the licensee did not ensure proper supervisory and management oversight of contractor work activities. Vendor calculations were used as the basis for an EOP set-point without taking into account specific plant design information such as instrument uncertainties, flow instrument calibration effects, and RHR minimum flow. (H.4(c)) (Section 1R21.3)
Inspection Report# : 2009006 (pdf)
Significance:      Aug 20, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate procedure for a Battery Room Flooding Event.
A finding of very low safety significance and associated Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the failure to provide adequate procedural direction to respond to a rupture of the service water piping in the battery rooms. As part of its corrective actions, the licensee revised OP-KW-AOP-MDS-001, Abnormal Operation of Miscellaneous Drains and Sumps, to correct the inadequate operator actions.
The finding was determined to be more than minor because the licensee failed to provide adequate procedural direction for a battery room A or B flood caused by a rupture of the SW piping to/from the battery room fan coil unit in the affected battery room, which ensured the protection of the battery in the unaffected room not associated with the initial flooding event. This finding is of very low safety significance (Green) because it did not result in a loss of operability, did not represent an actual loss of safety function, and is not potentially risk-significant due to external events. The cause of this finding is related to the cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, because the licensee did not fully evaluate the battery room flooding event (an issue potentially impacting nuclear safety) such that the resolution addressed causes, and extent of condition as necessary, to assure nuclear safety. (P.1(c)). (Section 1R21.6).
Inspection Report# : 2009006 (pdf)
Significance: SL-IV Jun 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Emergency Diesel Generator Air System May Not Be Appropriately Qualified A finding of very low safety significance (Green) and an associated Non-Cited Violation of 10 CFR 50.59 were identified by the inspectors for the licensees failure to obtain a license amendment when it failed to properly assess a quality assurance typing change to the emergency diesel generator starting air compressors. Violations of 10 CFR 50.59 are considered to be violations that potentially impede or impact the regulatory process, and are dispositioned using the traditional enforcement process. The licensee entered this issue into its corrective action program as Condition Report (CR) 326432 for evaluation and development of corrective actions, as appropriate.
Supplement I of the Enforcement Policy was used to determine the severity of the underlying technical issue evaluated under the Significance Determination Process (SDP). The issue was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of design control and adversely 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 inspectors evaluated the finding using the SDP and the inspectors screened the issue as very low safety significance (Green) because the quantity of air available to
 
supply air for five start attempts of the diesels and to supply support systems for the emergency diesel generators always exceeded that needed for 24 hours of operation, thereby, resulting in the probabilistic risk assessment function for the diesels being met. The inspectors determined this issue had a cross-cutting aspect in the area of Problem Identification and Resolution, trending and assessment, because the licensee failed to perform aggregate assessments that could have identified and prevented this and related issues earlier (P.1(b)).
Inspection Report# : 2009003 (pdf)
Significance:      Jun 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure To Ensure That Motor Control Circuit Control Circuits Have Adequate Voltage To Operate During Design Basis Accident Conditions
. A finding of very low safety significance (Green) and an associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, were identified by the inspectors for the failure to verify that motor control center (MCC) control circuits for some ventilation fans and safety injection system isolation valves would have adequate voltage to operate and, therefore, could result in a loss of function of the circuits during a design basis accident. To address this issue, the licensee modified several MCC starter circuits, which entailed replacement of some inadequately-sized control power transformers, starters, and fuses, and implemented procedures changes to reduce MCC loads.
The finding was determined to be more than minor because the calculation errors resulted in four inoperable components and a condition where there was reasonable doubt on the operability of several other safety-related loads.
The inspectors assessed the significance of this finding for each affected component and determined that the finding did not either relate to a containment structure, system, or component or containment status that had an impact on large early release frequency, or did not result in loss of operability or functionality of the safety injection system because the discharge isolation valves were aligned in their required accident positions and de-powered. In addition, the inspectors assessed the impact on the components powered from the MCCs and determined that the overall failure to ensure adequate voltage at the MCCs as having very low safety significance. Therefore, the finding screened as of very low safety significance (Green). The inspectors determined that there was no cross-cutting aspect to this finding because the cause of the problem occurred many years and was not indicative of current performance.
Inspection Report# : 2009003 (pdf)
Barrier Integrity Significance:      Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Containment Isolation Valve Inoperable With No Technical Specification Action requirement Entry A finding of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the licensees failure to have adequate procedures that ensured technical specifications were entered and followed for containment isolation valves. The licensee entered the issue into their corrective action program as Condition Report 344856 and Condition Report 350526A, and provided additional guidance to operations personnel. At the end of the inspection period, the licensee continued to perform a causal analysis.
The inspectors determined that the issue was more than minor because the finding, if left uncorrected, would become a more significant safety concern. Specifically, not entering the appropriate technical specification action requirements, when necessary, would lead to more significant safety concerns. The inspectors determined the finding could be evaluated using the Significance Determination Process in accordance with Inspection Manual Chapter 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of findings, Table 4a for the Barrier Integrity Cornerstone. The inspectors answered no to the Barrier Integrity questions
 
and screened the finding as having very low safety significance (Green). The finding has a cross-cutting aspect in the area of human performance, resources, because the licensee did not have complete, accurate and up-to-date design documentation, procedures and work packages (H.2(c)).
Inspection Report# : 2009004 (pdf)
Significance:      Aug 20, 2009 Identified By: NRC Item Type: NCV NonCited Violation Non-Conservative Main Steam Line Break Analysis The inspectors identified a finding of very low safety significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to correctly translate the design bases for the maximum steam generator narrow range level into procedures and instructions. This finding was entered into the licensees corrective action program.
The finding was determined to be more than minor because an evaluation was required to ensure that accident analysis requirements for peak containment pressure were met. The finding also impacted the Barrier Integrity cornerstone attribute of procedure quality, and affected the cornerstone objective of maintaining the functionality of containment to protect the public from radionuclide releases caused by accidents or events. Procedural guidance was not adequate to maintain the plant within the parameters specified in the analysis for containment operability. The finding screened as having very low safety significance (Green) because there was no actual barrier degradation. The inspectors determined there was no cross-cutting aspect associated with this finding. (Section 1R21.4)
Inspection Report# : 2009006 (pdf)
Significance: SL-IV Aug 20, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate 50.59 Evaluation of Shutdown Loss of Coolant Accident Procedure.
The inspectors identified a finding of very low safety significance and associated Non-Cited Violation of 10 CFR 50.59(d)(1) for the licensees failure to perform an adequate review of an abnormal operating procedure associated with a shutdown loss of coolant accident. As part of its corrective actions, the licensee revised procedure OP-KW-AOP-RHR-002 to remove the procedure applicability to the Cold Shutdown mode and Refueling mode with reactor vessel head on.
The inspectors determined that the finding was more than minor because it could not reasonably be determined that the activity would not ultimately have required NRC approval. Operation in accordance with the procedure may have challenged the reactor coolant system barrier. The inspectors determined that the finding did not require a quantitative assessment per IMC 0609, Appendix G. Therefore, the finding screened as having very low safety significance (Green) and was determined to be a Severity Level IV violation. The cause of this finding is related to the cross-cutting aspect in the area of Human Performance, Decision Making, because the licensee failed to use conservative assumptions in decision making to demonstrate that the proposed action to include additional modes of applicability for the Shutdown LOCA procedure was safe in order to proceed. (H.1(b)) (Section 1R21.6)
Inspection Report# : 2009006 (pdf)
Significance:      Jun 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failed Backdraft Damper Renders Containment Fan Coil Unit Inoperable
. A finding of very low safety significance (Green) and an associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, were self revealed for the licensees failure to maintain adequate procedures for the inspection and verification-of-operation for the A containment fan coil unit backdraft dampers. The licensee entered this issue into the licensees corrective action program as Condition Report (CR) 328191; immediate corrective actions were accomplished to repair the affected components.
The finding was determined to be more than minor because the finding was associated with the Barrier Integrity Cornerstone attribute of design control and adversely affected the cornerstone objective to provide reasonable
 
assurance that physical design barriers, specifically containment, protect the public from radionuclide releases caused by accidents or events. The inspectors determined the finding could be evaluated using the Significance Determination Process (SDP) and screened the finding as having very low safety significance (Green). This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, operating experience, because the licensee did not use operating experience to support plant safety (P.2(b)).
Inspection Report# : 2009003 (pdf)
Significance:        Jun 30, 2009 Identified By: NRC Item Type: FIN Finding Failure To Update Procedures As Required By Commitments A finding of very low safety significance (Green) was identified by the inspectors for the licensees failure to update procedures as required by NRC commitments. Specifically, a procedure for fuel oil sampling and a procedure for steam generator tube inspections were not maintained as required by the referenced commitments. The inspectors determined that the issues constituted a finding relating to management of commitments as required by Nuclear Energy Institute 99-04, Guidelines for Managing NRC Commitment Changes. The licensee has entered this issue into its corrective action program as Condition Report (CR) 340864 to assess the failure to the effects of revisions to reference and end-use documents on each other.
The inspectors concluded that the issue was more than minor because the integration of vendor/industry guidance was related to a commitment to the NRC for steam generator tube inspections, and the failure to appropriately manage the commitments impacted the regulatory process. The issue was administrative in nature and did not impact any safety or risk significant systems, therefore, the issue was determined to be of very low safety significance (Green). The inspectors determined that the issue had a cross cutting aspect related in the area of Human Performance, resources, because the licensee failed to maintain the related procedures complete, accurate, and up to date (H.2(c)).
Inspection Report# : 2009003 (pdf)
Emergency Preparedness Significance:        Jun 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Seismic Monitoring System Repeatedly Fails Surveillance A finding of very low safety significance (Green) and an associated Non-Cited Violation were identified by the inspectors for the licensees failure to maintain radiation monitoring instrumentation operable that was required by its emergency plans to meet the standards set forth in 10 CFR 50.47(b). Specifically, seismic instrumentation needed for two Emergency Action Levels, HU1.1 and HA1.1, was not maintained operable such that a related Unusual Event notification and an Alert declaration could have been made under certain conditions. Corrective actions were taken for this issue and included revising Emergency Action Level (EAL) requirements to values within the range of the instrumentation.
The inspectors determined that the issue was more than minor because the finding, if left uncorrected, would become a more significant safety concern. Specifically, for a seismic event, the deficiency could lead to the failure to declare an Unusual Event for a Natural and Destructive Phenomena Affecting the Plant Protected Area, HU1.1, and an Alert for a Natural and Destructive Phenomena Affecting the Plant Vital Area, HA1.1. The inspectors determined the finding could be evaluated using the Significance Determincation Process (SDP) and concluded that the risk significant planning standard problem was not a functional failure, nor did it represent a degraded function and, therefore, screened as an issue of very low safety significance (Green). The inspectors determined this was a Green risk significant planning standard problem, rather than degraded or failed risk significant planning standard function, because the process failure affected only one Unusual Event and one Alert emergency classification. The inspectors determined this issue had a cross-cutting aspect in the area of Problem Identification and Resolution, trending and
 
assessment, because the licensee failed to perform aggregate assessments that could have identified and prevented this and related issues earlier (P.1(b)).
Inspection Report# : 2009003 (pdf)
Occupational Radiation Safety Public Radiation Safety Significance:      Jun 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Calibration Of Radiation Monitor R-19 A finding of very low safety significance (Green) and an associated Non-Cited Violation of Technical Specification 4.1 were identified by the inspectors for the licensees calibration practices for process radiation instrument R-19 that did not qualify as a Channel Calibration as required by technical specifications. Specifically, the sources for calibration of R-19 were not of sufficient strength to test the instrument in the range where alarms were required to be operable. Corrective actions were taken for this issue and included revising Emergency Action Level (EAL) requirements to values within the range of the instrumentation.
The inspectors determined that the issue was more than minor because it was associated with the Public Radiation Safety Cornerstone attribute of equipment and instrumentation and adversely affected the cornerstone objective to ensure protection of public health and safety from exposure to radioactive materials released into the public domain.
The inspectors used Inspection Manual Chapter (IMC) 0609, Significance Determination Process, Attachment D, Public Radiation Safety Significance Determination Process, dated February 12, 2008, and determined that the finding was in the licensees radiological effluent monitoring program and was contrary to a technical specification requirement. However, the finding was not related to a spill or release of radioactive material to the environment and, therefore, screened as an issue of very low safety significance (Green). The inspectors reviewed this issue for a cross-cutting aspect and determined that no cross-cutting aspect was applicable.
Inspection Report# : 2009003 (pdf)
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.
Miscellaneous Significance: SL-IV Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure To Follow Independent Spent Fuel Storage Installation Loading Procedure Step The inspectors identified a Severity Level IV Non-Cited Violation of 10 CFR 72.150, Instructions, Procedures, and Drawings, during the Independent Spent Fuel Storage Installation loading campaign. The licensee failed to follow
 
procedure OP KW NOP ISF 001, Dry Shielded Canister Loading. The inspectors determined that the licensees failure to follow step 5.2.6 of Procedure OP-KW-NOP-ISF-001 to perform a crane brake check was contrary to 10 CFR 72.150. The licensee immediately evaluated the situation and discussed the need to check the crane brakes when lifting loads approaching the rated loads with the refueling crew to prevent missing this step in the future.
The inspectors determined that the violation had more than minor safety significance because the failure to check the crane brakes, results in not knowing if the brakes are functioning properly, which may lead to a failure of the brakes while lifting a loaded spent fuel canister. The issue was addressed by traditional enforcement since 10 CFR Part 72 is not risk based and is not covered under the reactor oversight process. Because this violation was of very low safety significance, was non-repetitive and non-willful, and was entered into the corrective action program, this violation is being treated as a Non-Cited Violation of 10 CFR 72.150 consistent with Section VI.A.1 of the Enforcement Policy.
The inspectors determined that there was no cross-cutting aspect associated with this finding.
Inspection Report# : 2009004 (pdf)
Last modified : May 26, 2010
 
Kewaunee 2Q/2010 Plant Inspection Findings Initiating Events Significance:        Mar 31, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Fuel Loading Occurs With Boron Concentration Below Required Minimum A finding of very low safety significance and associated Non-Cited Violation of Technical Specification 3.8.a.5 was self-revealed when the licensee loaded fuel into the reactor with reactor coolant system boron sample results less than the minimum boron concentration as specified in the core operating limits report. Once the licensee believed the boron concentration samples were accurate and that boron concentration was below the required minimum, operators stopped moving fuel until the boron concentration was restored to acceptable limits. The licensee entered the issue into the corrective action program as Condition Report 351923. The licensee conducted an apparent cause evaluation and proposed long-term corrective actions, including procedure enhancements, operator training on the event, and conservative decision making training.
This finding was determined to be more than minor because it was associated with the Initiating Events Cornerstone attribute of human performance and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown operations. Specifically, the licensee did not believe the initial boron sample results and continued to move fuel with actual boron concentrations below the minimum value specified in the core operating limits report. The inspectors determined that the finding could be evaluated in accordance with Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process. The inspectors used Checklist 4 contained in Attachment 1 and determined that the finding did not require a phase 2 or phase 3 analysis and screened as very low safety significance (Green).
This finding has a cross-cutting aspect in the area of human performance, decision-making, because the licensee failed to use conservative assumptions when making decisions and did not demonstrate that nuclear safety was an overriding priority (H.1(b)).
Inspection Report# : 2010002 (pdf)
Significance:        Mar 31, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Incorrect Settings On Differential Relay Results In Loss Of Tertiary Auxiliary Transformer A finding of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was self-revealed for the failure to establish adequate measures to identify and control design interfaces and coordinate among participating design organizations. Specifically, the licensee failed to adequately control all required tertiary auxiliary transformer relay inputs/settings that interfaced with the existing plant design. This adversely impacted associated equipment and caused an unanticipated system response. The licensee promptly cleared tags on the reserve auxiliary transformer to restore a normal offsite power source to one of the two 4160-volt safeguards buses. The licensee performed a root cause evaluation and implemented corrective actions, some of which included: modifying the design change process to ensure that all programmable digital device setpoints and inputs were
 
identified; documenting the basis for each setpoint or input in the design change documentation; and providing programmable digital device training for design engineering and maintenance personnel. The licensee entered the issue into its corrective action program as CR 352878.
The finding was determined to be more than minor because the finding was associated with the Initiating Events Cornerstone attribute of design control and adversely affected the cornerstone objective of limiting 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 adequately control all required tertiary auxiliary transformer relay inputs/settings adversely impacted the associated equipment, which caused an unanticipated system response and challenged core shutdown cooling.
The inspectors determined that the finding could be evaluated in accordance with Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process. The inspectors used Checklist 4, contained in Attachment 1, and determined that the finding required a Phase 2 analysis because it degraded the ability to recover the decay heat removal system. The Region III senior reactor analyst performed a phase 2 and subsequently a phase 3 analysis and determined the finding was of very low safety significance (Green).
This finding has a cross-cutting aspect in the area of human performance, resources, because the licensee did not maintain complete, accurate, and up-to-date design documentation (H.2(c)).
Inspection Report# : 2010002 (pdf)
Significance:      Dec 31, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate Work Instructions Lead to Component Cooling Water Relief Valve Lift And Surge Tank Level Drop A finding of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was self-revealed for the failure to have adequate work instructions in place during the isolation of component cooling water (CCW) flow in the reactor coolant pump vaults.
Specifically, the inadequate valve isolation sequence and the speed at which the outlet valves were closed caused CCW system relief valves to lift and rapidly drain the component cooling water surge tank while the CCW system was supporting the residual heat removal system for decay heat removal. In response to the issue, the licensee implemented compensatory corrective actions to modify the tagout and hang tags on the appropriate CCW isolation valves.
The inspectors determined that the finding was more than minor because it was associated with the Initiating Events Cornerstone attribute of configuration control and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations.
The inspectors determined that the finding could be evaluated in accordance with Inspection Manual Chapter 0609, Appendix G, "Shutdown Operations Significance Determination Process." The inspectors used Checklist 3 contained in Attachment 1 and determined that the finding required a Phase 2 analysis since the finding increased the likelihood that a loss of decay heat removal would occur. The Region III senior reactor analyst performed the assessment using Appendix G, Attachment 2, "Phase 2 Significance Determination Process Template for PWR [Pressurized Water Reactor] During Shutdown," and determined that this issue is best characterized as a finding of very low safety significance (Green). This finding has a cross-cutting aspect in the area of human performance, resources component, because the licensee did not maintain long-term plant safety by maintenance of design margins. Specifically, the work instruction did not adequately account for the low design margin that existed between the system operating pressure and the relief valve setpoints when both CCW pumps were running (H.2(a)).
Inspection Report# : 2009005 (pdf)
Significance:      Dec 31, 2009 Identified By: Self-Revealing
 
Item Type: NCV NonCited Violation Procedure Inadequacy Results In The Tertiary Auxiliary Transformer Breaker Reopening After Alignment To The Bus A finding of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was self-revealed for the licensee's failure to have adequate procedures to ensure that steps were sequenced such that unplanned transients were not initiated. Specifically, the procedure for performing emergency diesel generator train "A" automatic testing allowed steps to be sequenced in a manner such that a jumper used to simulate a station blackout signal was left installed during the restoration of offsite power. Because of the installed jumpers, a transient was initiated on the associated bus and attached equipment during the restoration from testing. In response to the issue, the licensee implemented compensatory corrective actions and corrected the procedure deficiency prior to conducting the same test on the opposite train.
The inspectors determined that the finding was more than minor because it was associated with the Initiating Events Cornerstone attribute of procedure quality 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. The inspectors evaluated the significance of the issue using Inspection Manual Chapter 0609, Appendix G, Checklist 3, and determined that the power availability guidelines were met. Because the finding did not increase the likelihood of a loss of offsite power or degrade the licensee's ability to cope with a loss of offsite power, the finding screened as having very low safety significance (Green). The finding has a cross-cutting aspect in the area of human performance, work practices component, because the procedure was not adequately verified when steps were changed from being sequence-dependent to allow for completion in any order. Specifically, personnel proceeded to change procedure without implementing peer-checking during the validation process to ensure that the change was applicable to all plant conditions (H.4(a)).
Inspection Report# : 2009005 (pdf)
Significance:        Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure To Adequately Analyze The Automatic Fast Transfer Feature That Allowed Operation With Both 4.16-kiloVolt Safety-Related Buses 1-5 And 1-6 Connected To The Reserve Auxiliary Transformer A finding of very low safety-significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," was identified by the inspectors for the failure to perform a power system analysis calculation that would have identified that the fast transfer design feature/scheme was deficient, in that, it allowed an unanalyzed electrical power system alignment where both redundant 4.16-kiloVolt safety-related buses were being supplied by an offsite source via the same transformer. Use of this electrical configuration could have resulted in an out-of-phase transfer, loss of available offsite power to the buses and potential damaging effects on redundant safety related equipment, during a design basis event such as initiation of safety injection signal. When identified, the licensee entered this issue into their corrective action program and implemented interim actions to prohibit use of the fast transfer feature or manually aligning two safety-related buses to be fed from the same transformer during plant operation.
This performance deficiency was more than minor because the failure to perform the required calculation resulted in a condition where the plant was being operated in an unanalyzed configuration where there was reasonable doubt as to the operability of redundant safeguard loads; this concern resulted in issuance of a Licensee Event Report 2007-007-00 on May 21, 2007. Consequently, the potential for damage or loss of power to safety-related loads during an event could have led to unacceptable consequences. The finding screened as being of very low safety-significance (Green) for the Initiating Events Cornerstone because the finding did not contribute to both the likelihood of a reactor trip and the likelihood that mitigating equipment or functions will not be available. The inspectors did not identify a cross cutting aspect associated with this finding because the cause of the performance deficiency was related to a historical design issue and not indicative of current licensee performance.
Inspection Report# : 2009004 (pdf)
Significance:        Sep 30, 2009
 
Identified By: NRC Item Type: NCV NonCited Violation Inadequate Design Analysis For 105-Ton Transfer Cask Lifting Beam A finding of very low safety significance and associated Non-Cited Violation of Title 10 Code of Federal Regulations Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors for the licensees failure to provide an adequate single failure proof design basis analysis for the 105-ton transfer cask-lifting beam. The licensee entered this issue into their corrective action program as condition report CR339267. The licensee revised the design calculation for the 105-ton transfer cask-lifting beam and demonstrated compliance with single failure proof acceptance criteria.
The finding was determined to be more than minor because the finding was associated with the Initiating Events Cornerstone attribute of equipment performance 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. The finding was determined to be of very low safety significance by the NRCs significance determination process because the transfer cask-lifting beam had not been previously used at the Kewaunee Power Station. This finding has a cross-cutting aspect in the area of human performance, work practices, because the licensee did not ensure supervisory and management oversight of work activities, including contractors, such that nuclear safety is supported, in that, the licensee failed to perform an effective owners review to assure that appropriate design methods are used in calculations that demonstrate nuclear safety (H.4(c)).
Inspection Report# : 2009004 (pdf)
Mitigating Systems Significance:        Jun 30, 2010 Identified By: NRC Item Type: FIN Finding Inappropriate Use of a Probabilistic Methodology in an Operability Determination A finding of very low safety significance was identified by the inspectors for an inadequate operability determination performed for the emergency diesel generators. Specifically, the licensee used TORMIS, a computer code and probabilistic-based methodology, for assessing tornado missile protection and confirming operability of their emergency diesel generator fuel oil day tank vents and storage tank vents. Probabilistic risk assessments were not allowed for confirming operability under both NRC guidance and the licensees procedures. The licensee entered this issue into their corrective action program as condition report 347741, performed a causal evaluation and took compensatory measures until modifications were complete in September 2009.
The finding was determined to be more than minor because the finding was associated with the Mitigating Systems Cornerstone attribute of protection against external events 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, the closure of the emergency diesel generator fuel oil day tank or storage tank vent path as a result of tornado-generated missile striking the vent lines would adversely affect the availability, reliability, and capability of the emergency diesel generators. 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 Mitigating Systems Cornerstone. The inspectors answered "no" to the Mitigating Systems questions and screened the finding as having very low significance (Green).
The inspectors did not identify a cross cutting aspect associated with this finding.
Inspection Report# : 2010003 (pdf)
Significance:        Mar 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Incorrect Curve Was Incorporated Into Calibration Surveillance Procedures A finding of very low safety-significance and associated Non-Cited Violation of
 
10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for inadequate surveillance calibration procedures.
Specifically, calibration surveillance procedure SP-06-034B-1, Steam Generator Flow Mismatch and Steam Pressure Instrument Channel 1, failed to have the correct negative ramp curve. The curve was required to ensure that the low steam line pressure safety injection lag circuitry unit did not exceed the Technical Specification setpoint value. This condition also existed in calibration procedures for channels 2, 3, and 4.
The licensee subsequently entered the issue into its corrective action program as CR 367826 and CR 367932. The licensee conducted an apparent cause evaluation and corrective actions were in progress at the conclusion of the inspection period.
The finding was determined to be more than minor because it was associated with the Mitigating System Cornerstone attribute of procedure quality and adversely affected the cornerstone objective of ensuring the reliability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to ensure that the low steam line pressure safety injection lag circuitry units did not exceed the Technical Specification value of less than or equal to 2 seconds. The finding was of very low safety-significance (Green) based on a phase 1 screening in accordance with Inspection Manual Chapter 0609, Appendix A, ASignificance Determination of Reactor Inspection Findings for At-Power Situations." The finding has a cross-cutting aspect in the areas of human performance, work practices, because the licensee failed to ensure that the calculation upon which the surveillance procedure was based, was approved prior to issuance of the procedure (H.4(b)).
Inspection Report# : 2010002 (pdf)
Significance:        Feb 12, 2010 Identified By: NRC Item Type: NCV NonCited Violation Calculation Methodology Did Not Represent Actual Plant Equipment Configuration A finding of very low safety-significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors for the licensees failure to assure that the calculation methodology represented the actual plant equipment configuration and that adequate design reviews were performed for verifying or checking the adequacy of design. Specifically, the licensee failed to assure that the methodology used in calculation C11716, MCC [Motor Control Center] Control Circuit Voltage Drop, Revision 1, correctly represented the sequence of operation for the various devices contained within the plant equipments control circuitry, such that the minimum required MCC voltage was available for proper circuit operation. Upon discovery of this condition, the licensee performed a preliminary evaluation and entered the finding into their corrective action program (CR366627 and CR366865).
This finding was more than minor in accordance with IMC 0612, Appendix B because the finding was associated with the design control attribute of the mitigating systems cornerstone and affected the cornerstones objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the inadequate MCC voltages could render the safety-related loads required to mitigate the consequences of a design basis accident inoperable and not available. In addition, as a result of the calculation errors, the inspectors were concerned that unsubstantiated MCC voltage values could be used in future calculations and modifications to plant equipment. To resolve the inspectors concerns, the licensee completed an interim evaluation, which evaluated the calculations other circuit models and associated cases. Although, by the end of the inspection, the licensee was able to demonstrate operability; at the time of discovery there was reasonable doubt on the operability of the control circuits modeled in the calculation. The finding was of very low safety-significance based on a Phase 1 screening in accordance with IMC 0609, Significance Determination Process, 609.04, Phase 1 - Initial Screening and Characterization of Findings, Table 4a.
This finding has a cross-cutting aspect in the area of human performance, work practices because the licensee did not ensure supervisory and management oversight of work activities, including contractors, such that nuclear safety was supported. Specifically, the licensee failed to assure that the calculation methodology represented the actual plant equipment configuration and that adequate design reviews were performed for verifying or checking the adequacy of
 
design. (H.4(c)) (Section 1R17.2b)
Inspection Report# : 2010007 (pdf)
Significance:        Dec 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure To Perform Dye Penetrant Examinations Of The Full Code Required Exam Surfaces The inspectors identified a finding of very low safety significance and an associated Non-Cited Violation of 10 CFR 50.55a(g)(4) for the failure to perform dye penetrant examinations of the full required exam surface on safety injection (SI) gas collection chamber welds (SI-W603, SI-W604, and SI-H109) in accordance with the American Society of Mechanical Engineers Section XI Code. Specifically, the examiner proceeded with the examination without anticipating the effects of the increased dwell and drying times of the developer due to cooler ambient temperature than those he had been working under previously. The developer, which would normally dry to a white residue shortly after application to a warm surface and aid in determining the extent of application, remained somewhat translucent when applied to the cooler surface, masking the extent of coverage. This resulted in the examiner's failure to coat the full required Code areas of the welds he was examining and his failure to recognize the lack of coverage. The licensee subsequently re-performed the dye penetrant examinations and entered this issue into their corrective action program.
The inspectors determined that the finding was more than minor because it was associated with the Mitigating System Cornerstone attribute of equipment performance and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Absent NRC intervention, the licensee would not have performed the full Code required examination of welds SI-W603, SI W604, and SI-H109 for an indefinite period of service, which would have placed the reactor coolant pressure boundary at increased risk for unanalyzed cracking, leakage, or component failure. This finding was of very low safety significance because a qualified examination was subsequently performed with no relevant indications detected.
In particular, it did not result in the loss of function of the mitigating system. The inspectors determined that the finding had a cross-cutting aspect in the area of human performance, work practices component, because the licensee proceeded in the face of uncertainty or unexpected circumstances (H.4(a)).
Inspection Report# : 2009005 (pdf)
Significance:        Dec 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Latching Pawl On Safety-Related Bus Tie Breakers Fails To Engage Due To Grease Hardening The inspectors identified a finding of very low safety significance and an associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," for the failure to promptly identify and correct deficiencies that had caused 4160-Volt alternating current breaker failures, which, if corrected, may have prevented subsequent similar failures. Specifically, the licensee did not evaluate other safety-related breakers after hardened grease was identified in the safety-related bus 5 to bus 6 crosstie breakers. In response to this finding, the licensee entered the issue into its corrective action program as Condition Report (CR) 360677.
The inspectors determined that the finding was more than minor because it was associated with the Mitigating Systems Cornerstone attribute of procedure quality 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 inspectors determined the finding could be evaluated using the Significance Determination Process in accordance with Inspection Manual Chapter 0609, "Significance Determination Process," Attachment 0609.04, "Phase 1, Initial Screening and Characterization of Findings," Table 4a, for the Mitigating Systems Cornerstone, dated January 10, 2008. The significance of the finding was determined to be of very low safety significance because the inspectors answered "no" to all of the questions in the Mitigating Systems Cornerstone column. The inspectors determined that the issue had a cross-cutting aspect in human performance, work practices component, because licensee staff did not comply with the timeliness aspects of completing an apparent cause evaluation in accordance with procedure guidance (H.4(b)).
Inspection Report# : 2009005 (pdf)
 
Significance:      Aug 20, 2009 Identified By: NRC Item Type: NCV NonCited Violation Improper Application of 440Vac Rated Motors The inspectors identified a finding of very low safety-significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to ensure the proper application of safety-related 440Vac motors.
Specifically, eight 440Vac safety-related motors were not suitable for operation at analyzed voltages. This finding was entered into the licensees corrective action program.
The finding was more than minor because if left uncorrected it could result in the loss of safety-related 440Vac motors by overstressing of the motor windings through exposure to higher than design rated voltages, and in the failure of motor drive components caused by increased torque produced at the higher voltages. The finding was determined to be of very low safety-significance (Green) because it did not result in a loss of operability. The cause of this finding is related to the cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, because the licensee did not identify this issue completely, accurately, and in a timely manner. The values were produced in a calculation but the licensee did not identify that they exceeded the acceptance criteria. (P.1(a)) (Section 1R21.3)
Inspection Report# : 2009006 (pdf)
Significance:      Aug 20, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inaccurate Minimum Low Head Safety Injection Flow Specified in Emergency Operating Procedure The inspectors identified a finding of very low safety-significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to specify the appropriate quantitative acceptance criterion to assure that adequate Emergency Core Cooling System flow would be delivered to the core following switchover to containment sump recirculation. This finding was entered into the licensees corrective action program.
The finding was determined to be more than minor because the licensee failed to include the appropriate quantitative set-point value for the minimum low-head safety injection train flow following switchover to containment sump recirculation to assure sufficient reactor coolant was available. This finding is of very low safety-significance (Green) because it did not result in a loss of operability, did not represent an actual loss of safety function, and is not potentially risk-significant due to external events. The cause of this finding is related to the cross-cutting aspect in the area of Human Performance, Work Practices, because the licensee did not ensure proper supervisory and management oversight of contractor work activities. Vendor calculations were used as the basis for an EOP set-point without taking into account specific plant design information such as instrument uncertainties, flow instrument calibration effects, and RHR minimum flow. (H.4(c)) (Section 1R21.3)
Inspection Report# : 2009006 (pdf)
Significance:      Aug 20, 2009 Identified By: NRC Item Type: NCV NonCited Violation Inadequate procedure for a Battery Room Flooding Event.
A finding of very low safety significance and associated Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the failure to provide adequate procedural direction to respond to a rupture of the service water piping in the battery rooms. As part of its corrective actions, the licensee revised OP-KW-AOP-MDS-001, Abnormal Operation of Miscellaneous Drains and Sumps, to correct the inadequate operator actions.
The finding was determined to be more than minor because the licensee failed to provide adequate procedural direction for a battery room A or B flood caused by a rupture of the SW piping to/from the battery room fan coil unit in the affected battery room, which ensured the protection of the battery in the unaffected room not associated with the
 
initial flooding event. This finding is of very low safety significance (Green) because it did not result in a loss of operability, did not represent an actual loss of safety function, and is not potentially risk-significant due to external events. The cause of this finding is related to the cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, because the licensee did not fully evaluate the battery room flooding event (an issue potentially impacting nuclear safety) such that the resolution addressed causes, and extent of condition as necessary, to assure nuclear safety. (P.1(c)). (Section 1R21.6).
Inspection Report# : 2009006 (pdf)
Barrier Integrity Significance:      Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Containment Isolation Valve Inoperable With No Technical Specification Action requirement Entry A finding of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the licensees failure to have adequate procedures that ensured technical specifications were entered and followed for containment isolation valves. The licensee entered the issue into their corrective action program as Condition Report 344856 and Condition Report 350526A, and provided additional guidance to operations personnel. At the end of the inspection period, the licensee continued to perform a causal analysis.
The inspectors determined that the issue was more than minor because the finding, if left uncorrected, would become a more significant safety concern. Specifically, not entering the appropriate technical specification action requirements, when necessary, would lead to more significant safety concerns. The inspectors determined the finding could be evaluated using the Significance Determination Process in accordance with Inspection Manual Chapter 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of findings, Table 4a for the Barrier Integrity Cornerstone. The inspectors answered no to the Barrier Integrity questions and screened the finding as having very low safety significance (Green). The finding has a cross-cutting aspect in the area of human performance, resources, because the licensee did not have complete, accurate and up-to-date design documentation, procedures and work packages (H.2(c)).
Inspection Report# : 2009004 (pdf)
Significance:      Aug 20, 2009 Identified By: NRC Item Type: NCV NonCited Violation Non-Conservative Main Steam Line Break Analysis The inspectors identified a finding of very low safety significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to correctly translate the design bases for the maximum steam generator narrow range level into procedures and instructions. This finding was entered into the licensees corrective action program.
The finding was determined to be more than minor because an evaluation was required to ensure that accident analysis requirements for peak containment pressure were met. The finding also impacted the Barrier Integrity cornerstone attribute of procedure quality, and affected the cornerstone objective of maintaining the functionality of containment to protect the public from radionuclide releases caused by accidents or events. Procedural guidance was not adequate to maintain the plant within the parameters specified in the analysis for containment operability. The finding screened as having very low safety significance (Green) because there was no actual barrier degradation. The inspectors determined there was no cross-cutting aspect associated with this finding. (Section 1R21.4)
Inspection Report# : 2009006 (pdf)
Significance: SL-IV Aug 20, 2009 Identified By: NRC Item Type: NCV NonCited Violation
 
Inadequate 50.59 Evaluation of Shutdown Loss of Coolant Accident Procedure.
The inspectors identified a finding of very low safety significance and associated Non-Cited Violation of 10 CFR 50.59(d)(1) for the licensees failure to perform an adequate review of an abnormal operating procedure associated with a shutdown loss of coolant accident. As part of its corrective actions, the licensee revised procedure OP-KW-AOP-RHR-002 to remove the procedure applicability to the Cold Shutdown mode and Refueling mode with reactor vessel head on.
The inspectors determined that the finding was more than minor because it could not reasonably be determined that the activity would not ultimately have required NRC approval. Operation in accordance with the procedure may have challenged the reactor coolant system barrier. The inspectors determined that the finding did not require a quantitative assessment per IMC 0609, Appendix G. Therefore, the finding screened as having very low safety significance (Green) and was determined to be a Severity Level IV violation. The cause of this finding is related to the cross-cutting aspect in the area of Human Performance, Decision Making, because the licensee failed to use conservative assumptions in decision making to demonstrate that the proposed action to include additional modes of applicability for the Shutdown LOCA procedure was safe in order to proceed. (H.1(b)) (Section 1R21.6)
Inspection Report# : 2009006 (pdf)
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.
Miscellaneous Significance: SL-IV Sep 30, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure To Follow Independent Spent Fuel Storage Installation Loading Procedure Step The inspectors identified a Severity Level IV Non-Cited Violation of 10 CFR 72.150, Instructions, Procedures, and Drawings, during the Independent Spent Fuel Storage Installation loading campaign. The licensee failed to follow procedure OP KW NOP ISF 001, Dry Shielded Canister Loading. The inspectors determined that the licensees failure to follow step 5.2.6 of Procedure OP-KW-NOP-ISF-001 to perform a crane brake check was contrary to 10 CFR 72.150. The licensee immediately evaluated the situation and discussed the need to check the crane brakes when lifting loads approaching the rated loads with the refueling crew to prevent missing this step in the future.
The inspectors determined that the violation had more than minor safety significance because the failure to check the crane brakes, results in not knowing if the brakes are functioning properly, which may lead to a failure of the brakes while lifting a loaded spent fuel canister. The issue was addressed by traditional enforcement since 10 CFR Part 72 is
 
not risk based and is not covered under the reactor oversight process. Because this violation was of very low safety significance, was non-repetitive and non-willful, and was entered into the corrective action program, this violation is being treated as a Non-Cited Violation of 10 CFR 72.150 consistent with Section VI.A.1 of the Enforcement Policy.
The inspectors determined that there was no cross-cutting aspect associated with this finding.
Inspection Report# : 2009004 (pdf)
Last modified : September 02, 2010
 
Kewaunee 3Q/2010 Plant Inspection Findings Initiating Events Significance:        Mar 31, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Fuel Loading Occurs With Boron Concentration Below Required Minimum A finding of very low safety significance and associated Non-Cited Violation of Technical Specification 3.8.a.5 was self-revealed when the licensee loaded fuel into the reactor with reactor coolant system boron sample results less than the minimum boron concentration as specified in the core operating limits report. Once the licensee believed the boron concentration samples were accurate and that boron concentration was below the required minimum, operators stopped moving fuel until the boron concentration was restored to acceptable limits. The licensee entered the issue into the corrective action program as Condition Report 351923. The licensee conducted an apparent cause evaluation and proposed long-term corrective actions, including procedure enhancements, operator training on the event, and conservative decision making training.
This finding was determined to be more than minor because it was associated with the Initiating Events Cornerstone attribute of human performance and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown operations. Specifically, the licensee did not believe the initial boron sample results and continued to move fuel with actual boron concentrations below the minimum value specified in the core operating limits report. The inspectors determined that the finding could be evaluated in accordance with Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process. The inspectors used Checklist 4 contained in Attachment 1 and determined that the finding did not require a phase 2 or phase 3 analysis and screened as very low safety significance (Green).
This finding has a cross-cutting aspect in the area of human performance, decision-making, because the licensee failed to use conservative assumptions when making decisions and did not demonstrate that nuclear safety was an overriding priority (H.1(b)).
Inspection Report# : 2010002 (pdf)
Significance:        Mar 31, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Incorrect Settings On Differential Relay Results In Loss Of Tertiary Auxiliary Transformer A finding of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was self-revealed for the failure to establish adequate measures to identify and control design interfaces and coordinate among participating design organizations. Specifically, the licensee failed to adequately control all required tertiary auxiliary transformer relay inputs/settings that interfaced with the existing plant design. This adversely impacted associated equipment and caused an unanticipated system response. The licensee promptly cleared tags on the reserve auxiliary transformer to restore a normal offsite power source to one of the two 4160-volt safeguards buses. The licensee performed a root cause evaluation and implemented corrective actions, some of which included: modifying the design change process to ensure that all programmable digital device setpoints and inputs were
 
identified; documenting the basis for each setpoint or input in the design change documentation; and providing programmable digital device training for design engineering and maintenance personnel. The licensee entered the issue into its corrective action program as CR 352878.
The finding was determined to be more than minor because the finding was associated with the Initiating Events Cornerstone attribute of design control and adversely affected the cornerstone objective of limiting 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 adequately control all required tertiary auxiliary transformer relay inputs/settings adversely impacted the associated equipment, which caused an unanticipated system response and challenged core shutdown cooling.
The inspectors determined that the finding could be evaluated in accordance with Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process. The inspectors used Checklist 4, contained in Attachment 1, and determined that the finding required a Phase 2 analysis because it degraded the ability to recover the decay heat removal system. The Region III senior reactor analyst performed a phase 2 and subsequently a phase 3 analysis and determined the finding was of very low safety significance (Green).
This finding has a cross-cutting aspect in the area of human performance, resources, because the licensee did not maintain complete, accurate, and up-to-date design documentation (H.2(c)).
Inspection Report# : 2010002 (pdf)
Significance:      Dec 31, 2009 Identified By: Self-Revealing Item Type: NCV NonCited Violation Inadequate Work Instructions Lead to Component Cooling Water Relief Valve Lift And Surge Tank Level Drop A finding of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was self-revealed for the failure to have adequate work instructions in place during the isolation of component cooling water (CCW) flow in the reactor coolant pump vaults.
Specifically, the inadequate valve isolation sequence and the speed at which the outlet valves were closed caused CCW system relief valves to lift and rapidly drain the component cooling water surge tank while the CCW system was supporting the residual heat removal system for decay heat removal. In response to the issue, the licensee implemented compensatory corrective actions to modify the tagout and hang tags on the appropriate CCW isolation valves.
The inspectors determined that the finding was more than minor because it was associated with the Initiating Events Cornerstone attribute of configuration control and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations.
The inspectors determined that the finding could be evaluated in accordance with Inspection Manual Chapter 0609, Appendix G, "Shutdown Operations Significance Determination Process." The inspectors used Checklist 3 contained in Attachment 1 and determined that the finding required a Phase 2 analysis since the finding increased the likelihood that a loss of decay heat removal would occur. The Region III senior reactor analyst performed the assessment using Appendix G, Attachment 2, "Phase 2 Significance Determination Process Template for PWR [Pressurized Water Reactor] During Shutdown," and determined that this issue is best characterized as a finding of very low safety significance (Green). This finding has a cross-cutting aspect in the area of human performance, resources component, because the licensee did not maintain long-term plant safety by maintenance of design margins. Specifically, the work instruction did not adequately account for the low design margin that existed between the system operating pressure and the relief valve setpoints when both CCW pumps were running (H.2(a)).
Inspection Report# : 2009005 (pdf)
Significance:      Dec 31, 2009 Identified By: Self-Revealing
 
Item Type: NCV NonCited Violation Procedure Inadequacy Results In The Tertiary Auxiliary Transformer Breaker Reopening After Alignment To The Bus A finding of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," was self-revealed for the licensee's failure to have adequate procedures to ensure that steps were sequenced such that unplanned transients were not initiated. Specifically, the procedure for performing emergency diesel generator train "A" automatic testing allowed steps to be sequenced in a manner such that a jumper used to simulate a station blackout signal was left installed during the restoration of offsite power. Because of the installed jumpers, a transient was initiated on the associated bus and attached equipment during the restoration from testing. In response to the issue, the licensee implemented compensatory corrective actions and corrected the procedure deficiency prior to conducting the same test on the opposite train.
The inspectors determined that the finding was more than minor because it was associated with the Initiating Events Cornerstone attribute of procedure quality 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. The inspectors evaluated the significance of the issue using Inspection Manual Chapter 0609, Appendix G, Checklist 3, and determined that the power availability guidelines were met. Because the finding did not increase the likelihood of a loss of offsite power or degrade the licensee's ability to cope with a loss of offsite power, the finding screened as having very low safety significance (Green). The finding has a cross-cutting aspect in the area of human performance, work practices component, because the procedure was not adequately verified when steps were changed from being sequence-dependent to allow for completion in any order. Specifically, personnel proceeded to change procedure without implementing peer-checking during the validation process to ensure that the change was applicable to all plant conditions (H.4(a)).
Inspection Report# : 2009005 (pdf)
Mitigating Systems Significance:      Sep 30, 2010 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Emergency Operating Procedure A finding of very low safety significance and associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the licensees failure to have an adequate emergency operating procedure for an activity affecting quality. Specifically, emergency operating procedure E 2, Faulted Steam Generator Isolation, did not prescribe actions to manually close the steam supplies to the turbine-driven auxiliary feedwater pump in the event the control room switches failed to operate. The licensee initiated condition report (CR) CR391458 and took immediate corrective actions to correct the deficient procedure and informed the licensed operators.
The finding was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of procedure quality and adversely affected the cornerstone objective of ensuring the reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to ensure that Emergency Operating Procedure E 2 contained all the required actions to ensure successful isolation of a faulted steam generator. The inspectors determined the finding could be evaluated using the significance determination process (SDP) in accordance with Inspection Manual Chapter (IMC) 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Tables 3b and 4a, for the Mitigating Systems Cornerstone. The inspectors answered "no" to the Mitigating Systems questions and screened the finding as having very low significance (Green). The inspectors determined that this finding did not reflect present performance since the procedure error was introduced greater than three years ago; therefore, there was no cross cutting aspect associated with this finding.
Inspection Report# : 2010004 (pdf)
 
Significance:        Sep 30, 2010 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Barrier Control Procedures Result In Exposed Service Water Pumps A finding of very low safety significance and associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the failure to have adequate procedures to address the removal of the screenhouse traveling water screen covers, an activity affecting quality. Consequently, the covers were removed and safety related equipment was exposed to the environment without adequate planning of mitigation actions in the event of inclement weather. The licensee initiated condition reports (CR) CR394670, CR395541, and CR395717 to document the issue. At the end of the inspection period, the licensee was performing a causal evaluation and developing corrective actions to address the issue.
The finding was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of protection against external factors and adversely 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 inspectors determined the finding could be evaluated using the significance determination process (SDP) in accordance with Inspection Manual Chapter (IMC) 0609, Significance Determination Process, 609.04, Phase 1 Initial Screening and Characterization of Findings, Tables 3b, 4a, and 4b for the Mitigating Systems Cornerstone. The inspectors determined that the screenhouse covers were designed to prevent tornado missiles from damaging the safety related equipment housed inside the screenhouse and that two trains of the service water system would be degraded; therefore, the inspectors answered yes to the Table 4b seismic, flooding, and severe weather screening criteria questions 1 and 2. The inspectors contacted the RIII senior reactor analyst who determined, using NUREG/CR 4461, Tornado Climatology of the Contiguous United States, and the number of days the covers were removed that the performance deficiency risk was of very low safety significance (Green). The finding has a cross cutting aspect in the area of human performance, Decision Making, because the licensee failed to make safety significant or risk significant decisions using a systematic process to ensure safety is maintained.
Specifically, the licensee applied an incorrect evaluation to a situation that resulted in the multiple trains of service water pumps being unprotected from tornado missiles (H.1(a)).
Inspection Report# : 2010004 (pdf)
Significance: SL-IV Sep 30, 2010 Identified By: NRC Item Type: NCV NonCited Violation Replacement of Automatic Action With An Operator Manual Action Without Prior NRC Approval A Severity Level IV non-cited violation (NCV) of 10 CFR 50.59(d)(1), Changes, Tests, and Experiments, was identified by the inspectors for the failure to document an evaluation that provided a basis for the determination that the changes implemented in DCR 3163 and Emergency Operating Procedure ES 1.3, Transfer to Sump Recirculation, in 2001 did not require a license amendment. Specifically, the licensee failed to provide an evaluation that adequately documented why replacing the automatic opening of the service water (SW) valves SW 1300A and SW 1300B upon a safety injection signal (to support the service water safety function of loss of coolant accident (LOCA) recirculation operation) with a manual action to open the valves in Emergency Operating Procedure ES 1.3, did not present more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component (SSC) important to safety previously evaluated in the updated safety analysis report. The licensee initiated CR389330 and, at the end of the inspection period, planned to submit a license amendment request to the NRC for this design change.
The violation was determined to be more than minor because the inspectors could not reasonably determine that the changes would not have ultimately required prior NRC approval. Violations of 10 CFR 50.59 are dispositioned using the traditional enforcement process instead of the significance determination process (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 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, Tables 3b and 4a, for the Mitigating Systems Cornerstone. The inspectors answered "yes" to question 1 of the Mitigating Systems Cornerstone column of the Phase 1 worksheet because the inspectors concluded that this was a design basis deficiency confirmed not to result in the loss of operability. Based upon this Phase 1 screening, the inspectors
 
concluded that the issue was of very low safety significance (Green). In accordance with Section 6.1.d.2 of the NRC Enforcement Policy this violation is categorized as Severity Level IV because the resulting changes were evaluated by the SDP as having very low safety significance. The inspectors determined that this finding did not reflect present performance since the error was introduced in a design change that was greater than three years old; therefore, there was no cross cutting aspect associated with this finding.
Inspection Report# : 2010004 (pdf)
Significance:        Jun 30, 2010 Identified By: NRC Item Type: FIN Finding Inappropriate Use of a Probabilistic Methodology in an Operability Determination A finding of very low safety significance was identified by the inspectors for an inadequate operability determination performed for the emergency diesel generators. Specifically, the licensee used TORMIS, a computer code and probabilistic-based methodology, for assessing tornado missile protection and confirming operability of their emergency diesel generator fuel oil day tank vents and storage tank vents. Probabilistic risk assessments were not allowed for confirming operability under both NRC guidance and the licensees procedures. The licensee entered this issue into their corrective action program as condition report 347741, performed a causal evaluation and took compensatory measures until modifications were complete in September 2009.
The finding was determined to be more than minor because the finding was associated with the Mitigating Systems Cornerstone attribute of protection against external events 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, the closure of the emergency diesel generator fuel oil day tank or storage tank vent path as a result of tornado-generated missile striking the vent lines would adversely affect the availability, reliability, and capability of the emergency diesel generators. 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 Mitigating Systems Cornerstone. The inspectors answered "no" to the Mitigating Systems questions and screened the finding as having very low significance (Green).
The inspectors did not identify a cross cutting aspect associated with this finding.
Inspection Report# : 2010003 (pdf)
Significance:        Mar 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Incorrect Curve Was Incorporated Into Calibration Surveillance Procedures A finding of very low safety-significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for inadequate surveillance calibration procedures.
Specifically, calibration surveillance procedure SP-06-034B-1, Steam Generator Flow Mismatch and Steam Pressure Instrument Channel 1, failed to have the correct negative ramp curve. The curve was required to ensure that the low steam line pressure safety injection lag circuitry unit did not exceed the Technical Specification setpoint value. This condition also existed in calibration procedures for channels 2, 3, and 4.
The licensee subsequently entered the issue into its corrective action program as CR 367826 and CR 367932. The licensee conducted an apparent cause evaluation and corrective actions were in progress at the conclusion of the inspection period.
The finding was determined to be more than minor because it was associated with the Mitigating System Cornerstone attribute of procedure quality and adversely affected the cornerstone objective of ensuring the reliability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to ensure that the low steam line pressure safety injection lag circuitry units did not exceed the Technical Specification value of less than or equal to 2 seconds. The finding was of very low safety-significance (Green) based on a phase 1 screening in accordance with Inspection Manual Chapter 0609, Appendix A, ASignificance Determination of Reactor Inspection Findings for
 
At-Power Situations." The finding has a cross-cutting aspect in the areas of human performance, work practices, because the licensee failed to ensure that the calculation upon which the surveillance procedure was based, was approved prior to issuance of the procedure (H.4(b)).
Inspection Report# : 2010002 (pdf)
Significance:        Feb 12, 2010 Identified By: NRC Item Type: NCV NonCited Violation Calculation Methodology Did Not Represent Actual Plant Equipment Configuration A finding of very low safety-significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors for the licensees failure to assure that the calculation methodology represented the actual plant equipment configuration and that adequate design reviews were performed for verifying or checking the adequacy of design. Specifically, the licensee failed to assure that the methodology used in calculation C11716, MCC [Motor Control Center] Control Circuit Voltage Drop, Revision 1, correctly represented the sequence of operation for the various devices contained within the plant equipments control circuitry, such that the minimum required MCC voltage was available for proper circuit operation. Upon discovery of this condition, the licensee performed a preliminary evaluation and entered the finding into their corrective action program (CR366627 and CR366865).
This finding was more than minor in accordance with IMC 0612, Appendix B because the finding was associated with the design control attribute of the mitigating systems cornerstone and affected the cornerstones objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the inadequate MCC voltages could render the safety-related loads required to mitigate the consequences of a design basis accident inoperable and not available. In addition, as a result of the calculation errors, the inspectors were concerned that unsubstantiated MCC voltage values could be used in future calculations and modifications to plant equipment. To resolve the inspectors concerns, the licensee completed an interim evaluation, which evaluated the calculations other circuit models and associated cases. Although, by the end of the inspection, the licensee was able to demonstrate operability; at the time of discovery there was reasonable doubt on the operability of the control circuits modeled in the calculation. The finding was of very low safety-significance based on a Phase 1 screening in accordance with IMC 0609, Significance Determination Process, 609.04, Phase 1 - Initial Screening and Characterization of Findings, Table 4a.
This finding has a cross-cutting aspect in the area of human performance, work practices because the licensee did not ensure supervisory and management oversight of work activities, including contractors, such that nuclear safety was supported. Specifically, the licensee failed to assure that the calculation methodology represented the actual plant equipment configuration and that adequate design reviews were performed for verifying or checking the adequacy of design. (H.4(c)) (Section 1R17.2b)
Inspection Report# : 2010007 (pdf)
Significance:        Dec 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Failure To Perform Dye Penetrant Examinations Of The Full Code Required Exam Surfaces The inspectors identified a finding of very low safety significance and an associated Non-Cited Violation of 10 CFR 50.55a(g)(4) for the failure to perform dye penetrant examinations of the full required exam surface on safety injection (SI) gas collection chamber welds (SI-W603, SI-W604, and SI-H109) in accordance with the American Society of Mechanical Engineers Section XI Code. Specifically, the examiner proceeded with the examination without anticipating the effects of the increased dwell and drying times of the developer due to cooler ambient temperature than those he had been working under previously. The developer, which would normally dry to a white residue shortly after application to a warm surface and aid in determining the extent of application, remained somewhat translucent when applied to the cooler surface, masking the extent of coverage. This resulted in the examiner's failure to coat the full required Code areas of the welds he was examining and his failure to recognize the lack of coverage. The licensee subsequently re-performed the dye penetrant examinations and entered this issue into their corrective action program.
 
The inspectors determined that the finding was more than minor because it was associated with the Mitigating System Cornerstone attribute of equipment performance and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Absent NRC intervention, the licensee would not have performed the full Code required examination of welds SI-W603, SI W604, and SI-H109 for an indefinite period of service, which would have placed the reactor coolant pressure boundary at increased risk for unanalyzed cracking, leakage, or component failure. This finding was of very low safety significance because a qualified examination was subsequently performed with no relevant indications detected.
In particular, it did not result in the loss of function of the mitigating system. The inspectors determined that the finding had a cross-cutting aspect in the area of human performance, work practices component, because the licensee proceeded in the face of uncertainty or unexpected circumstances (H.4(a)).
Inspection Report# : 2009005 (pdf)
Significance:        Dec 31, 2009 Identified By: NRC Item Type: NCV NonCited Violation Latching Pawl On Safety-Related Bus Tie Breakers Fails To Engage Due To Grease Hardening The inspectors identified a finding of very low safety significance and an associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," for the failure to promptly identify and correct deficiencies that had caused 4160-Volt alternating current breaker failures, which, if corrected, may have prevented subsequent similar failures. Specifically, the licensee did not evaluate other safety-related breakers after hardened grease was identified in the safety-related bus 5 to bus 6 crosstie breakers. In response to this finding, the licensee entered the issue into its corrective action program as Condition Report (CR) 360677.
The inspectors determined that the finding was more than minor because it was associated with the Mitigating Systems Cornerstone attribute of procedure quality 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 inspectors determined the finding could be evaluated using the Significance Determination Process in accordance with Inspection Manual Chapter 0609, "Significance Determination Process," Attachment 0609.04, "Phase 1, Initial Screening and Characterization of Findings," Table 4a, for the Mitigating Systems Cornerstone, dated January 10, 2008. The significance of the finding was determined to be of very low safety significance because the inspectors answered "no" to all of the questions in the Mitigating Systems Cornerstone column. The inspectors determined that the issue had a cross-cutting aspect in human performance, work practices component, because licensee staff did not comply with the timeliness aspects of completing an apparent cause evaluation in accordance with procedure guidance (H.4(b)).
Inspection Report# : 2009005 (pdf)
Barrier Integrity Significance:        Sep 03, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct the Classification of a Containment Isolation Valve.
A finding of very low safety significance and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified by the inspectors for the failure to correct a condition adverse to quality.
Specifically, the licensee failed to provide their licensed operators with correct procedures and instructions for determining which valves were containment isolation valves. The condition was previously identified on August 12, 2009, when the inspectors found MS 100A, the steam supply to the turbine driven auxiliary feedwater pump, open without the capability to be remotely closed from the control room and without a technical specification entry for the containment isolation function. The licensee entered the issue, during the current inspection, into their corrective action program and took short-term corrective actions of placing a standing order in the control room directing operators to enter the appropriate containment isolation technical specifications for the valves in question.
 
The finding was determined to be more than minor, because, if left uncorrected, has the potential to lead to a more significant safety concern. The inspectors concluded this finding was associated with the Barrier Integrity Cornerstone. 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. The inspectors answered no to the Barrier Integrity Cornerstone questions and screened the finding as having very low safety significance (Green). This finding has a cross-cutting aspect in the area of human performance within the resources component because the licensee did not maintain complete, accurate and up-to-date design documentation (H.2(c)). (Section 4OA2.1.b(2))
Inspection Report# : 2010006 (pdf)
Significance: SL-IV Sep 03, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Update the Updated Safety Analysis Report to Include Containment Penetration Leakage Testing Information.
The inspectors identified a Severity Level IV, non-cited violation of 10 CFR 50.71(e), Maintenance of Records, Making of Reports, having very low safety significance. The inspectors found that the licensee failed to update the Updated Safety Analysis Report (USAR) to describe for each containment penetration, the penetration category, the type of leakage test required, and the applicable leakage test method. The licensee entered this into their corrective action program. The inspectors found the violation to be more than minor in accordance with the NRC Enforcement Policy, Section 6.1.d, Example 3, in that the failure to update the Final Safety Analysis Report (FSAR) would not have a material impact on safety or licensed activities. This issue was determined to be a Severity Level IV violation since it was similar to a Severity Level IV violation example in the NRC Enforcement Policy. Additionally, in accordance with the Enforcement Policy, this violation is categorized as Severity Level IV because the resulting changes were evaluated by the SDP as having very low safety significance (Green).
Violations of 10 CFR 50.71 are dispositioned using the traditional enforcement process instead of the significance determination process (SDP) because they are considered to be violations that potentially impede or impact the regulatory process. The underlying finding is evaluated under the SDP to determine the significance of the violation.
In this case, the finding was determined to be more than minor because, if left uncorrected, it had the potential to lead to a more significant safety concern. (Section 4OA2.1.b(2))
The SDP portion of this issue is tracked as item 2010-006-03.
Inspection Report# : 2010006 (pdf)
Significance:        Sep 03, 2010 Identified By: NRC Item Type: FIN Finding Failure to Update the Updated Safety Analysis Report to Include Containment Penetration Leakage Testing Information.
The inspectors identified a finding associated with a traditional enforcement Severity Level IV, non-cited violation of 10 CFR 50.71(e), Maintenance of Records, Making of Reports, having very low safety significance. The resulting changes were evaluated by the SDP as having very low safety significance (Green).
The underlying finding was evaluated under the SDP to determine the significance of the violation. In this case, the finding was determined to be more than minor because, if left uncorrected, it had the potential to lead to a more significant safety concern. 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. The inspectors answered no to the Barrier Integrity Cornerstone questions and screened the finding as having very low safety significance (Green). The inspectors did not identify a cross-cutting aspect associated with the finding because the finding was not representative of current performance. (Section 4OA2.1.b(2))
The Traditional Enforcement portion of this issue is tracked as item 2010-006-02.
Inspection Report# : 2010006 (pdf)
 
Emergency Preparedness Significance: SL-IV Sep 07, 2010 Identified By: NRC Item Type: NCV NonCited Violation Changes to EAL Technical Bases Document Decreases the Effectiveness of the Plan without Prior NRC Approval.
The inspector identified a Severity Level IV 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. Specifically, the licensee changed wording in their EAL technical basis document for EAL SU5 and CU1, RCS Leakage. The new wording eliminates leakage from the charging and letdown systems from consideration as RCS Leakage and therefore, leakage from these systems that meet the EAL thresholds would not constitute an Unusual Event declaration, using the licensees revised wording. This change was made without prior NRC approval.
The Green finding associated with this Item 05000305/2010502-02.
Inspection Report# : 2010502 (pdf)
Significance:      Sep 07, 2010 Identified By: NRC Item Type: FIN Finding Changes Made to EAL Technical Bases that Decreased the Effectiveness The inspector identified a Green finding 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. Specifically, the licensee changed wording in their EAL technical basis document for EAL SU5 and CU1, RCS Leakage. The new wording eliminates leakage from the charging and letdown systems from consideration as RCS Leakage and therefore, leakage from these systems that meet the EAL thresholds would not constitute an Unusual Event declaration, using the licensees revised wording. This change was made without prior NRC approval.
The performance deficiency was more than minor and of very low safety-significance using MC 0612 and MC 0609, Appendix B, 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. Using MC 0609, Appendix B, the inspectors determined that the finding had a very low safety significance. The inspectors also determined that the finding had a cross-cutting aspect in the area of Human Performance, decision making because the licensee did not recognize that the change that was made to the EAL Technical Basis document decreased the effectiveness of the emergency plan.
(H.1.(b)) (Section 1EP4)
The associated SLIV is Item 05000305/2010502-01.
Inspection Report# : 2010502 (pdf)
Occupational Radiation Safety Significance:      Sep 30, 2010 Identified By: NRC Item Type: NCV NonCited Violation
 
Unauthorized Entry into an HRA A finding of very low safety significance and an associated non-cited violation (NCV) of Technical Specification 6.13 was identified by the inspectors after a worker entered a high radiation area on October 15, 2009. Radiation protection did not authorize the worker to enter the area nor was the worker made knowledgeable of the dose rate level in the area. The work was temporarily assigned from the turbine building to the containment building to assist with the cleaning of containment in preparation for containment close out. The worker received a briefing from radiation protection regarding the radiological condition of containment, but was instructed not to enter any high radiation areas. The worker entered the radiological controlled area on radiation work permit 09-0202-1, which allowed access to containment but did not allow access to high radiation areas and the electronic dosimeter worn by the worker was set to alarm at 50 mrem/hour. During the course of the work activity, the worker was instructed to retrieve a piece of equipment from the basement elevation of containment. An unknown individual held the swing gate open, which also blocked the HRA posting, and the worker entered the basement elevation of containment. The worker, alerted to the higher dose rate conditions through an electronic dosimeter alarm, then exited the work area. The worker immediately reported the event to the radiation protection staff who confirmed the basement elevation of containment was a posted HRA and the dose rates were greater than 100 mrem/hour. The maximum dose rate measured by the ED was 106 mrem/hour. The corrective actions taken by the licensee included temporarily restricting the individual's further access to the radiologically controlled area and counseling of the individual by the licensee's Radiation Protection Manager.
The inspectors identified Example 6(h) of inspection manual chapter (IMC) 0612, Appendix E, as similar to the performance issue, in that, the worker was neither authorized by radiation protection to work in specific locations within containment, nor was the worker made knowledgeable of the dose rate level in the area. Therefore, in accordance with IMC 0612 and Example 6(h) of Appendix E, the inspectors determined that the performance deficiency was more than minor. Additionally, the performance deficiency impacted 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, unauthorized entry into areas without knowledge of the radiological conditions placed the worker at increased risk for unnecessary radiation exposure. The finding was assessed using the Occupational Radiation Safety significance determination process (SDP) and was determined to be of very low safety significance because the problem was not an as low as is reasonably achievable planning issue, there were no overexposures nor substantial potential for overexposures given the worker's reaction to the electronic dosimeter alarm and the dose rate ranges, and the licensees ability to assess dose was not compromised. The inspectors determined that the cause of this incident involved a cross cutting component in the human performance area for inadequate work control. Specifically, the licensee did not appropriately coordinate work activities by incorporating necessary to assure human performance (H.3(b)).
Inspection Report# : 2010004 (pdf)
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.
Miscellaneous Last modified : November 29, 2010
 
Kewaunee 4Q/2010 Plant Inspection Findings Initiating Events Significance:      Dec 31, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure To Follow Red Channel Instrument Test Procedure A finding of very low safety significance and associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed when a nuclear control operator (NCO) failed to perform a procedure step, which resulted in the main feedwater regulating valve FW 7A partially closing while the reactor was at full power. Specifically, Step 6.11.2 of procedure SP-47-316A, Channel 1 (Red) Instrument Channel Test Channel Operational Test, directed the NCO to place the main feedwater regulating valve FW 7A in manual to preclude valve movement during a simulated portion of the test; however, the NCO marked the step "not applicable" and subsequently did not perform it. The licensee initiated condition reports (CRs) CR396649 and CR405809, performed an apparent cause evaluation (ACE), and initiated corrective actions (CAs) to address the issues identified in the causal evaluation.
The finding was determined to be more than minor in accordance with Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated December 24, 2009, because it was associated with the Initiating Events Cornerstone attribute of human performance and adversely affected the cornerstone objective of limiting 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 follow the procedure initiated a secondary-side plant transient. The inspectors determined the finding could be evaluated using the Significance Determination Process (SDP) in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Tables 3b and 4a for the Initiating Events Cornerstone, dated January 10, 2008. The inspectors answered "no" to the Initiating Events Cornerstone Transient Initiator question and screened the finding as having very low significance (Green). The finding has a cross-cutting aspect in the area of human performance, Work Practices, because the personnel work practices did not support human performance. Specifically, licensee personnel failed to follow procedures (H.4(b)).
Inspection Report# : 2010005 (pdf)
Significance:      Mar 31, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Fuel Loading Occurs With Boron Concentration Below Required Minimum A finding of very low safety significance and associated Non-Cited Violation of Technical Specification 3.8.a.5 was self-revealed when the licensee loaded fuel into the reactor with reactor coolant system boron sample results less than the minimum boron concentration as specified in the core operating limits report. Once the licensee believed the boron concentration samples were accurate and that boron concentration was below the required minimum, operators stopped moving fuel until the boron concentration was restored to acceptable limits. The licensee entered the issue into the corrective action program as Condition Report 351923. The licensee conducted an apparent cause evaluation and proposed long-term corrective actions, including procedure enhancements, operator training on the event, and conservative decision making training.
This finding was determined to be more than minor because it was associated with the Initiating Events Cornerstone attribute of human performance and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and
 
challenge critical safety functions during shutdown operations. Specifically, the licensee did not believe the initial boron sample results and continued to move fuel with actual boron concentrations below the minimum value specified in the core operating limits report. The inspectors determined that the finding could be evaluated in accordance with Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process. The inspectors used Checklist 4 contained in Attachment 1 and determined that the finding did not require a phase 2 or phase 3 analysis and screened as very low safety significance (Green).
This finding has a cross-cutting aspect in the area of human performance, decision-making, because the licensee failed to use conservative assumptions when making decisions and did not demonstrate that nuclear safety was an overriding priority (H.1(b)).
Inspection Report# : 2010002 (pdf)
Significance:        Mar 31, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Incorrect Settings On Differential Relay Results In Loss Of Tertiary Auxiliary Transformer A finding of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was self-revealed for the failure to establish adequate measures to identify and control design interfaces and coordinate among participating design organizations. Specifically, the licensee failed to adequately control all required tertiary auxiliary transformer relay inputs/settings that interfaced with the existing plant design. This adversely impacted associated equipment and caused an unanticipated system response. The licensee promptly cleared tags on the reserve auxiliary transformer to restore a normal offsite power source to one of the two 4160-volt safeguards buses. The licensee performed a root cause evaluation and implemented corrective actions, some of which included: modifying the design change process to ensure that all programmable digital device setpoints and inputs were identified; documenting the basis for each setpoint or input in the design change documentation; and providing programmable digital device training for design engineering and maintenance personnel. The licensee entered the issue into its corrective action program as CR 352878.
The finding was determined to be more than minor because the finding was associated with the Initiating Events Cornerstone attribute of design control and adversely affected the cornerstone objective of limiting 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 adequately control all required tertiary auxiliary transformer relay inputs/settings adversely impacted the associated equipment, which caused an unanticipated system response and challenged core shutdown cooling.
The inspectors determined that the finding could be evaluated in accordance with Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process. The inspectors used Checklist 4, contained in Attachment 1, and determined that the finding required a Phase 2 analysis because it degraded the ability to recover the decay heat removal system. The Region III senior reactor analyst performed a phase 2 and subsequently a phase 3 analysis and determined the finding was of very low safety significance (Green).
This finding has a cross-cutting aspect in the area of human performance, resources, because the licensee did not maintain complete, accurate, and up-to-date design documentation (H.2(c)).
Inspection Report# : 2010002 (pdf)
Mitigating Systems
 
Significance:        Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Incorrect Containment Fan Coil Unit Acceptance Criteria A finding of very low safety significance and associated non cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors for the failure to correctly translate the applicable regulatory requirements and the design basis into procedures and instructions. Specifically, the licensee failed to adequately translate the containment fan coil unit (CFCU) service water flow acceptance criteria from the current design basis calculations into the CFCU performance monitoring procedures, which resulted in the incorrect acceptance criteria in plant test procedures. The licensee took immediate corrective actions to correct the acceptance criteria in the test procedures and to perform an operability determination on CFCU C, the only one of the four CFCUs that showed a recent decrease in flow. At the end of the inspection period, the licensee was completing an apparent cause evaluation and developing additional long term corrective actions.
The finding was determined to be more than minor in accordance with Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated December 24, 2009, because it was associated with the Mitigating Systems Cornerstone attribute of procedure quality and adversely affected the cornerstone objective of ensuring the reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to ensure that procedure PMP 18 13, "Containment Fan Coil Unit Performance Monitoring (AQ-1)," contained the correct acceptance criteria for testing the CFCUs. The inspectors determined the finding could be evaluated using the Significance Determination Process (SDP) in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Tables 3b and 4a for the Mitigating Systems Cornerstone, dated January 10, 2008.
The inspectors answered "no" to the Mitigating Systems questions and screened the finding as having very low significance (Green). This finding has a cross cutting aspect in the area of human performance, Resources, because the licensee did not maintain complete, accurate, and up to date procedures. Specifically, the correct acceptance criteria for testing the CFCUs from the design basis calculations were not specified in the CFCU testing procedure (H.2(c)).
Inspection Report# : 2010005 (pdf)
Significance:        Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Unacceptable Preconditioning of Safety-Related Pressure Switches A finding of very low significance and associated non cited violation (NCV) of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the failure to develop and implement an adequate surveillance test procedure to accurately assess the as found trip setpoint for the pressure switches associated with the turbine building service water isolation function and various other safety related functions. The licensee initiated condition report CR401813, performed an apparent cause evaluation, and initiated corrective actions to address the issues identified in the casual evaluation.
The finding was determined to be more than minor in accordance with Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated December 24, 2009, because it was associated with the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences and affected the cornerstone attribute of Equipment Performance. The inspectors determined the finding could be evaluated using the Significance Determination Process (SDP) in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Tables 3b and 4a for the Mitigating Systems Cornerstone, dated January 10, 2008. The inspectors answered no to the Mitigating Systems questions and screened the finding as having very low significance (Green). This finding has a cross cutting aspect in the area of problem identification and resolution, Operating Experience, because the licensee did not evaluate and communicate external operating experience to internal stakeholders in a timely manner (P.2(a)).
Inspection Report# : 2010005 (pdf)
Significance:        Dec 31, 2010
 
Identified By: NRC Item Type: NCV NonCited Violation Inappropriate Isolation of the Safety Injection Pump Minimum Flow Recirculation Lines A finding of very low safety significance and associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for multiple inadequate procedures, which directed closing the common train safety injection minimum flow recirculation line valves, an activity affecting quality. Specifically, station procedures directed operators to close the safety injection pump minimum flow recirculation valves in order to complete valve timing tests, and to engage an interlock that allowed closure of the containment sump recirculation valves. However, the procedures and licensed operators failed to recognize that closure of either minimum flow recirculation valve affected the operability and availability of both safety injection pumps for certain design basis accidents because the minimum flow recirculation path was isolated.
The licensee subsequently entered the issue into its corrective action program as CR393930. The licensee corrected the procedure inadequacies and completed a root cause evaluation that recommended several corrective actions to prevent recurrence.
The finding was determined to be more than minor in accordance with Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated December 24, 2009, because it was associated with the Mitigating System Cornerstone attribute of procedure quality, 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 ensure that procedures implemented during power operations ensured the operability of both trains of safety injection. The inspectors determined the finding could be evaluated using the Significance Determination Process (SDP) in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Tables 3b and 4a for the Mitigating Systems Cornerstone, dated January 10, 2008. The inspectors answered "yes" to the Mitigating Systems question that confirmed the finding represented a loss of system safety function. The Region III Senior Reactor Analyst (SRA) performed an SDP Phase 2 analysis and a Phase 3 analysis. The Phase 3 analysis determined that the resultant delta core damage frequency (CDF) was less than 1E 6 and delta large early release frequency (LERF) was less than 1E 7, which represented a Green finding. The dominant scenario involved a small break loss of coolant accident with operator failure to perform a rapid cool down. The finding has a cross cutting aspect in the area of human performance, Decision Making, because although the licensee procedures cautioned that starting a safety injection pump following the closure of a minimum flow recirculation valve would result in damage to the pump, the licensee staff failed to use conservative decision making to question the adequacy of the prescribed procedure actions (H.1(b)).
Inspection Report# : 2010005 (pdf)
Significance:        Sep 30, 2010 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Emergency Operating Procedure A finding of very low safety significance and associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the licensees failure to have an adequate emergency operating procedure for an activity affecting quality. Specifically, emergency operating procedure E 2, Faulted Steam Generator Isolation, did not prescribe actions to manually close the steam supplies to the turbine-driven auxiliary feedwater pump in the event the control room switches failed to operate. The licensee initiated condition report (CR) CR391458 and took immediate corrective actions to correct the deficient procedure and informed the licensed operators.
The finding was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of procedure quality and adversely affected the cornerstone objective of ensuring the reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to ensure that Emergency Operating Procedure E 2 contained all the required actions to ensure successful isolation of a faulted steam generator. The inspectors determined the finding could be evaluated using the significance determination process (SDP) in accordance with Inspection Manual Chapter (IMC) 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Tables 3b and 4a, for the Mitigating Systems Cornerstone. The inspectors answered "no" to the Mitigating Systems questions
 
and screened the finding as having very low significance (Green). The inspectors determined that this finding did not reflect present performance since the procedure error was introduced greater than three years ago; therefore, there was no cross cutting aspect associated with this finding.
Inspection Report# : 2010004 (pdf)
Significance:        Sep 30, 2010 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Barrier Control Procedures Result In Exposed Service Water Pumps A finding of very low safety significance and associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the failure to have adequate procedures to address the removal of the screenhouse traveling water screen covers, an activity affecting quality. Consequently, the covers were removed and safety related equipment was exposed to the environment without adequate planning of mitigation actions in the event of inclement weather. The licensee initiated condition reports (CR) CR394670, CR395541, and CR395717 to document the issue. At the end of the inspection period, the licensee was performing a causal evaluation and developing corrective actions to address the issue.
The finding was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of protection against external factors and adversely 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 inspectors determined the finding could be evaluated using the significance determination process (SDP) in accordance with Inspection Manual Chapter (IMC) 0609, Significance Determination Process, 609.04, Phase 1 Initial Screening and Characterization of Findings, Tables 3b, 4a, and 4b for the Mitigating Systems Cornerstone. The inspectors determined that the screenhouse covers were designed to prevent tornado missiles from damaging the safety related equipment housed inside the screenhouse and that two trains of the service water system would be degraded; therefore, the inspectors answered yes to the Table 4b seismic, flooding, and severe weather screening criteria questions 1 and 2. The inspectors contacted the RIII senior reactor analyst who determined, using NUREG/CR 4461, Tornado Climatology of the Contiguous United States, and the number of days the covers were removed that the performance deficiency risk was of very low safety significance (Green). The finding has a cross cutting aspect in the area of human performance, Decision Making, because the licensee failed to make safety significant or risk significant decisions using a systematic process to ensure safety is maintained.
Specifically, the licensee applied an incorrect evaluation to a situation that resulted in the multiple trains of service water pumps being unprotected from tornado missiles (H.1(a)).
Inspection Report# : 2010004 (pdf)
Significance: SL-IV Sep 30, 2010 Identified By: NRC Item Type: NCV NonCited Violation Replacement of Automatic Action With An Operator Manual Action Without Prior NRC Approval A Severity Level IV non-cited violation (NCV) of 10 CFR 50.59(d)(1), Changes, Tests, and Experiments, was identified by the inspectors for the failure to document an evaluation that provided a basis for the determination that the changes implemented in DCR 3163 and Emergency Operating Procedure ES 1.3, Transfer to Sump Recirculation, in 2001 did not require a license amendment. Specifically, the licensee failed to provide an evaluation that adequately documented why replacing the automatic opening of the service water (SW) valves SW 1300A and SW 1300B upon a safety injection signal (to support the service water safety function of loss of coolant accident (LOCA) recirculation operation) with a manual action to open the valves in Emergency Operating Procedure ES 1.3, did not present more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component (SSC) important to safety previously evaluated in the updated safety analysis report. The licensee initiated CR389330 and, at the end of the inspection period, planned to submit a license amendment request to the NRC for this design change.
The violation was determined to be more than minor because the inspectors could not reasonably determine that the changes would not have ultimately required prior NRC approval. Violations of 10 CFR 50.59 are dispositioned using the traditional enforcement process instead of the significance determination process (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 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, Tables 3b and 4a, for the Mitigating Systems Cornerstone. The inspectors answered "yes" to question 1 of the Mitigating Systems Cornerstone column of the Phase 1 worksheet because the inspectors concluded that this was a design basis deficiency confirmed not to result in the loss of operability. Based upon this Phase 1 screening, the inspectors concluded that the issue was of very low safety significance (Green). In accordance with Section 6.1.d.2 of the NRC Enforcement Policy this violation is categorized as Severity Level IV because the resulting changes were evaluated by the SDP as having very low safety significance. The inspectors determined that this finding did not reflect present performance since the error was introduced in a design change that was greater than three years old; therefore, there was no cross cutting aspect associated with this finding.
See related performance deficinecy item 2010-004-04.
Inspection Report# : 2010004 (pdf)
Significance:      Sep 30, 2010 Identified By: NRC Item Type: FIN Finding Performance Deficiency Associated With SLIV NCV - Replace of Automatic Actions With an Operator Manual Action Without Prior NRC Approval.
A Severity Level IV NCV of 10 CFR 50.59(d)(1), Changes, Tests, and Experiments, was identified by the inspectors for the failure to document an evaluation that provided a basis for the determination that the changes implemented in DCR 3163 and Emergency Operating Procedure ES 1.3, Transfer to Sump Recirculation, in 2001 did not require a license amendment. Specifically, the licensee failed to provide an evaluation that adequately documented why replacing the automatic opening of the service water (SW) valves SW 1300A and SW 1300B upon a safety injection signal (to support the service water safety function of loss of coolant accident (LOCA) recirculation operation) with a manual action to open the valves in Emergency Operating Procedure ES 1.3, did not present more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component (SSC) important to safety previously evaluated in the updated safety analysis report. The licensee initiated CR389330 and, at the end of the inspection period, planned to submit a license amendment request to the NRC for this design change.
The violation was determined to be more than minor because the inspectors could not reasonably determine that the changes would not have ultimately required prior NRC approval. 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 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, Tables 3b and 4a, for the Mitigating Systems Cornerstone. The inspectors answered "yes" to question 1 of the Mitigating Systems Cornerstone column of the Phase 1 worksheet because the inspectors concluded that this was a design basis deficiency confirmed not to result in the loss of operability. Based upon this Phase 1 screening, the inspectors concluded that the issue was of very low safety significance (Green). In accordance with Section 6.1.d.2 of the NRC Enforcement Policy this violation is categorized as Severity Level IV because the resulting changes were evaluated by the SDP as having very low safety significance.
The inspectors determined that this finding did not reflect present performance since the error was introduced in a design change that was greater than three years old; therefore, there was no cross cutting aspect associated with this finding.
See related SL IV NCV item 2010-004-03.
Inspection Report# : 2010004 (pdf)
Significance:      Jun 30, 2010 Identified By: NRC Item Type: FIN Finding Inappropriate Use of a Probabilistic Methodology in an Operability Determination A finding of very low safety significance was identified by the inspectors for an inadequate operability determination performed for the emergency diesel generators. Specifically, the licensee used TORMIS, a computer code and
 
probabilistic-based methodology, for assessing tornado missile protection and confirming operability of their emergency diesel generator fuel oil day tank vents and storage tank vents. Probabilistic risk assessments were not allowed for confirming operability under both NRC guidance and the licensees procedures. The licensee entered this issue into their corrective action program as condition report 347741, performed a causal evaluation and took compensatory measures until modifications were complete in September 2009.
The finding was determined to be more than minor because the finding was associated with the Mitigating Systems Cornerstone attribute of protection against external events 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, the closure of the emergency diesel generator fuel oil day tank or storage tank vent path as a result of tornado-generated missile striking the vent lines would adversely affect the availability, reliability, and capability of the emergency diesel generators. 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 Mitigating Systems Cornerstone. The inspectors answered "no" to the Mitigating Systems questions and screened the finding as having very low significance (Green).
The inspectors did not identify a cross cutting aspect associated with this finding.
Inspection Report# : 2010003 (pdf)
Significance:        Mar 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Incorrect Curve Was Incorporated Into Calibration Surveillance Procedures A finding of very low safety-significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for inadequate surveillance calibration procedures.
Specifically, calibration surveillance procedure SP-06-034B-1, Steam Generator Flow Mismatch and Steam Pressure Instrument Channel 1, failed to have the correct negative ramp curve. The curve was required to ensure that the low steam line pressure safety injection lag circuitry unit did not exceed the Technical Specification setpoint value. This condition also existed in calibration procedures for channels 2, 3, and 4.
The licensee subsequently entered the issue into its corrective action program as CR 367826 and CR 367932. The licensee conducted an apparent cause evaluation and corrective actions were in progress at the conclusion of the inspection period.
The finding was determined to be more than minor because it was associated with the Mitigating System Cornerstone attribute of procedure quality and adversely affected the cornerstone objective of ensuring the reliability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to ensure that the low steam line pressure safety injection lag circuitry units did not exceed the Technical Specification value of less than or equal to 2 seconds. The finding was of very low safety-significance (Green) based on a phase 1 screening in accordance with Inspection Manual Chapter 0609, Appendix A, ASignificance Determination of Reactor Inspection Findings for At-Power Situations." The finding has a cross-cutting aspect in the areas of human performance, work practices, because the licensee failed to ensure that the calculation upon which the surveillance procedure was based, was approved prior to issuance of the procedure (H.4(b)).
Inspection Report# : 2010002 (pdf)
Significance:        Feb 12, 2010 Identified By: NRC Item Type: NCV NonCited Violation Calculation Methodology Did Not Represent Actual Plant Equipment Configuration A finding of very low safety-significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors for the licensees failure to assure that the calculation methodology
 
represented the actual plant equipment configuration and that adequate design reviews were performed for verifying or checking the adequacy of design. Specifically, the licensee failed to assure that the methodology used in calculation C11716, MCC [Motor Control Center] Control Circuit Voltage Drop, Revision 1, correctly represented the sequence of operation for the various devices contained within the plant equipments control circuitry, such that the minimum required MCC voltage was available for proper circuit operation. Upon discovery of this condition, the licensee performed a preliminary evaluation and entered the finding into their corrective action program (CR366627 and CR366865).
This finding was more than minor in accordance with IMC 0612, Appendix B because the finding was associated with the design control attribute of the mitigating systems cornerstone and affected the cornerstones objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the inadequate MCC voltages could render the safety-related loads required to mitigate the consequences of a design basis accident inoperable and not available. In addition, as a result of the calculation errors, the inspectors were concerned that unsubstantiated MCC voltage values could be used in future calculations and modifications to plant equipment. To resolve the inspectors concerns, the licensee completed an interim evaluation, which evaluated the calculations other circuit models and associated cases. Although, by the end of the inspection, the licensee was able to demonstrate operability; at the time of discovery there was reasonable doubt on the operability of the control circuits modeled in the calculation. The finding was of very low safety-significance based on a Phase 1 screening in accordance with IMC 0609, Significance Determination Process, 609.04, Phase 1 - Initial Screening and Characterization of Findings, Table 4a.
This finding has a cross-cutting aspect in the area of human performance, work practices because the licensee did not ensure supervisory and management oversight of work activities, including contractors, such that nuclear safety was supported. Specifically, the licensee failed to assure that the calculation methodology represented the actual plant equipment configuration and that adequate design reviews were performed for verifying or checking the adequacy of design. (H.4(c))
Inspection Report# : 2010007 (pdf)
Barrier Integrity Significance:        Sep 03, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct the Classification of a Containment Isolation Valve A finding of very low safety significance and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified by the inspectors for the failure to correct a condition adverse to quality.
Specifically, the licensee failed to provide their licensed operators with correct procedures and instructions for determining which valves were containment isolation valves. The condition was previously identified on August 12, 2009, when the inspectors found MS 100A, the steam supply to the turbine driven auxiliary feedwater pump, open without the capability to be remotely closed from the control room and without a technical specification entry for the containment isolation function. The licensee entered the issue, during the current inspection, into their corrective action program and took short-term corrective actions of placing a standing order in the control room directing operators to enter the appropriate containment isolation technical specifications for the valves in question.
The finding was determined to be more than minor, because, if left uncorrected, has the potential to lead to a more significant safety concern. The inspectors concluded this finding was associated with the Barrier Integrity Cornerstone. 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. The inspectors answered no to the Barrier Integrity Cornerstone questions and screened the finding as having very low safety significance (Green). This finding has a cross-cutting aspect in the area of human performance within the resources component because the licensee did not maintain complete, accurate and up-to-date design documentation (H.2(c)).
Inspection Report# : 2010006 (pdf)
 
Significance: SL-IV Sep 03, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Update the Updated Safety Analysis Report to Include Containment Penetration Leakage Testing Information The inspectors identified a Severity Level IV, non-cited violation of 10 CFR 50.71(e), Maintenance of Records, Making of Reports, having very low safety significance. The inspectors found that the licensee failed to update the Updated Safety Analysis Report (USAR) to describe for each containment penetration, the penetration category, the type of leakage test required, and the applicable leakage test method. The licensee entered this into their corrective action program. The inspectors found the violation to be more than minor in accordance with the NRC Enforcement Policy, Section 6.1.d, Example 3, in that the failure to update the Final Safety Analysis Report (FSAR) would not have a material impact on safety or licensed activities. This issue was determined to be a Severity Level IV violation since it was similar to a Severity Level IV violation example in the NRC Enforcement Policy. Additionally, in accordance with the Enforcement Policy, this violation is categorized as Severity Level IV because the resulting changes were evaluated by the SDP as having very low safety significance (Green).
Violations of 10 CFR 50.71 are dispositioned using the traditional enforcement process instead of the significance determination process (SDP) because they are considered to be violations that potentially impede or impact the regulatory process. The underlying finding is evaluated under the SDP to determine the significance of the violation.
In this case, the finding was determined to be more than minor because, if left uncorrected, it had the potential to lead to a more significant safety concern.
The SDP portion of this issue is tracked as item 2010-006-03.
Inspection Report# : 2010006 (pdf)
Significance:        Sep 03, 2010 Identified By: NRC Item Type: FIN Finding Failure to Update the Updated Safety Analysis Report to Include Containment Penetration Leakage Testing Information.
The inspectors identified a finding associated with a traditional enforcement Severity Level IV, non-cited violation of 10 CFR 50.71(e), Maintenance of Records, Making of Reports, having very low safety significance. The resulting changes were evaluated by the SDP as having very low safety significance (Green).
The underlying finding was evaluated under the SDP to determine the significance of the violation. In this case, the finding was determined to be more than minor because, if left uncorrected, it had the potential to lead to a more significant safety concern. 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. The inspectors answered no to the Barrier Integrity Cornerstone questions and screened the finding as having very low safety significance (Green). The inspectors did not identify a cross-cutting aspect associated with the finding because the finding was not representative of current performance.
The Traditional Enforcement portion of this issue is tracked as item 2010-006-02.
Inspection Report# : 2010006 (pdf)
Emergency Preparedness Significance: SL-IV Sep 07, 2010 Identified By: NRC Item Type: NCV NonCited Violation Changes to EAL Technical Bases Document Decreases the Effectiveness of the Plan Without Prior NRC Approval The inspector identified a Severity Level IV NCV of 10 CFR 50.54(q) associated with 10 CFR 50.47(b)(4) because
 
the licensee failed to obtain prior NRC approval for a change made to its emergency plan that decreased the effectiveness of the plan. Specifically, the licensee changed wording in their EAL technical basis document for EAL SU5 and CU1, RCS Leakage. The new wording eliminates leakage from the charging and letdown systems from consideration as RCS Leakage and therefore, leakage from these systems that meet the EAL thresholds would not constitute an Unusual Event declaration, using the licensees revised wording. This change was made without prior NRC approval.
The Green finding associated with this Item 05000305/2010502-02.
Inspection Report# : 2010502 (pdf)
Significance:      Sep 07, 2010 Identified By: NRC Item Type: FIN Finding Changes Made to EAL Technical Bases that Decreased the Effectiveness The inspector identified a Green finding 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. Specifically, the licensee changed wording in their EAL technical basis document for EAL SU5 and CU1, RCS Leakage. The new wording eliminates leakage from the charging and letdown systems from consideration as RCS Leakage and therefore, leakage from these systems that meet the EAL thresholds would not constitute an Unusual Event declaration, using the licensees revised wording. This change was made without prior NRC approval.
The performance deficiency was more than minor and of very low safety-significance using MC 0612 and MC 0609, Appendix B, 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. Using MC 0609, Appendix B, the inspectors determined that the finding had a very low safety significance. The inspectors also determined that the finding had a cross-cutting aspect in the area of Human Performance, decision making because the licensee did not recognize that the change that was made to the EAL Technical Basis document decreased the effectiveness of the emergency plan.
(H.1.(b)) (Section 1EP4)
The associated SLIV is Item 05000305/2010502-01.
Inspection Report# : 2010502 (pdf)
Occupational Radiation Safety Significance:      Sep 30, 2010 Identified By: NRC Item Type: NCV NonCited Violation Unauthorized Entry into an HRA A finding of very low safety significance and an associated non-cited violation (NCV) of Technical Specification 6.13 was identified by the inspectors after a worker entered a high radiation area on October 15, 2009. Radiation protection did not authorize the worker to enter the area nor was the worker made knowledgeable of the dose rate level in the area. The work was temporarily assigned from the turbine building to the containment building to assist with the cleaning of containment in preparation for containment close out. The worker received a briefing from radiation protection regarding the radiological condition of containment, but was instructed not to enter any high radiation areas. The worker entered the radiological controlled area on radiation work permit 09-0202-1, which allowed access to containment but did not allow access to high radiation areas and the electronic dosimeter worn by the worker was set to alarm at 50 mrem/hour. During the course of the work activity, the worker was instructed to retrieve a piece of equipment from the basement elevation of containment. An unknown individual held the swing gate open, which also
 
blocked the HRA posting, and the worker entered the basement elevation of containment. The worker, alerted to the higher dose rate conditions through an electronic dosimeter alarm, then exited the work area. The worker immediately reported the event to the radiation protection staff who confirmed the basement elevation of containment was a posted HRA and the dose rates were greater than 100 mrem/hour. The maximum dose rate measured by the ED was 106 mrem/hour. The corrective actions taken by the licensee included temporarily restricting the individual's further access to the radiologically controlled area and counseling of the individual by the licensee's Radiation Protection Manager.
The inspectors identified Example 6(h) of inspection manual chapter (IMC) 0612, Appendix E, as similar to the performance issue, in that, the worker was neither authorized by radiation protection to work in specific locations within containment, nor was the worker made knowledgeable of the dose rate level in the area. Therefore, in accordance with IMC 0612 and Example 6(h) of Appendix E, the inspectors determined that the performance deficiency was more than minor. Additionally, the performance deficiency impacted 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, unauthorized entry into areas without knowledge of the radiological conditions placed the worker at increased risk for unnecessary radiation exposure. The finding was assessed using the Occupational Radiation Safety significance determination process (SDP) and was determined to be of very low safety significance because the problem was not an as low as is reasonably achievable planning issue, there were no overexposures nor substantial potential for overexposures given the worker's reaction to the electronic dosimeter alarm and the dose rate ranges, and the licensees ability to assess dose was not compromised. The inspectors determined that the cause of this incident involved a cross cutting component in the human performance area for inadequate work control. Specifically, the licensee did not appropriately coordinate work activities by incorporating necessary to assure human performance (H.3(b)).
Inspection Report# : 2010004 (pdf)
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.
Miscellaneous Last modified : March 03, 2011
 
Kewaunee 1Q/2011 Plant Inspection Findings Initiating Events Significance:        Mar 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Misapplication Of Code Acceptance Criteria For Weld Flaws A finding of very low safety-significance and associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion IX, Control of Special Processes, was identified by the inspectors on March 3, 2011, for the licensees failure to establish a procedure that incorporated the American Society of Mechanical Engineers Code acceptance criteria for evaluation of flaws detected during ultrasonic examinations. Consequently, the licensee applied incorrect acceptance criteria to the flaws identified during ultrasonic examination of a weld on the chemical and volume control system seal water injection filter 1A housing. Licensee corrective actions included: evaluation of weld flaws to ensure they met applicable Code criteria and revision of a site procedure to incorporate appropriate Code acceptance criteria.
The finding was determined to be more than minor because the finding, if left uncorrected, would become a more significant safety concern. Absent NRC identification, the failure to provide Code acceptance criteria could have allowed components with unacceptable cracks to be returned to service. Cracks in components returned to service would place safety related piping systems at increased risk for through wall leakage and/or failure. The licensee promptly corrected this issue before components with unacceptable flaws were returned to service. The inspectors answered No to the Significance Determination Process Phase I screening question, Assuming worst case degradation, would the finding result in exceeding the Technical Specification (TS) limit for any reactor coolant system leakage or could the finding have likely affected other mitigation systems resulting in a total loss of their safety function assuming the worst case degradation? Therefore, this finding screened as having very low safety-significance (Green). This finding has a cross-cutting aspect in the area of human performance, work practices, because the licensee did not effectively implement human error prevention techniques. Specifically, the lack of procedure acceptance criteria was caused by inadequate peer checking during the licensees review and approval of the procedure for evaluation of non destructive examination data (H.4(a)).
Inspection Report# : 2011002 (pdf)
Significance:        Mar 31, 2011 Identified By: Self-Revealing Item Type: FIN Finding Partial Loss Of Offsite Power Caused By Less Than Adequate Interface And Oversight Of Switchyard Modification Work A finding of very low safety-significance was self-revealed for the failure to adequately control relay testing for switchyard breaker installations under Design Change WO KW100691871. Specifically, on March 10, 2011, Dominion Electrical Transmission technicians deviated from standard work practices to test a relay via an internal corporate server, which caused a partial loss of offsite power to the plant through the loss of the main auxiliary transformer backfeed to safety-related bus 6. Licensee corrective actions included a human performance and safety stand down for substation personnel on the day of the event, the development of a mitigating strategy that outlined expectations and implemented increased direct supervision on critical tasks, and the development of a formal memo describing expectations related to the restricted use of the server for performing remote testing of control functions.
The finding was determined to be more than minor because, if left uncorrected, the finding had the potential to lead to a more significant safety concern. Specifically, had a different breaker been inappropriately tripped, the station could have experienced a total loss of offsite power. The inspectors concluded that the finding could be evaluated using Inspection Manual Chapter 0609, Appendix M, Significance Determination Process Using Qualitative Criteria.
Specifically, the inspectors qualitatively evaluated the finding by applying the spent fuel pool questions in the Fuel Barrier column of Table 4a, Attachment 4. The inspectors answered "No" to all three questions and determined that
 
the finding was of very low safety-significance (Green). The finding has a cross-cutting aspect in the areas of human performance, work practices, because supervisory and management oversight of work activities, including contractors, was not implemented for this evolution (H.4(c)).
Inspection Report# : 2011002 (pdf)
Significance:      Dec 31, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure To Follow Red Channel Instrument Test Procedure A finding of very low safety significance and associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed when a nuclear control operator (NCO) failed to perform a procedure step, which resulted in the main feedwater regulating valve FW 7A partially closing while the reactor was at full power. Specifically, Step 6.11.2 of procedure SP-47-316A, Channel 1 (Red) Instrument Channel Test Channel Operational Test, directed the NCO to place the main feedwater regulating valve FW 7A in manual to preclude valve movement during a simulated portion of the test; however, the NCO marked the step "not applicable" and subsequently did not perform it. The licensee initiated condition reports (CRs) CR396649 and CR405809, performed an apparent cause evaluation (ACE), and initiated corrective actions (CAs) to address the issues identified in the causal evaluation.
The finding was determined to be more than minor in accordance with Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated December 24, 2009, because it was associated with the Initiating Events Cornerstone attribute of human performance and adversely affected the cornerstone objective of limiting 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 follow the procedure initiated a secondary-side plant transient. The inspectors determined the finding could be evaluated using the Significance Determination Process (SDP) in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Tables 3b and 4a for the Initiating Events Cornerstone, dated January 10, 2008. The inspectors answered "no" to the Initiating Events Cornerstone Transient Initiator question and screened the finding as having very low significance (Green). The finding has a cross-cutting aspect in the area of human performance, Work Practices, because the personnel work practices did not support human performance. Specifically, licensee personnel failed to follow procedures (H.4(b)).
Inspection Report# : 2010005 (pdf)
Mitigating Systems Significance:      Mar 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Work Instructions Results In Potential Orange Path A finding of very low safety-significance and associated non-cited violation (NCV) of Technical Specification 5.4.1, Procedures, was identified by the inspectors for the failure to implement procedures for shutdown operations involving shutdown operations safety assessments. Specifically, OU KW 201, Shutdown Safety Assessment Checklist, step 3.3.1, stated, in part, that a shutdown safety assessment was required to be completed in accordance with the procedure for core cooling; however, the inspectors noted that the February 28, 2011, 6:00 p.m. analysis credited the safety injection system feed and bleed as an available alternate decay heat removal system when the system was not available as described in Section 5.3.2, Available/Availability, for work scheduled at that time on the emergency core cooling system (ECCS) sump. The licensee initiated condition report CR415539, and at the end of the inspection period, the licensee was performing a causal evaluation to determine the causes of the event and develop corrective actions. On February 28, as a remedial corrective action prior to the start of work, additional steps to the work instructions were added to ensure the equipment would meet the intended function, operators were designated to perform the local manual operations and a pre job brief was conducted that provided training for using the equipment in the given situation.
 
The finding was determined to be more than minor because the finding was associated with the Mitigating Systems Cornerstone attribute of human error (pre event) and adversely 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). Specifically, the availability of the ECCS sump was integral to ensuring that the plant was not in an orange risk path for the evolutions completed on February 28. The inspectors screened the finding as of very low safety-significance (Green) because the finding did not degrade the licensees ability to establish an alternate core cooling path if decay heat removal could not be re established and, therefore, did not require a Significance Determination Process phase 2 or phase 3 analysis. The finding has a cross-cutting aspect in the areas of human performance, work control, because the licensee failed to plan the work activities by incorporating the need for planned contingencies and compensatory actions to ensure the ECCS sump was available to ensure an orange risk path for core cooling was not entered (H.3(a)).
Inspection Report# : 2011002 (pdf)
Significance:        Mar 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Unintended Voiding Of The Reactor Vessel Closure Head A finding of very low safety-significance and associated non-cited violation (NCV) of Technical Specification 5.4.1, Procedures, was identified by the inspectors for the failure to establish, implement, and maintain procedures for shutdown operations involving the draining of reactor coolant system (RCS) inventory. Specifically, on March 21, 2011, during a pressurizer draindown evolution, licensed operators unknowingly created a gas void in the reactor vessel closure head (RVCH) that displaced water to a level near the RVCH flange. Subsequent evaluation determined that the procedure for draining the RCS did not contain adequate guidance to ensure that an unacceptable void in the RVCH was not present prior to or formed during operations draindown activities. The licensee subsequently entered the issue into its corrective action program as CR418537 and performed a remedial corrective action of removing the gas void that accumulated in the RVCH. At the end of the inspection period, the licensee was performing an apparent cause evaluation to determine the causes of the event and develop additional corrective actions.
The finding was determined to be more than minor because the finding was associated with the Mitigating Systems Cornerstone attribute of operating procedure quality and adversely 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). Specifically, the formation of the gas void in the RVCH displaced RCS inventory and could have challenged the ability to remove decay heat in the event of a loss of shutdown cooling. The Region III senior reactor analyst determined that this issue is best characterized as a finding of very low safety-significance (Green). The finding has a cross-cutting aspect in the areas of human performance, work practices, because operations personnel did not follow or implement the guidance contained in plant procedures. Specifically, procedure OP KW AOP RC 002 prescribed actions to take if a gas void formed in the RVCH that resulted in RVLIS level readings less than 88 percent, which had occurred several hours prior to the start of a pressurizer draining evolution (H.4(b)).
Inspection Report# : 2011002 (pdf)
Significance:        Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Incorrect Containment Fan Coil Unit Acceptance Criteria A finding of very low safety significance and associated non cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors for the failure to correctly translate the applicable regulatory requirements and the design basis into procedures and instructions. Specifically, the licensee failed to adequately translate the containment fan coil unit (CFCU) service water flow acceptance criteria from the current design basis calculations into the CFCU performance monitoring procedures, which resulted in the incorrect acceptance criteria in plant test procedures. The licensee took immediate corrective actions to correct the acceptance criteria in the test procedures and to perform an operability determination on CFCU C, the only one of the four CFCUs that showed a recent decrease in flow. At the end of the inspection period, the licensee was completing an apparent cause evaluation and developing additional long term corrective actions.
The finding was determined to be more than minor in accordance with Inspection Manual Chapter (IMC) 0612,
 
Power Reactor Inspection Reports, Appendix B, Issue Screening, dated December 24, 2009, because it was associated with the Mitigating Systems Cornerstone attribute of procedure quality and adversely affected the cornerstone objective of ensuring the reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to ensure that procedure PMP 18 13, "Containment Fan Coil Unit Performance Monitoring (AQ-1)," contained the correct acceptance criteria for testing the CFCUs. The inspectors determined the finding could be evaluated using the Significance Determination Process (SDP) in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Tables 3b and 4a for the Mitigating Systems Cornerstone, dated January 10, 2008.
The inspectors answered "no" to the Mitigating Systems questions and screened the finding as having very low significance (Green). This finding has a cross cutting aspect in the area of human performance, Resources, because the licensee did not maintain complete, accurate, and up to date procedures. Specifically, the correct acceptance criteria for testing the CFCUs from the design basis calculations were not specified in the CFCU testing procedure (H.2(c)).
Inspection Report# : 2010005 (pdf)
Significance:      Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Unacceptable Preconditioning of Safety-Related Pressure Switches A finding of very low significance and associated non cited violation (NCV) of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the failure to develop and implement an adequate surveillance test procedure to accurately assess the as found trip setpoint for the pressure switches associated with the turbine building service water isolation function and various other safety related functions. The licensee initiated condition report CR401813, performed an apparent cause evaluation, and initiated corrective actions to address the issues identified in the casual evaluation.
The finding was determined to be more than minor in accordance with Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated December 24, 2009, because it was associated with the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences and affected the cornerstone attribute of Equipment Performance. The inspectors determined the finding could be evaluated using the Significance Determination Process (SDP) in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Tables 3b and 4a for the Mitigating Systems Cornerstone, dated January 10, 2008. The inspectors answered no to the Mitigating Systems questions and screened the finding as having very low significance (Green). This finding has a cross cutting aspect in the area of problem identification and resolution, Operating Experience, because the licensee did not evaluate and communicate external operating experience to internal stakeholders in a timely manner (P.2(a)).
Inspection Report# : 2010005 (pdf)
Significance:      Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Inappropriate Isolation of the Safety Injection Pump Minimum Flow Recirculation Lines A finding of very low safety significance and associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for multiple inadequate procedures, which directed closing the common train safety injection minimum flow recirculation line valves, an activity affecting quality. Specifically, station procedures directed operators to close the safety injection pump minimum flow recirculation valves in order to complete valve timing tests, and to engage an interlock that allowed closure of the containment sump recirculation valves. However, the procedures and licensed operators failed to recognize that closure of either minimum flow recirculation valve affected the operability and availability of both safety injection pumps for certain design basis accidents because the minimum flow recirculation path was isolated.
The licensee subsequently entered the issue into its corrective action program as CR393930. The licensee corrected the procedure inadequacies and completed a root cause evaluation that recommended several corrective actions to prevent recurrence.
The finding was determined to be more than minor in accordance with Inspection Manual Chapter (IMC) 0612,
 
Power Reactor Inspection Reports, Appendix B, Issue Screening, dated December 24, 2009, because it was associated with the Mitigating System Cornerstone attribute of procedure quality, 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 ensure that procedures implemented during power operations ensured the operability of both trains of safety injection. The inspectors determined the finding could be evaluated using the Significance Determination Process (SDP) in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Tables 3b and 4a for the Mitigating Systems Cornerstone, dated January 10, 2008. The inspectors answered "yes" to the Mitigating Systems question that confirmed the finding represented a loss of system safety function. The Region III Senior Reactor Analyst (SRA) performed an SDP Phase 2 analysis and a Phase 3 analysis. The Phase 3 analysis determined that the resultant delta core damage frequency (CDF) was less than 1E 6 and delta large early release frequency (LERF) was less than 1E 7, which represented a Green finding. The dominant scenario involved a small break loss of coolant accident with operator failure to perform a rapid cool down. The finding has a cross cutting aspect in the area of human performance, Decision Making, because although the licensee procedures cautioned that starting a safety injection pump following the closure of a minimum flow recirculation valve would result in damage to the pump, the licensee staff failed to use conservative decision making to question the adequacy of the prescribed procedure actions (H.1(b)).
Inspection Report# : 2010005 (pdf)
Significance:        Sep 30, 2010 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Emergency Operating Procedure A finding of very low safety significance and associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the licensees failure to have an adequate emergency operating procedure for an activity affecting quality. Specifically, emergency operating procedure E 2, Faulted Steam Generator Isolation, did not prescribe actions to manually close the steam supplies to the turbine-driven auxiliary feedwater pump in the event the control room switches failed to operate. The licensee initiated condition report (CR) CR391458 and took immediate corrective actions to correct the deficient procedure and informed the licensed operators.
The finding was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of procedure quality and adversely affected the cornerstone objective of ensuring the reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to ensure that Emergency Operating Procedure E 2 contained all the required actions to ensure successful isolation of a faulted steam generator. The inspectors determined the finding could be evaluated using the significance determination process (SDP) in accordance with Inspection Manual Chapter (IMC) 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Tables 3b and 4a, for the Mitigating Systems Cornerstone. The inspectors answered "no" to the Mitigating Systems questions and screened the finding as having very low significance (Green). The inspectors determined that this finding did not reflect present performance since the procedure error was introduced greater than three years ago; therefore, there was no cross cutting aspect associated with this finding.
Inspection Report# : 2010004 (pdf)
Significance:        Sep 30, 2010 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Barrier Control Procedures Result In Exposed Service Water Pumps A finding of very low safety significance and associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the failure to have adequate procedures to address the removal of the screenhouse traveling water screen covers, an activity affecting quality. Consequently, the covers were removed and safety related equipment was exposed to the environment without adequate planning of mitigation actions in the event of inclement weather. The licensee initiated condition reports (CR) CR394670, CR395541, and CR395717 to document the issue. At the end of the inspection period, the
 
licensee was performing a causal evaluation and developing corrective actions to address the issue.
The finding was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of protection against external factors and adversely 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 inspectors determined the finding could be evaluated using the significance determination process (SDP) in accordance with Inspection Manual Chapter (IMC) 0609, Significance Determination Process, 609.04, Phase 1 Initial Screening and Characterization of Findings, Tables 3b, 4a, and 4b for the Mitigating Systems Cornerstone. The inspectors determined that the screenhouse covers were designed to prevent tornado missiles from damaging the safety related equipment housed inside the screenhouse and that two trains of the service water system would be degraded; therefore, the inspectors answered yes to the Table 4b seismic, flooding, and severe weather screening criteria questions 1 and 2. The inspectors contacted the RIII senior reactor analyst who determined, using NUREG/CR 4461, Tornado Climatology of the Contiguous United States, and the number of days the covers were removed that the performance deficiency risk was of very low safety significance (Green). The finding has a cross cutting aspect in the area of human performance, Decision Making, because the licensee failed to make safety significant or risk significant decisions using a systematic process to ensure safety is maintained.
Specifically, the licensee applied an incorrect evaluation to a situation that resulted in the multiple trains of service water pumps being unprotected from tornado missiles (H.1(a)).
Inspection Report# : 2010004 (pdf)
Significance: SL-IV Sep 30, 2010 Identified By: NRC Item Type: NCV NonCited Violation Replacement of Automatic Action With An Operator Manual Action Without Prior NRC Approval A Severity Level IV non-cited violation (NCV) of 10 CFR 50.59(d)(1), Changes, Tests, and Experiments, was identified by the inspectors for the failure to document an evaluation that provided a basis for the determination that the changes implemented in DCR 3163 and Emergency Operating Procedure ES 1.3, Transfer to Sump Recirculation, in 2001 did not require a license amendment. Specifically, the licensee failed to provide an evaluation that adequately documented why replacing the automatic opening of the service water (SW) valves SW 1300A and SW 1300B upon a safety injection signal (to support the service water safety function of loss of coolant accident (LOCA) recirculation operation) with a manual action to open the valves in Emergency Operating Procedure ES 1.3, did not present more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component (SSC) important to safety previously evaluated in the updated safety analysis report. The licensee initiated CR389330 and, at the end of the inspection period, planned to submit a license amendment request to the NRC for this design change.
The violation was determined to be more than minor because the inspectors could not reasonably determine that the changes would not have ultimately required prior NRC approval. Violations of 10 CFR 50.59 are dispositioned using the traditional enforcement process instead of the significance determination process (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 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, Tables 3b and 4a, for the Mitigating Systems Cornerstone. The inspectors answered "yes" to question 1 of the Mitigating Systems Cornerstone column of the Phase 1 worksheet because the inspectors concluded that this was a design basis deficiency confirmed not to result in the loss of operability. Based upon this Phase 1 screening, the inspectors concluded that the issue was of very low safety significance (Green). In accordance with Section 6.1.d.2 of the NRC Enforcement Policy this violation is categorized as Severity Level IV because the resulting changes were evaluated by the SDP as having very low safety significance. The inspectors determined that this finding did not reflect present performance since the error was introduced in a design change that was greater than three years old; therefore, there was no cross cutting aspect associated with this finding.
See related performance deficinecy item 2010-004-04.
Inspection Report# : 2010004 (pdf)
Significance:        Sep 30, 2010
 
Identified By: NRC Item Type: FIN Finding Performance Deficiency Associated With SLIV NCV - Replace of Automatic Actions With an Operator Manual Action Without Prior NRC Approval.
A Severity Level IV NCV of 10 CFR 50.59(d)(1), Changes, Tests, and Experiments, was identified by the inspectors for the failure to document an evaluation that provided a basis for the determination that the changes implemented in DCR 3163 and Emergency Operating Procedure ES 1.3, Transfer to Sump Recirculation, in 2001 did not require a license amendment. Specifically, the licensee failed to provide an evaluation that adequately documented why replacing the automatic opening of the service water (SW) valves SW 1300A and SW 1300B upon a safety injection signal (to support the service water safety function of loss of coolant accident (LOCA) recirculation operation) with a manual action to open the valves in Emergency Operating Procedure ES 1.3, did not present more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component (SSC) important to safety previously evaluated in the updated safety analysis report. The licensee initiated CR389330 and, at the end of the inspection period, planned to submit a license amendment request to the NRC for this design change.
The violation was determined to be more than minor because the inspectors could not reasonably determine that the changes would not have ultimately required prior NRC approval. 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 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, Tables 3b and 4a, for the Mitigating Systems Cornerstone. The inspectors answered "yes" to question 1 of the Mitigating Systems Cornerstone column of the Phase 1 worksheet because the inspectors concluded that this was a design basis deficiency confirmed not to result in the loss of operability. Based upon this Phase 1 screening, the inspectors concluded that the issue was of very low safety significance (Green). In accordance with Section 6.1.d.2 of the NRC Enforcement Policy this violation is categorized as Severity Level IV because the resulting changes were evaluated by the SDP as having very low safety significance.
The inspectors determined that this finding did not reflect present performance since the error was introduced in a design change that was greater than three years old; therefore, there was no cross cutting aspect associated with this finding.
See related SL IV NCV item 2010-004-03.
Inspection Report# : 2010004 (pdf)
Significance:        Jun 30, 2010 Identified By: NRC Item Type: FIN Finding Inappropriate Use of a Probabilistic Methodology in an Operability Determination A finding of very low safety significance was identified by the inspectors for an inadequate operability determination performed for the emergency diesel generators. Specifically, the licensee used TORMIS, a computer code and probabilistic-based methodology, for assessing tornado missile protection and confirming operability of their emergency diesel generator fuel oil day tank vents and storage tank vents. Probabilistic risk assessments were not allowed for confirming operability under both NRC guidance and the licensees procedures. The licensee entered this issue into their corrective action program as condition report 347741, performed a causal evaluation and took compensatory measures until modifications were complete in September 2009.
The finding was determined to be more than minor because the finding was associated with the Mitigating Systems Cornerstone attribute of protection against external events 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, the closure of the emergency diesel generator fuel oil day tank or storage tank vent path as a result of tornado-generated missile striking the vent lines would adversely affect the availability, reliability, and capability of the emergency diesel generators. 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 Mitigating Systems Cornerstone. The inspectors answered "no" to the Mitigating Systems questions and screened the finding as having very low significance (Green).
The inspectors did not identify a cross cutting aspect associated with this finding.
 
Inspection Report# : 2010003 (pdf)
Barrier Integrity Significance:        Sep 03, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct the Classification of a Containment Isolation Valve A finding of very low safety significance and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified by the inspectors for the failure to correct a condition adverse to quality.
Specifically, the licensee failed to provide their licensed operators with correct procedures and instructions for determining which valves were containment isolation valves. The condition was previously identified on August 12, 2009, when the inspectors found MS 100A, the steam supply to the turbine driven auxiliary feedwater pump, open without the capability to be remotely closed from the control room and without a technical specification entry for the containment isolation function. The licensee entered the issue, during the current inspection, into their corrective action program and took short-term corrective actions of placing a standing order in the control room directing operators to enter the appropriate containment isolation technical specifications for the valves in question.
The finding was determined to be more than minor, because, if left uncorrected, has the potential to lead to a more significant safety concern. The inspectors concluded this finding was associated with the Barrier Integrity Cornerstone. 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. The inspectors answered no to the Barrier Integrity Cornerstone questions and screened the finding as having very low safety significance (Green). This finding has a cross-cutting aspect in the area of human performance within the resources component because the licensee did not maintain complete, accurate and up-to-date design documentation (H.2(c)).
Inspection Report# : 2010006 (pdf)
Significance: SL-IV Sep 03, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Update the Updated Safety Analysis Report to Include Containment Penetration Leakage Testing Information The inspectors identified a Severity Level IV, non-cited violation of 10 CFR 50.71(e), Maintenance of Records, Making of Reports, having very low safety significance. The inspectors found that the licensee failed to update the Updated Safety Analysis Report (USAR) to describe for each containment penetration, the penetration category, the type of leakage test required, and the applicable leakage test method. The licensee entered this into their corrective action program. The inspectors found the violation to be more than minor in accordance with the NRC Enforcement Policy, Section 6.1.d, Example 3, in that the failure to update the Final Safety Analysis Report (FSAR) would not have a material impact on safety or licensed activities. This issue was determined to be a Severity Level IV violation since it was similar to a Severity Level IV violation example in the NRC Enforcement Policy. Additionally, in accordance with the Enforcement Policy, this violation is categorized as Severity Level IV because the resulting changes were evaluated by the SDP as having very low safety significance (Green).
Violations of 10 CFR 50.71 are dispositioned using the traditional enforcement process instead of the significance determination process (SDP) because they are considered to be violations that potentially impede or impact the regulatory process. The underlying finding is evaluated under the SDP to determine the significance of the violation.
In this case, the finding was determined to be more than minor because, if left uncorrected, it had the potential to lead to a more significant safety concern.
The SDP portion of this issue is tracked as item 2010-006-03.
Inspection Report# : 2010006 (pdf)
 
Significance:      Sep 03, 2010 Identified By: NRC Item Type: FIN Finding Failure to Update the Updated Safety Analysis Report to Include Containment Penetration Leakage Testing Information.
The inspectors identified a finding associated with a traditional enforcement Severity Level IV, non-cited violation of 10 CFR 50.71(e), Maintenance of Records, Making of Reports, having very low safety significance. The resulting changes were evaluated by the SDP as having very low safety significance (Green).
The underlying finding was evaluated under the SDP to determine the significance of the violation. In this case, the finding was determined to be more than minor because, if left uncorrected, it had the potential to lead to a more significant safety concern. 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. The inspectors answered no to the Barrier Integrity Cornerstone questions and screened the finding as having very low safety significance (Green). The inspectors did not identify a cross-cutting aspect associated with the finding because the finding was not representative of current performance.
The Traditional Enforcement portion of this issue is tracked as item 2010-006-02.
Inspection Report# : 2010006 (pdf)
Emergency Preparedness Significance: SL-IV Sep 07, 2010 Identified By: NRC Item Type: NCV NonCited Violation Changes to EAL Technical Bases Document Decreases the Effectiveness of the Plan Without Prior NRC Approval The inspector identified a Severity Level IV NCV of 10 CFR 50.54(q) associated with 10 CFR 50.47(b)(4) because the licensee failed to obtain prior NRC approval for a change made to its emergency plan that decreased the effectiveness of the plan. Specifically, the licensee changed wording in their EAL technical basis document for EAL SU5 and CU1, RCS Leakage. The new wording eliminates leakage from the charging and letdown systems from consideration as RCS Leakage and therefore, leakage from these systems that meet the EAL thresholds would not constitute an Unusual Event declaration, using the licensees revised wording. This change was made without prior NRC approval.
The Green finding associated with this Item 05000305/2010502-02.
Inspection Report# : 2010502 (pdf)
Significance:      Sep 07, 2010 Identified By: NRC Item Type: FIN Finding Changes Made to EAL Technical Bases that Decreased the Effectiveness The inspector identified a Green finding 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. Specifically, the licensee changed wording in their EAL technical basis document for EAL SU5 and CU1, RCS Leakage. The new wording eliminates leakage from the charging and letdown systems from consideration as RCS Leakage and therefore, leakage from these systems that meet the EAL thresholds would not constitute an Unusual Event declaration, using the licensees revised wording. This change was made without prior NRC approval.
The performance deficiency was more than minor and of very low safety-significance using MC 0612 and MC 0609,
 
Appendix B, 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. Using MC 0609, Appendix B, the inspectors determined that the finding had a very low safety significance. The inspectors also determined that the finding had a cross-cutting aspect in the area of Human Performance, decision making because the licensee did not recognize that the change that was made to the EAL Technical Basis document decreased the effectiveness of the emergency plan.
(H.1.(b)) (Section 1EP4)
The associated SLIV is Item 05000305/2010502-01.
Inspection Report# : 2010502 (pdf)
Occupational Radiation Safety Significance:      Sep 30, 2010 Identified By: NRC Item Type: NCV NonCited Violation Unauthorized Entry into an HRA A finding of very low safety significance and an associated non-cited violation (NCV) of Technical Specification 6.13 was identified by the inspectors after a worker entered a high radiation area on October 15, 2009. Radiation protection did not authorize the worker to enter the area nor was the worker made knowledgeable of the dose rate level in the area. The work was temporarily assigned from the turbine building to the containment building to assist with the cleaning of containment in preparation for containment close out. The worker received a briefing from radiation protection regarding the radiological condition of containment, but was instructed not to enter any high radiation areas. The worker entered the radiological controlled area on radiation work permit 09-0202-1, which allowed access to containment but did not allow access to high radiation areas and the electronic dosimeter worn by the worker was set to alarm at 50 mrem/hour. During the course of the work activity, the worker was instructed to retrieve a piece of equipment from the basement elevation of containment. An unknown individual held the swing gate open, which also blocked the HRA posting, and the worker entered the basement elevation of containment. The worker, alerted to the higher dose rate conditions through an electronic dosimeter alarm, then exited the work area. The worker immediately reported the event to the radiation protection staff who confirmed the basement elevation of containment was a posted HRA and the dose rates were greater than 100 mrem/hour. The maximum dose rate measured by the ED was 106 mrem/hour. The corrective actions taken by the licensee included temporarily restricting the individual's further access to the radiologically controlled area and counseling of the individual by the licensee's Radiation Protection Manager.
The inspectors identified Example 6(h) of inspection manual chapter (IMC) 0612, Appendix E, as similar to the performance issue, in that, the worker was neither authorized by radiation protection to work in specific locations within containment, nor was the worker made knowledgeable of the dose rate level in the area. Therefore, in accordance with IMC 0612 and Example 6(h) of Appendix E, the inspectors determined that the performance deficiency was more than minor. Additionally, the performance deficiency impacted 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, unauthorized entry into areas without knowledge of the radiological conditions placed the worker at increased risk for unnecessary radiation exposure. The finding was assessed using the Occupational Radiation Safety significance determination process (SDP) and was determined to be of very low safety significance because the problem was not an as low as is reasonably achievable planning issue, there were no overexposures nor substantial potential for overexposures given the worker's reaction to the electronic dosimeter alarm and the dose rate ranges, and the licensees ability to assess dose was not compromised. The inspectors determined that the cause of this incident involved a cross cutting component in the human performance area for inadequate work control. Specifically, the licensee did not appropriately coordinate work activities by incorporating necessary to assure human performance (H.3(b)).
Inspection Report# : 2010004 (pdf)
 
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.
Miscellaneous Last modified : June 07, 2011
 
Kewaunee 2Q/2011 Plant Inspection Findings Initiating Events Significance:        Mar 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Misapplication Of Code Acceptance Criteria For Weld Flaws A finding of very low safety-significance and associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion IX, Control of Special Processes, was identified by the inspectors on March 3, 2011, for the licensees failure to establish a procedure that incorporated the American Society of Mechanical Engineers Code acceptance criteria for evaluation of flaws detected during ultrasonic examinations. Consequently, the licensee applied incorrect acceptance criteria to the flaws identified during ultrasonic examination of a weld on the chemical and volume control system seal water injection filter 1A housing. Licensee corrective actions included: evaluation of weld flaws to ensure they met applicable Code criteria and revision of a site procedure to incorporate appropriate Code acceptance criteria.
The finding was determined to be more than minor because the finding, if left uncorrected, would become a more significant safety concern. Absent NRC identification, the failure to provide Code acceptance criteria could have allowed components with unacceptable cracks to be returned to service. Cracks in components returned to service would place safety related piping systems at increased risk for through wall leakage and/or failure. The licensee promptly corrected this issue before components with unacceptable flaws were returned to service. The inspectors answered No to the Significance Determination Process Phase I screening question, Assuming worst case degradation, would the finding result in exceeding the Technical Specification (TS) limit for any reactor coolant system leakage or could the finding have likely affected other mitigation systems resulting in a total loss of their safety function assuming the worst case degradation? Therefore, this finding screened as having very low safety-significance (Green). This finding has a cross-cutting aspect in the area of human performance, work practices, because the licensee did not effectively implement human error prevention techniques. Specifically, the lack of procedure acceptance criteria was caused by inadequate peer checking during the licensees review and approval of the procedure for evaluation of non destructive examination data (H.4(a)).
Inspection Report# : 2011002 (pdf)
Significance:        Mar 31, 2011 Identified By: Self-Revealing Item Type: FIN Finding Partial Loss Of Offsite Power Caused By Less Than Adequate Interface And Oversight Of Switchyard Modification Work A finding of very low safety-significance was self-revealed for the failure to adequately control relay testing for switchyard breaker installations under Design Change WO KW100691871. Specifically, on March 10, 2011, Dominion Electrical Transmission technicians deviated from standard work practices to test a relay via an internal corporate server, which caused a partial loss of offsite power to the plant through the loss of the main auxiliary transformer backfeed to safety-related bus 6. Licensee corrective actions included a human performance and safety stand down for substation personnel on the day of the event, the development of a mitigating strategy that outlined expectations and implemented increased direct supervision on critical tasks, and the development of a formal memo describing expectations related to the restricted use of the server for performing remote testing of control functions.
The finding was determined to be more than minor because, if left uncorrected, the finding had the potential to lead to a more significant safety concern. Specifically, had a different breaker been inappropriately tripped, the station could have experienced a total loss of offsite power. The inspectors concluded that the finding could be evaluated using Inspection Manual Chapter 0609, Appendix M, Significance Determination Process Using Qualitative Criteria.
Specifically, the inspectors qualitatively evaluated the finding by applying the spent fuel pool questions in the Fuel Barrier column of Table 4a, Attachment 4. The inspectors answered "No" to all three questions and determined that
 
the finding was of very low safety-significance (Green). The finding has a cross-cutting aspect in the areas of human performance, work practices, because supervisory and management oversight of work activities, including contractors, was not implemented for this evolution (H.4(c)).
Inspection Report# : 2011002 (pdf)
Significance:      Dec 31, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure To Follow Red Channel Instrument Test Procedure A finding of very low safety significance and associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed when a nuclear control operator (NCO) failed to perform a procedure step, which resulted in the main feedwater regulating valve FW 7A partially closing while the reactor was at full power. Specifically, Step 6.11.2 of procedure SP-47-316A, Channel 1 (Red) Instrument Channel Test Channel Operational Test, directed the NCO to place the main feedwater regulating valve FW 7A in manual to preclude valve movement during a simulated portion of the test; however, the NCO marked the step "not applicable" and subsequently did not perform it. The licensee initiated condition reports (CRs) CR396649 and CR405809, performed an apparent cause evaluation (ACE), and initiated corrective actions (CAs) to address the issues identified in the causal evaluation.
The finding was determined to be more than minor in accordance with Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated December 24, 2009, because it was associated with the Initiating Events Cornerstone attribute of human performance and adversely affected the cornerstone objective of limiting 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 follow the procedure initiated a secondary-side plant transient. The inspectors determined the finding could be evaluated using the Significance Determination Process (SDP) in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Tables 3b and 4a for the Initiating Events Cornerstone, dated January 10, 2008. The inspectors answered "no" to the Initiating Events Cornerstone Transient Initiator question and screened the finding as having very low significance (Green). The finding has a cross-cutting aspect in the area of human performance, Work Practices, because the personnel work practices did not support human performance. Specifically, licensee personnel failed to follow procedures (H.4(b)).
Inspection Report# : 2010005 (pdf)
Mitigating Systems Significance:      Jun 30, 2011 Identified By: Self-Revealing Item Type: FIN Finding Technical Support Center Diesel Generator Output Breaker Fails To Close A finding of very low safety significance was self revealed for the failure to perform adequate preventive maintenance on latching relay VR1/B46, a relay required for closure of the technical support center (TSC) diesel generators (DG's) output breaker and automatic restoration of bus 1-46, which powers the TSC DGs cooling system. Specifically, on March 20, 2011, during a partial loss of offsite power event, the TSC DG started but failed to load onto bus 1-46.
After approximately 43 minutes of operation, the DG automatically shut down from an over-temperature condition, as designed. The licensee initiated condition report 417289 and performed apparent cause evaluation 018573. The licensees short-term corrective actions included troubleshooting the initial failure, repairing relay VR1/B46, and restoring the TSC DG to functional status. The licensees long-term corrective actions were in-progress at the completion of this inspection period.
The finding 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.
Specifically, the failure of the output breaker to close and energize bus 1-46 caused the TSC DG to overheat and
 
automatically shut down during a partial loss of offsite power. The inspectors concluded the finding could be evaluated in accordance with Inspection Manual Chapter 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Table 4a, for the Mitigating Systems Cornerstone. The inspectors answered Yes to questions 2 and 4 of the Mitigating Systems Cornerstone column and determined that the finding required a Phase 2 analysis. The Region III senior reactor analyst completed a Phase 2 analysis and determined the risk significance of the issue to be very low (Green). The finding has a cross-cutting aspect in the area of human performance, resources, because a licensee effort to review various plant components for possible inclusion in a preventive maintenance optimization project had assigned a low priority to this relay (H.2(a)).
Inspection Report# : 2011003 (pdf)
Significance:      Jun 30, 2011 Identified By: NRC Item Type: FIN Finding Inadequate Operability Determination Of A Heat Exchanger Leak On Emergency Diesel Generator A A finding of very low safety significance was identified by the inspectors for the failure to adequately assess operability of the service water system in operability determination 413, EDG A Jacket Water Expansion Tank Overflow, in accordance with site Procedure OP-AA-102-1001, Development of Technical Basis to Support Operability Determinations. At the end of the inspection period, the licensee was completing an apparent cause evaluation to determine the cause and develop corrective actions.
The finding was determined to be more than minor because the finding, if left uncorrected, had the potential to become a more significant safety concern. Specifically, the failure to perform operability evaluations on degraded safety-related systems could lead to situations where systems needed to mitigate design basis accidents were not capable of performing their required safety functions. The inspectors determined the finding could be evaluated using Inspection Manual Chapter 0609, Significance Determination Process, Attachment 0609.04, Phase 1- Initial Screening and Characterization of Findings, Table 4a, for the Mitigating Systems Cornerstone. The inspectors answered "No" to the Mitigating Systems questions and screened the finding as having very low safety significance (Green). The finding has a cross-cutting aspect in the area of human performance, work practices, because the licensee failed to communicate decisions and the bases for decisions to personnel who had a need to know the information in order to perform work safely. Specifically, the licensee failed to effectively communicate the expectation to assess operability of the service water system in the pre-job brief and peer review (H.1(c)).
Inspection Report# : 2011003 (pdf)
Significance:      Jun 30, 2011 Identified By: NRC Item Type: FIN Finding Failure To Review And Update Severe Accident Management Guidelines In Accordance With An Established Program A finding of very low safety significance was identified by the inspectors for the licensees failure to perform reviews and update the Severe Accident Management Guidelines (SAMGs) in accordance with the licensees nuclear administrative directives (NADs). Specifically, Procedure NAD 14.06 required that the engineering group review industry correspondence related to SAMGs and implement appropriate changes, and that the emergency preparedness group conduct biennial reviews of the SAMGs. The inspectors identified that neither group had performed the reviews. As a result, the SAMGs were not adequately updated. The licensee entered this issue into their corrective action program as condition reports 424681, 424855, 424865, 424866, 425092, 426999, and 427092, and was still evaluating the cause for this condition at the end of this inspection period. The licensee scheduled the revision of the SAMGs for completion by December 2011.
The finding was determined to be more than minor because, if left uncorrected, the finding had the potential to lead to a more significant safety concern. Specifically, the failure to review and update the SAMGs would have hampered the licensees response in the unlikely event of a severe accident, because the SAMGs were not current. The inspectors, in consultation with the Region III senior reactor analyst, determined that the finding could be evaluated using the Significance Determination Process in accordance with Inspection Manual Chapter 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Table 4a, for the Mitigating Systems Cornerstone. The inspectors answered "No" to the Mitigating Systems questions and screened the
 
finding as having very low safety significance (Green). The finding has a cross-cutting aspect in the area of problem identification and resolution, corrective action program, because the licensee failed to take appropriate corrective actions to address safety issues and adverse trends in a timely manner, commensurate with their safety significance and complexity. Specifically, the licensee identified in an apparent cause evaluation initiated in April 2010 that the emergency preparedness organization had not performed the required reviews and updates of emergency preparedness procedures, and the SAMGs were identified in the licensees extent of condition. However, the inspectors identified that the corrective actions issued for this extent of condition did not address the SAMGs and, therefore, no corrective actions were taken (P.1(d)).
Inspection Report# : 2011003 (pdf)
Significance:        Mar 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Work Instructions Results In Potential Orange Path A finding of very low safety-significance and associated non-cited violation (NCV) of Technical Specification 5.4.1, Procedures, was identified by the inspectors for the failure to implement procedures for shutdown operations involving shutdown operations safety assessments. Specifically, OU KW 201, Shutdown Safety Assessment Checklist, step 3.3.1, stated, in part, that a shutdown safety assessment was required to be completed in accordance with the procedure for core cooling; however, the inspectors noted that the February 28, 2011, 6:00 p.m. analysis credited the safety injection system feed and bleed as an available alternate decay heat removal system when the system was not available as described in Section 5.3.2, Available/Availability, for work scheduled at that time on the emergency core cooling system (ECCS) sump. The licensee initiated condition report CR415539, and at the end of the inspection period, the licensee was performing a causal evaluation to determine the causes of the event and develop corrective actions. On February 28, as a remedial corrective action prior to the start of work, additional steps to the work instructions were added to ensure the equipment would meet the intended function, operators were designated to perform the local manual operations and a pre job brief was conducted that provided training for using the equipment in the given situation.
The finding was determined to be more than minor because the finding was associated with the Mitigating Systems Cornerstone attribute of human error (pre event) and adversely 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). Specifically, the availability of the ECCS sump was integral to ensuring that the plant was not in an orange risk path for the evolutions completed on February 28. The inspectors screened the finding as of very low safety-significance (Green) because the finding did not degrade the licensees ability to establish an alternate core cooling path if decay heat removal could not be re established and, therefore, did not require a Significance Determination Process phase 2 or phase 3 analysis. The finding has a cross-cutting aspect in the areas of human performance, work control, because the licensee failed to plan the work activities by incorporating the need for planned contingencies and compensatory actions to ensure the ECCS sump was available to ensure an orange risk path for core cooling was not entered (H.3(a)).
Inspection Report# : 2011002 (pdf)
Significance:        Mar 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Unintended Voiding Of The Reactor Vessel Closure Head A finding of very low safety-significance and associated non-cited violation (NCV) of Technical Specification 5.4.1, Procedures, was identified by the inspectors for the failure to establish, implement, and maintain procedures for shutdown operations involving the draining of reactor coolant system (RCS) inventory. Specifically, on March 21, 2011, during a pressurizer draindown evolution, licensed operators unknowingly created a gas void in the reactor vessel closure head (RVCH) that displaced water to a level near the RVCH flange. Subsequent evaluation determined that the procedure for draining the RCS did not contain adequate guidance to ensure that an unacceptable void in the RVCH was not present prior to or formed during operations draindown activities. The licensee subsequently entered the issue into its corrective action program as CR418537 and performed a remedial corrective action of removing the gas void that accumulated in the RVCH. At the end of the inspection period, the licensee was performing an apparent cause evaluation to determine the causes of the event and develop additional corrective actions.
 
The finding was determined to be more than minor because the finding was associated with the Mitigating Systems Cornerstone attribute of operating procedure quality and adversely 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). Specifically, the formation of the gas void in the RVCH displaced RCS inventory and could have challenged the ability to remove decay heat in the event of a loss of shutdown cooling. The Region III senior reactor analyst determined that this issue is best characterized as a finding of very low safety-significance (Green). The finding has a cross-cutting aspect in the areas of human performance, work practices, because operations personnel did not follow or implement the guidance contained in plant procedures. Specifically, procedure OP KW AOP RC 002 prescribed actions to take if a gas void formed in the RVCH that resulted in RVLIS level readings less than 88 percent, which had occurred several hours prior to the start of a pressurizer draining evolution (H.4(b)).
Inspection Report# : 2011002 (pdf)
Significance:        Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Incorrect Containment Fan Coil Unit Acceptance Criteria A finding of very low safety significance and associated non cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors for the failure to correctly translate the applicable regulatory requirements and the design basis into procedures and instructions. Specifically, the licensee failed to adequately translate the containment fan coil unit (CFCU) service water flow acceptance criteria from the current design basis calculations into the CFCU performance monitoring procedures, which resulted in the incorrect acceptance criteria in plant test procedures. The licensee took immediate corrective actions to correct the acceptance criteria in the test procedures and to perform an operability determination on CFCU C, the only one of the four CFCUs that showed a recent decrease in flow. At the end of the inspection period, the licensee was completing an apparent cause evaluation and developing additional long term corrective actions.
The finding was determined to be more than minor in accordance with Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated December 24, 2009, because it was associated with the Mitigating Systems Cornerstone attribute of procedure quality and adversely affected the cornerstone objective of ensuring the reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to ensure that procedure PMP 18 13, "Containment Fan Coil Unit Performance Monitoring (AQ-1)," contained the correct acceptance criteria for testing the CFCUs. The inspectors determined the finding could be evaluated using the Significance Determination Process (SDP) in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Tables 3b and 4a for the Mitigating Systems Cornerstone, dated January 10, 2008.
The inspectors answered "no" to the Mitigating Systems questions and screened the finding as having very low significance (Green). This finding has a cross cutting aspect in the area of human performance, Resources, because the licensee did not maintain complete, accurate, and up to date procedures. Specifically, the correct acceptance criteria for testing the CFCUs from the design basis calculations were not specified in the CFCU testing procedure (H.2(c)).
Inspection Report# : 2010005 (pdf)
Significance:        Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Unacceptable Preconditioning of Safety-Related Pressure Switches A finding of very low significance and associated non cited violation (NCV) of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the failure to develop and implement an adequate surveillance test procedure to accurately assess the as found trip setpoint for the pressure switches associated with the turbine building service water isolation function and various other safety related functions. The licensee initiated condition report CR401813, performed an apparent cause evaluation, and initiated corrective actions to address the issues identified in the casual evaluation.
The finding was determined to be more than minor in accordance with Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated December 24, 2009, because it was
 
associated with the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences and affected the cornerstone attribute of Equipment Performance. The inspectors determined the finding could be evaluated using the Significance Determination Process (SDP) in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Tables 3b and 4a for the Mitigating Systems Cornerstone, dated January 10, 2008. The inspectors answered no to the Mitigating Systems questions and screened the finding as having very low significance (Green). This finding has a cross cutting aspect in the area of problem identification and resolution, Operating Experience, because the licensee did not evaluate and communicate external operating experience to internal stakeholders in a timely manner (P.2(a)).
Inspection Report# : 2010005 (pdf)
Significance:        Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Inappropriate Isolation of the Safety Injection Pump Minimum Flow Recirculation Lines A finding of very low safety significance and associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for multiple inadequate procedures, which directed closing the common train safety injection minimum flow recirculation line valves, an activity affecting quality. Specifically, station procedures directed operators to close the safety injection pump minimum flow recirculation valves in order to complete valve timing tests, and to engage an interlock that allowed closure of the containment sump recirculation valves. However, the procedures and licensed operators failed to recognize that closure of either minimum flow recirculation valve affected the operability and availability of both safety injection pumps for certain design basis accidents because the minimum flow recirculation path was isolated.
The licensee subsequently entered the issue into its corrective action program as CR393930. The licensee corrected the procedure inadequacies and completed a root cause evaluation that recommended several corrective actions to prevent recurrence.
The finding was determined to be more than minor in accordance with Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated December 24, 2009, because it was associated with the Mitigating System Cornerstone attribute of procedure quality, 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 ensure that procedures implemented during power operations ensured the operability of both trains of safety injection. The inspectors determined the finding could be evaluated using the Significance Determination Process (SDP) in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Tables 3b and 4a for the Mitigating Systems Cornerstone, dated January 10, 2008. The inspectors answered "yes" to the Mitigating Systems question that confirmed the finding represented a loss of system safety function. The Region III Senior Reactor Analyst (SRA) performed an SDP Phase 2 analysis and a Phase 3 analysis. The Phase 3 analysis determined that the resultant delta core damage frequency (CDF) was less than 1E 6 and delta large early release frequency (LERF) was less than 1E 7, which represented a Green finding. The dominant scenario involved a small break loss of coolant accident with operator failure to perform a rapid cool down. The finding has a cross cutting aspect in the area of human performance, Decision Making, because although the licensee procedures cautioned that starting a safety injection pump following the closure of a minimum flow recirculation valve would result in damage to the pump, the licensee staff failed to use conservative decision making to question the adequacy of the prescribed procedure actions (H.1(b)).
Inspection Report# : 2010005 (pdf)
Significance:        Sep 30, 2010 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Emergency Operating Procedure A finding of very low safety significance and associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the licensees failure to have an adequate emergency operating procedure for an activity affecting quality. Specifically, emergency operating
 
procedure E 2, Faulted Steam Generator Isolation, did not prescribe actions to manually close the steam supplies to the turbine-driven auxiliary feedwater pump in the event the control room switches failed to operate. The licensee initiated condition report (CR) CR391458 and took immediate corrective actions to correct the deficient procedure and informed the licensed operators.
The finding was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of procedure quality and adversely affected the cornerstone objective of ensuring the reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to ensure that Emergency Operating Procedure E 2 contained all the required actions to ensure successful isolation of a faulted steam generator. The inspectors determined the finding could be evaluated using the significance determination process (SDP) in accordance with Inspection Manual Chapter (IMC) 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Tables 3b and 4a, for the Mitigating Systems Cornerstone. The inspectors answered "no" to the Mitigating Systems questions and screened the finding as having very low significance (Green). The inspectors determined that this finding did not reflect present performance since the procedure error was introduced greater than three years ago; therefore, there was no cross cutting aspect associated with this finding.
Inspection Report# : 2010004 (pdf)
Significance:        Sep 30, 2010 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Barrier Control Procedures Result In Exposed Service Water Pumps A finding of very low safety significance and associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the failure to have adequate procedures to address the removal of the screenhouse traveling water screen covers, an activity affecting quality. Consequently, the covers were removed and safety related equipment was exposed to the environment without adequate planning of mitigation actions in the event of inclement weather. The licensee initiated condition reports (CR) CR394670, CR395541, and CR395717 to document the issue. At the end of the inspection period, the licensee was performing a causal evaluation and developing corrective actions to address the issue.
The finding was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of protection against external factors and adversely 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 inspectors determined the finding could be evaluated using the significance determination process (SDP) in accordance with Inspection Manual Chapter (IMC) 0609, Significance Determination Process, 609.04, Phase 1 Initial Screening and Characterization of Findings, Tables 3b, 4a, and 4b for the Mitigating Systems Cornerstone. The inspectors determined that the screenhouse covers were designed to prevent tornado missiles from damaging the safety related equipment housed inside the screenhouse and that two trains of the service water system would be degraded; therefore, the inspectors answered yes to the Table 4b seismic, flooding, and severe weather screening criteria questions 1 and 2. The inspectors contacted the RIII senior reactor analyst who determined, using NUREG/CR 4461, Tornado Climatology of the Contiguous United States, and the number of days the covers were removed that the performance deficiency risk was of very low safety significance (Green). The finding has a cross cutting aspect in the area of human performance, Decision Making, because the licensee failed to make safety significant or risk significant decisions using a systematic process to ensure safety is maintained.
Specifically, the licensee applied an incorrect evaluation to a situation that resulted in the multiple trains of service water pumps being unprotected from tornado missiles (H.1(a)).
Inspection Report# : 2010004 (pdf)
Significance: SL-IV Sep 30, 2010 Identified By: NRC Item Type: NCV NonCited Violation Replacement of Automatic Action With An Operator Manual Action Without Prior NRC Approval A Severity Level IV non-cited violation (NCV) of 10 CFR 50.59(d)(1), Changes, Tests, and Experiments, was identified by the inspectors for the failure to document an evaluation that provided a basis for the determination that the changes implemented in DCR 3163 and Emergency Operating Procedure ES 1.3, Transfer to Sump Recirculation, in 2001 did not require a license amendment. Specifically, the licensee failed to provide an evaluation
 
that adequately documented why replacing the automatic opening of the service water (SW) valves SW 1300A and SW 1300B upon a safety injection signal (to support the service water safety function of loss of coolant accident (LOCA) recirculation operation) with a manual action to open the valves in Emergency Operating Procedure ES 1.3, did not present more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component (SSC) important to safety previously evaluated in the updated safety analysis report. The licensee initiated CR389330 and, at the end of the inspection period, planned to submit a license amendment request to the NRC for this design change.
The violation was determined to be more than minor because the inspectors could not reasonably determine that the changes would not have ultimately required prior NRC approval. Violations of 10 CFR 50.59 are dispositioned using the traditional enforcement process instead of the significance determination process (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 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, Tables 3b and 4a, for the Mitigating Systems Cornerstone. The inspectors answered "yes" to question 1 of the Mitigating Systems Cornerstone column of the Phase 1 worksheet because the inspectors concluded that this was a design basis deficiency confirmed not to result in the loss of operability. Based upon this Phase 1 screening, the inspectors concluded that the issue was of very low safety significance (Green). In accordance with Section 6.1.d.2 of the NRC Enforcement Policy this violation is categorized as Severity Level IV because the resulting changes were evaluated by the SDP as having very low safety significance. The inspectors determined that this finding did not reflect present performance since the error was introduced in a design change that was greater than three years old; therefore, there was no cross cutting aspect associated with this finding.
See related performance deficiency item 2010-004-04.
Inspection Report# : 2010004 (pdf)
Significance:      Sep 30, 2010 Identified By: NRC Item Type: FIN Finding Performance Deficiency Associated With SLIV NCV - Replace of Automatic Actions With an Operator Manual Action Without Prior NRC Approval A Severity Level IV NCV of 10 CFR 50.59(d)(1), Changes, Tests, and Experiments, was identified by the inspectors for the failure to document an evaluation that provided a basis for the determination that the changes implemented in DCR 3163 and Emergency Operating Procedure ES 1.3, Transfer to Sump Recirculation, in 2001 did not require a license amendment. Specifically, the licensee failed to provide an evaluation that adequately documented why replacing the automatic opening of the service water (SW) valves SW 1300A and SW 1300B upon a safety injection signal (to support the service water safety function of loss of coolant accident (LOCA) recirculation operation) with a manual action to open the valves in Emergency Operating Procedure ES 1.3, did not present more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component (SSC) important to safety previously evaluated in the updated safety analysis report. The licensee initiated CR389330 and, at the end of the inspection period, planned to submit a license amendment request to the NRC for this design change.
The violation was determined to be more than minor because the inspectors could not reasonably determine that the changes would not have ultimately required prior NRC approval. 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 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, Tables 3b and 4a, for the Mitigating Systems Cornerstone. The inspectors answered "yes" to question 1 of the Mitigating Systems Cornerstone column of the Phase 1 worksheet because the inspectors concluded that this was a design basis deficiency confirmed not to result in the loss of operability. Based upon this Phase 1 screening, the inspectors concluded that the issue was of very low safety significance (Green). In accordance with Section 6.1.d.2 of the NRC Enforcement Policy this violation is categorized as Severity Level IV because the resulting changes were evaluated by the SDP as having very low safety significance.
The inspectors determined that this finding did not reflect present performance since the error was introduced in a
 
design change that was greater than three years old; therefore, there was no cross cutting aspect associated with this finding.
See related SL IV NCV item 2010-004-03.
Inspection Report# : 2010004 (pdf)
Barrier Integrity Significance:        Jun 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failed Standoffs Result In An Inoperable Train of Shield Building Ventilation A finding of very low safety significance and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by inspectors for the failure to have and follow adequate procedures for evaluation and installation of components in shield building ventilation (SBV) train A. Specifically, the licensee failed to have adequate procedures to direct the completion of a subcomponent classification evaluation (SCE) and prevent non safety-related parts from being installed in safety-related applications; have torque specifications for the standoffs (spacers for circuit cards) in the work instructions; and properly accomplish the SCE procedure when evaluating the standoffs. The licensees initial short-term corrective actions removed the installed standoffs from both trains. The licensee also performed an extent of condition looking at previously completed item equivalency evaluations to determine if an SCE was needed or missing for newly installed components.
The finding was determined to be more than minor because the finding was associated with the Barrier Integrity Cornerstone attribute of procedure quality, 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 licensee failed to have and follow adequate procedures which led to the failure of SBV train A. The inspectors determined that this was a type B containment finding since it was related to a degraded condition that had potential important implications for the integrity of the containment, without affecting the likelihood of core damage.
The inspector evaluated the finding using the significance determination process (SDP) in accordance with Inspection Manual Chapter 0609, Appendix H, Containment Integrity SDP, Table 4.1, and determined that the finding did not relate to a containment structure, system, and component, nor containment status that had an impact on large early release frequency. Because of this, the issue screened as Green, using the flowchart in Figure 4.1. The finding has a cross-cutting aspect in the area of problem identification and resolution, corrective action program, because the licensee failed to thoroughly evaluate problems such that the resolutions address causes and extent of conditions, as necessary. This includes properly classifying, prioritizing, and evaluating for operability and reportability conditions adverse to quality. This also includes, for significant problems, conducting effectiveness reviews of corrective actions to ensure that the problems are resolved. Specifically, the licensee failed to properly evaluate and identify the cause of the SBV train A failure and produce a resolution that addressed the cause (P.1(c)).
Inspection Report# : 2011003 (pdf)
Significance:        Sep 03, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Correct the Classification of a Containment Isolation Valve A finding of very low safety significance and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified by the inspectors for the failure to correct a condition adverse to quality.
Specifically, the licensee failed to provide their licensed operators with correct procedures and instructions for determining which valves were containment isolation valves. The condition was previously identified on August 12, 2009, when the inspectors found MS 100A, the steam supply to the turbine driven auxiliary feedwater pump, open without the capability to be remotely closed from the control room and without a technical specification entry for the containment isolation function. The licensee entered the issue, during the current inspection, into their corrective action program and took short-term corrective actions of placing a standing order in the control room directing operators to enter the appropriate containment isolation technical specifications for the valves in question.
 
The finding was determined to be more than minor, because, if left uncorrected, has the potential to lead to a more significant safety concern. The inspectors concluded this finding was associated with the Barrier Integrity Cornerstone. 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. The inspectors answered no to the Barrier Integrity Cornerstone questions and screened the finding as having very low safety significance (Green). This finding has a cross-cutting aspect in the area of human performance within the resources component because the licensee did not maintain complete, accurate and up-to-date design documentation (H.2(c)).
Inspection Report# : 2010006 (pdf)
Significance: SL-IV Sep 03, 2010 Identified By: NRC Item Type: NCV NonCited Violation Failure to Update the Updated Safety Analysis Report to Include Containment Penetration Leakage Testing Information The inspectors identified a Severity Level IV, non-cited violation of 10 CFR 50.71(e), Maintenance of Records, Making of Reports, having very low safety significance. The inspectors found that the licensee failed to update the Updated Safety Analysis Report (USAR) to describe for each containment penetration, the penetration category, the type of leakage test required, and the applicable leakage test method. The licensee entered this into their corrective action program. The inspectors found the violation to be more than minor in accordance with the NRC Enforcement Policy, Section 6.1.d, Example 3, in that the failure to update the Final Safety Analysis Report (FSAR) would not have a material impact on safety or licensed activities. This issue was determined to be a Severity Level IV violation since it was similar to a Severity Level IV violation example in the NRC Enforcement Policy. Additionally, in accordance with the Enforcement Policy, this violation is categorized as Severity Level IV because the resulting changes were evaluated by the SDP as having very low safety significance (Green).
Violations of 10 CFR 50.71 are dispositioned using the traditional enforcement process instead of the significance determination process (SDP) because they are considered to be violations that potentially impede or impact the regulatory process. The underlying finding is evaluated under the SDP to determine the significance of the violation.
In this case, the finding was determined to be more than minor because, if left uncorrected, it had the potential to lead to a more significant safety concern.
The SDP portion of this issue is tracked as item 2010-006-03.
Inspection Report# : 2010006 (pdf)
Significance:        Sep 03, 2010 Identified By: NRC Item Type: FIN Finding Failure to Update the Updated Safety Analysis Report to Include Containment Penetration Leakage Testing Information The inspectors identified a finding associated with a traditional enforcement Severity Level IV, non-cited violation of 10 CFR 50.71(e), Maintenance of Records, Making of Reports, having very low safety significance. The resulting changes were evaluated by the SDP as having very low safety significance (Green).
The underlying finding was evaluated under the SDP to determine the significance of the violation. In this case, the finding was determined to be more than minor because, if left uncorrected, it had the potential to lead to a more significant safety concern. 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. The inspectors answered no to the Barrier Integrity Cornerstone questions and screened the finding as having very low safety significance (Green). The inspectors did not identify a cross-cutting aspect associated with the finding because the finding was not representative of current performance.
The Traditional Enforcement portion of this issue is tracked as item 2010-006-02.
Inspection Report# : 2010006 (pdf)
 
Emergency Preparedness Significance: SL-IV Sep 07, 2010 Identified By: NRC Item Type: NCV NonCited Violation Changes to Emergency Action Level (EAL) Technical Bases Document Decreases the Effectiveness of the Plan Without Prior NRC Approval The inspector identified a Severity Level IV NCV of 10 CFR 50.54(q) associated with 10 CFR 50.47(b)(4) because the licensee failed to obtain prior NRC approval for a change made to its emergency plan that decreased the effectiveness of the plan. Specifically, the licensee changed wording in their EAL technical basis document for EAL SU5 and CU1, RCS Leakage. The new wording eliminates leakage from the charging and letdown systems from consideration as RCS Leakage and therefore, leakage from these systems that meet the EAL thresholds would not constitute an Unusual Event declaration, using the licensees revised wording. This change was made without prior NRC approval.
The Green finding associated with this Item 05000305/2010502-02.
Inspection Report# : 2010502 (pdf)
Significance:      Sep 07, 2010 Identified By: NRC Item Type: FIN Finding Changes Made to Emergency Action Level (EAL) Technical Bases that Decreased the Effectiveness The inspector identified a Green finding 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. Specifically, the licensee changed wording in their EAL technical basis document for EAL SU5 and CU1, RCS Leakage. The new wording eliminates leakage from the charging and letdown systems from consideration as RCS Leakage and therefore, leakage from these systems that meet the EAL thresholds would not constitute an Unusual Event declaration, using the licensees revised wording. This change was made without prior NRC approval.
The performance deficiency was more than minor and of very low safety-significance using MC 0612 and MC 0609, Appendix B, 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. Using MC 0609, Appendix B, the inspectors determined that the finding had a very low safety significance. The inspectors also determined that the finding had a cross-cutting aspect in the area of Human Performance, decision making because the licensee did not recognize that the change that was made to the EAL Technical Basis document decreased the effectiveness of the emergency plan.
(H.1.(b)) (Section 1EP4)
The associated SLIV is Item 05000305/2010502-01.
Inspection Report# : 2010502 (pdf)
Occupational Radiation Safety Significance:      Sep 30, 2010 Identified By: NRC Item Type: NCV NonCited Violation
 
Unauthorized Entry into a High Radiation Area (HRA)
A finding of very low safety significance and an associated non-cited violation (NCV) of Technical Specification 6.13 was identified by the inspectors after a worker entered a high radiation area on October 15, 2009. Radiation protection did not authorize the worker to enter the area nor was the worker made knowledgeable of the dose rate level in the area. The work was temporarily assigned from the turbine building to the containment building to assist with the cleaning of containment in preparation for containment close out. The worker received a briefing from radiation protection regarding the radiological condition of containment, but was instructed not to enter any high radiation areas. The worker entered the radiological controlled area on radiation work permit 09-0202-1, which allowed access to containment but did not allow access to high radiation areas and the electronic dosimeter worn by the worker was set to alarm at 50 mrem/hour. During the course of the work activity, the worker was instructed to retrieve a piece of equipment from the basement elevation of containment. An unknown individual held the swing gate open, which also blocked the HRA posting, and the worker entered the basement elevation of containment. The worker, alerted to the higher dose rate conditions through an electronic dosimeter alarm, then exited the work area. The worker immediately reported the event to the radiation protection staff who confirmed the basement elevation of containment was a posted HRA and the dose rates were greater than 100 mrem/hour. The maximum dose rate measured by the ED was 106 mrem/hour. The corrective actions taken by the licensee included temporarily restricting the individual's further access to the radiologically controlled area and counseling of the individual by the licensee's Radiation Protection Manager.
The inspectors identified Example 6(h) of inspection manual chapter (IMC) 0612, Appendix E, as similar to the performance issue, in that, the worker was neither authorized by radiation protection to work in specific locations within containment, nor was the worker made knowledgeable of the dose rate level in the area. Therefore, in accordance with IMC 0612 and Example 6(h) of Appendix E, the inspectors determined that the performance deficiency was more than minor. Additionally, the performance deficiency impacted 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, unauthorized entry into areas without knowledge of the radiological conditions placed the worker at increased risk for unnecessary radiation exposure. The finding was assessed using the Occupational Radiation Safety significance determination process (SDP) and was determined to be of very low safety significance because the problem was not an as low as is reasonably achievable planning issue, there were no overexposures nor substantial potential for overexposures given the worker's reaction to the electronic dosimeter alarm and the dose rate ranges, and the licensees ability to assess dose was not compromised. The inspectors determined that the cause of this incident involved a cross cutting component in the human performance area for inadequate work control. Specifically, the licensee did not appropriately coordinate work activities by incorporating necessary to assure human performance (H.3(b)).
Inspection Report# : 2010004 (pdf)
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.
Miscellaneous Significance: SL-IV Jun 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure To Submit LER Per 10 CFR 50.73 A Severity Level IV non-cited violation of 10 CFR Part 50.73(a)(2)(i)(B) and 50.73(a)(2)(v)(C) was identified by the
 
inspectors for the failure of the licensee to report an event or condition that was prohibited by Technical Specifications, and an event or condition that could have prevented the fulfillment of a safety function, that is relied upon to control the release of radioactive material. Specifically, the licensee failed to report the inoperability of shield building ventilation train A from December 3, 2010, through January 26, 2011, a condition prohibited by Technical Specification 3.6.c.1, which allowed a single train outage time of seven days. Additionally, shield building ventilation train B was inoperable on multiple occasions during the same time period, requiring the licensee to also report an event or condition that could have prevented the fulfillment of a safety function, which is relied upon to control the release of radioactive material. At the end of the inspection period, the licensee was completing an apparent cause evaluation to determine the cause and develop corrective actions.
Because violations of 10 CFR 50.73 are considered to be violations that potentially impact the regulatory process, they are dispositioned using the traditional enforcement process instead of the Reactor Oversight Process Significance Determination Process. A cross-cutting aspect was not assigned to this violation. Per the NRC Enforcement Policy, Section 6.0, Violation Examples, a failure to submit a required licensee event report is categorized as a Severity Level IV violation.
Inspection Report# : 2011003 (pdf)
Last modified : October 14, 2011
 
Kewaunee 3Q/2011 Plant Inspection Findings Initiating Events Significance:        Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Repetitive Molded Case Circuit Breaker Failures A finding of very low significance and associated non-cited violation of Title 10 of the Code of Federal Regulations (CFR) 50.65(a)(3) was identified by the inspectors for the failure to incorporate industry operating experience into preventive maintenance activities when practical to do so. Specifically, the failure to incorporate the industry operating experience resulted in multiple molded case circuit breaker (MCCB) failures that could have been prevented by implementing an MCCB cycling program. The need to cycle MCCBs was identified in industry operating experience as well as the vendor's instructions for the breakers. The licensee was performing an apparent cause evaluation which was still in progress at the conclusion of the inspection period. Initial corrective actions included scheduling the MCCBs for the breaker cycling maintenance activity.
This finding was determined to be of greater than minor significance because it was associated with the Protection Against External Factors attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events, such as fire, that challenge critical safety functions during shutdown as well as power operations. Specifically, the lack of a cycling program for safety related MCCBs resulted in breakers remaining in the on position after an overcurrent condition. The inspectors determined the finding had very low safety significance (Green) because the breakers and associated cabling did not significantly affect safe shutdown defense in depth strategies and the finding did not involve a design or qualification deficiency, did not represent a loss of system safety function, did not represent a loss of Technical Specification equipment for greater than its allowed outage time, and did not affect risk significant equipment per 10 CFR 50.65. This finding has a cross-cutting aspect in the area of human performance, work control, because the licensee did not emphasize the need for work groups to communicate, coordinate, and cooperate with each other during activities in which interdepartmental coordination is necessary to assure plant and human performance.
Inspection Report# : 2011004 (pdf)
Significance:        Mar 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Misapplication Of Code Acceptance Criteria For Weld Flaws A finding of very low safety-significance and associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion IX, Control of Special Processes, was identified by the inspectors on March 3, 2011, for the licensees failure to establish a procedure that incorporated the American Society of Mechanical Engineers Code acceptance criteria for evaluation of flaws detected during ultrasonic examinations. Consequently, the licensee applied incorrect acceptance criteria to the flaws identified during ultrasonic examination of a weld on the chemical and volume control system seal water injection filter 1A housing. Licensee corrective actions included: evaluation of weld flaws to ensure they met applicable Code criteria and revision of a site procedure to incorporate appropriate Code acceptance criteria.
The finding was determined to be more than minor because the finding, if left uncorrected, would become a more significant safety concern. Absent NRC identification, the failure to provide Code acceptance criteria could have allowed components with unacceptable cracks to be returned to service. Cracks in components returned to service would place safety related piping systems at increased risk for through wall leakage and/or failure. The licensee promptly corrected this issue before components with unacceptable flaws were returned to service. The inspectors answered No to the Significance Determination Process Phase I screening question, Assuming worst case degradation, would the finding result in exceeding the Technical Specification (TS) limit for any reactor coolant system leakage or could the finding have likely affected other mitigation systems resulting in a total loss of their
 
safety function assuming the worst case degradation? Therefore, this finding screened as having very low safety-significance (Green). This finding has a cross-cutting aspect in the area of human performance, work practices, because the licensee did not effectively implement human error prevention techniques. Specifically, the lack of procedure acceptance criteria was caused by inadequate peer checking during the licensees review and approval of the procedure for evaluation of non destructive examination data (H.4(a)).
Inspection Report# : 2011002 (pdf)
Significance:      Mar 31, 2011 Identified By: Self-Revealing Item Type: FIN Finding Partial Loss Of Offsite Power Caused By Less Than Adequate Interface And Oversight Of Switchyard Modification Work A finding of very low safety-significance was self-revealed for the failure to adequately control relay testing for switchyard breaker installations under Design Change WO KW100691871. Specifically, on March 10, 2011, Dominion Electrical Transmission technicians deviated from standard work practices to test a relay via an internal corporate server, which caused a partial loss of offsite power to the plant through the loss of the main auxiliary transformer backfeed to safety-related bus 6. Licensee corrective actions included a human performance and safety stand down for substation personnel on the day of the event, the development of a mitigating strategy that outlined expectations and implemented increased direct supervision on critical tasks, and the development of a formal memo describing expectations related to the restricted use of the server for performing remote testing of control functions.
The finding was determined to be more than minor because, if left uncorrected, the finding had the potential to lead to a more significant safety concern. Specifically, had a different breaker been inappropriately tripped, the station could have experienced a total loss of offsite power. The inspectors concluded that the finding could be evaluated using Inspection Manual Chapter 0609, Appendix M, Significance Determination Process Using Qualitative Criteria.
Specifically, the inspectors qualitatively evaluated the finding by applying the spent fuel pool questions in the Fuel Barrier column of Table 4a, Attachment 4. The inspectors answered "No" to all three questions and determined that the finding was of very low safety-significance (Green). The finding has a cross-cutting aspect in the areas of human performance, work practices, because supervisory and management oversight of work activities, including contractors, was not implemented for this evolution (H.4(c)).
Inspection Report# : 2011002 (pdf)
Significance:      Dec 31, 2010 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure To Follow Red Channel Instrument Test Procedure A finding of very low safety significance and associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed when a nuclear control operator (NCO) failed to perform a procedure step, which resulted in the main feedwater regulating valve FW 7A partially closing while the reactor was at full power. Specifically, Step 6.11.2 of procedure SP-47-316A, Channel 1 (Red) Instrument Channel Test Channel Operational Test, directed the NCO to place the main feedwater regulating valve FW 7A in manual to preclude valve movement during a simulated portion of the test; however, the NCO marked the step "not applicable" and subsequently did not perform it. The licensee initiated condition reports (CRs) CR396649 and CR405809, performed an apparent cause evaluation (ACE), and initiated corrective actions (CAs) to address the issues identified in the causal evaluation.
The finding was determined to be more than minor in accordance with Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated December 24, 2009, because it was associated with the Initiating Events Cornerstone attribute of human performance and adversely affected the cornerstone objective of limiting 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 follow the procedure initiated a secondary-side plant transient. The inspectors determined the finding could be evaluated using the Significance Determination Process (SDP) in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Tables 3b and 4a for the Initiating Events Cornerstone, dated January 10, 2008. The inspectors answered "no" to the Initiating Events Cornerstone Transient
 
Initiator question and screened the finding as having very low significance (Green). The finding has a cross-cutting aspect in the area of human performance, Work Practices, because the personnel work practices did not support human performance. Specifically, licensee personnel failed to follow procedures (H.4(b)).
Inspection Report# : 2010005 (pdf)
Mitigating Systems Significance:        Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure To Maintain Fire Barrier And Automatic Fire Suppression A finding of very low safety significance and associated non-cited violation of license condition 2.C(3) of the Kewaunee Power Station Renewed Operating License was identified by inspectors for the failure to have a self-closing fire door that closed and latched each time it was open. License condition 2.C(3) requires, in part, that the licensee implement and maintain, in effect, all provisions of the approved fire protection program as described in the licensees fire plan. Appendix B of the Kewaunee Power Station Fire Protection Program Plan lists the 1975 edition of NFPA 80 [National Fire Protection Association], Fire Doors and Windows, as an applicable NFPA code. NFPA 80 states, in part, that a self closing door shall be equipped with a closing device to cause the door to close and latch each time it is opened. The licensee entered the issue into its corrective action program and adjusted the door closing device to ensure the door properly closed when the train A screenhouse ventilation fan was operating.
The inspectors determined that the failure of the door to close and latch was contrary to the requirements of NFPA 80 and was a performance deficiency. The finding was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of protection against external factors (Fire) and adversely 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). The inspectors screened the finding as having very low safety significance (Green). The 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 that the resolutions address causes and extent of conditions. This includes properly classifying, prioritizing, and evaluating for operability and reportability conditions adverse to quality.
Inspection Report# : 2011004 (pdf)
Significance:        Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Misapplication Of Technical Specification 3.1.6 Applicability Note A finding of very low safety significance and associated non-cited violation of Technical Specification (TS) 3.1.6, Control Bank Insertion Limits, was identified by the inspectors for the failure to comply with TS action condition 3.1.6.A due to incorrect use of the applicability note. Specifically, on August 30, 2011, during the performance of SP-49-075, Control Rod Exercise, operators received a rod control urgent failure while inserting control bank A group 1 control rods. The test was suspended for troubleshooting for approximately 20 hours with control bank A group 2 control rods, inserted one step below the control rod insertion limit in violation of TS 3.1.6.A action condition. The inspectors concluded that, once the test was suspended for troubleshooting activities, use of the applicability note was not appropriate; therefore, the operators should have complied with the TS 3.1.6.A action condition for control bank A group 2 control rods at that time. On August 31, operators withdrew control bank A group 2 rods one step, which restored the rods to within the limit specified in the core operating limits report. At the end of this inspection period, the licensee was still performing an apparent cause evaluation to determine the causes of the event and to develop corrective actions.
The finding was determined to be more than minor because the finding adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent desirable consequences. Specifically, the human performance attributes of the licensees failure to recognize the misapplication of the applicability note of the TS affected the capability of systems that respond to initiating events. The inspectors
 
screened the finding as having very low safety significance (Green) because an actual loss of safety function did not occur. The finding has a cross-cutting aspect in the area of human performance, decision making, because the licensee failed to use conservative assumptions and adopt a requirement to demonstrate that the proposed action is safe in order to proceed rather than a requirement to demonstrate that it is unsafe in order to disapprove the action.
Inspection Report# : 2011004 (pdf)
Significance:        Sep 30, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Incorrect Transformer Load Tap Changer Setting Causes Inoperable Offsite Power A finding of very low safety significance and associated non-cited violation of Technical Specification 3.8.1 was self revealed for the failure to maintain a switchyard transformer load tap changer (LTC) at the appropriate setting for the predicted post trip voltage of offsite power. The incorrect setting resulted in the inoperability of the Reserve Auxiliary Transformer (RAT) offsite power source. The licensees corrective actions included restoring the RAT supply transformer (RST) LTC to an appropriate setting, creating a short term standing order to prevent operation of the RST LTC outside settings that were supported by the existing interface agreement with the transmission system operator.
The licensee performed an apparent cause evaluation, a root cause analysis and also, as a long-term corrective action, modified procedure OP-KW-NOP SUB 003 to prevent operation of the RST LTC outside settings that were supported by the existing interface agreement with the transmission system operator.
The finding was determined to be more than minor because the finding was associated with the Mitigating Systems Cornerstone attribute of configuration control and adversely 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). The inspectors screened the finding as having very low safety significance (Green). This finding has a cross-cutting aspect in the area of human performance, resources, because the licensee did not maintain complete, accurate, and up to date procedures for the use of the RST LTC following its installation during the spring 2011 outage.
Inspection Report# : 2011004 (pdf)
Significance:        Jun 30, 2011 Identified By: Self-Revealing Item Type: FIN Finding Technical Support Center Diesel Generator Output Breaker Fails To Close A finding of very low safety significance was self revealed for the failure to perform adequate preventive maintenance on latching relay VR1/B46, a relay required for closure of the technical support center (TSC) diesel generators (DG's) output breaker and automatic restoration of bus 1-46, which powers the TSC DGs cooling system. Specifically, on March 20, 2011, during a partial loss of offsite power event, the TSC DG started but failed to load onto bus 1-46.
After approximately 43 minutes of operation, the DG automatically shut down from an over-temperature condition, as designed. The licensee initiated condition report 417289 and performed apparent cause evaluation 018573. The licensees short-term corrective actions included troubleshooting the initial failure, repairing relay VR1/B46, and restoring the TSC DG to functional status. The licensees long-term corrective actions were in-progress at the completion of this inspection period.
The finding 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.
Specifically, the failure of the output breaker to close and energize bus 1-46 caused the TSC DG to overheat and automatically shut down during a partial loss of offsite power. The inspectors concluded the finding could be evaluated in accordance with Inspection Manual Chapter 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Table 4a, for the Mitigating Systems Cornerstone. The inspectors answered Yes to questions 2 and 4 of the Mitigating Systems Cornerstone column and determined that the finding required a Phase 2 analysis. The Region III senior reactor analyst completed a Phase 2 analysis and determined the risk significance of the issue to be very low (Green). The finding has a cross-cutting aspect in the area of human performance, resources, because a licensee effort to review various plant components for possible inclusion in a preventive maintenance optimization project had assigned a low priority to this relay (H.2(a)).
 
Inspection Report# : 2011003 (pdf)
Significance:      Jun 30, 2011 Identified By: NRC Item Type: FIN Finding Inadequate Operability Determination Of A Heat Exchanger Leak On Emergency Diesel Generator A A finding of very low safety significance was identified by the inspectors for the failure to adequately assess operability of the service water system in operability determination 413, EDG A Jacket Water Expansion Tank Overflow, in accordance with site Procedure OP-AA-102-1001, Development of Technical Basis to Support Operability Determinations. At the end of the inspection period, the licensee was completing an apparent cause evaluation to determine the cause and develop corrective actions.
The finding was determined to be more than minor because the finding, if left uncorrected, had the potential to become a more significant safety concern. Specifically, the failure to perform operability evaluations on degraded safety-related systems could lead to situations where systems needed to mitigate design basis accidents were not capable of performing their required safety functions. The inspectors determined the finding could be evaluated using Inspection Manual Chapter 0609, Significance Determination Process, Attachment 0609.04, Phase 1- Initial Screening and Characterization of Findings, Table 4a, for the Mitigating Systems Cornerstone. The inspectors answered "No" to the Mitigating Systems questions and screened the finding as having very low safety significance (Green). The finding has a cross-cutting aspect in the area of human performance, work practices, because the licensee failed to communicate decisions and the bases for decisions to personnel who had a need to know the information in order to perform work safely. Specifically, the licensee failed to effectively communicate the expectation to assess operability of the service water system in the pre-job brief and peer review (H.1(c)).
Inspection Report# : 2011003 (pdf)
Significance:      Jun 30, 2011 Identified By: NRC Item Type: FIN Finding Failure To Review And Update Severe Accident Management Guidelines In Accordance With An Established Program A finding of very low safety significance was identified by the inspectors for the licensees failure to perform reviews and update the Severe Accident Management Guidelines (SAMGs) in accordance with the licensees nuclear administrative directives (NADs). Specifically, Procedure NAD 14.06 required that the engineering group review industry correspondence related to SAMGs and implement appropriate changes, and that the emergency preparedness group conduct biennial reviews of the SAMGs. The inspectors identified that neither group had performed the reviews. As a result, the SAMGs were not adequately updated. The licensee entered this issue into their corrective action program as condition reports 424681, 424855, 424865, 424866, 425092, 426999, and 427092, and was still evaluating the cause for this condition at the end of this inspection period. The licensee scheduled the revision of the SAMGs for completion by December 2011.
The finding was determined to be more than minor because, if left uncorrected, the finding had the potential to lead to a more significant safety concern. Specifically, the failure to review and update the SAMGs would have hampered the licensees response in the unlikely event of a severe accident, because the SAMGs were not current. The inspectors, in consultation with the Region III senior reactor analyst, determined that the finding could be evaluated using the Significance Determination Process in accordance with Inspection Manual Chapter 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Table 4a, for the Mitigating Systems Cornerstone. The inspectors answered "No" to the Mitigating Systems questions and screened the finding as having very low safety significance (Green). The finding has a cross-cutting aspect in the area of problem identification and resolution, corrective action program, because the licensee failed to take appropriate corrective actions to address safety issues and adverse trends in a timely manner, commensurate with their safety significance and complexity. Specifically, the licensee identified in an apparent cause evaluation initiated in April 2010 that the emergency preparedness organization had not performed the required reviews and updates of emergency preparedness procedures, and the SAMGs were identified in the licensees extent of condition. However, the inspectors identified that the corrective actions issued for this extent of condition did not address the SAMGs and, therefore, no corrective actions were taken (P.1(d)).
 
Inspection Report# : 2011003 (pdf)
Significance:        Mar 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Work Instructions Results In Potential Orange Path A finding of very low safety-significance and associated non-cited violation (NCV) of Technical Specification 5.4.1, Procedures, was identified by the inspectors for the failure to implement procedures for shutdown operations involving shutdown operations safety assessments. Specifically, OU KW 201, Shutdown Safety Assessment Checklist, step 3.3.1, stated, in part, that a shutdown safety assessment was required to be completed in accordance with the procedure for core cooling; however, the inspectors noted that the February 28, 2011, 6:00 p.m. analysis credited the safety injection system feed and bleed as an available alternate decay heat removal system when the system was not available as described in Section 5.3.2, Available/Availability, for work scheduled at that time on the emergency core cooling system (ECCS) sump. The licensee initiated condition report CR415539, and at the end of the inspection period, the licensee was performing a causal evaluation to determine the causes of the event and develop corrective actions. On February 28, as a remedial corrective action prior to the start of work, additional steps to the work instructions were added to ensure the equipment would meet the intended function, operators were designated to perform the local manual operations and a pre job brief was conducted that provided training for using the equipment in the given situation.
The finding was determined to be more than minor because the finding was associated with the Mitigating Systems Cornerstone attribute of human error (pre event) and adversely 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). Specifically, the availability of the ECCS sump was integral to ensuring that the plant was not in an orange risk path for the evolutions completed on February 28. The inspectors screened the finding as of very low safety-significance (Green) because the finding did not degrade the licensees ability to establish an alternate core cooling path if decay heat removal could not be re established and, therefore, did not require a Significance Determination Process phase 2 or phase 3 analysis. The finding has a cross-cutting aspect in the areas of human performance, work control, because the licensee failed to plan the work activities by incorporating the need for planned contingencies and compensatory actions to ensure the ECCS sump was available to ensure an orange risk path for core cooling was not entered (H.3(a)).
Inspection Report# : 2011002 (pdf)
Significance:        Mar 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Unintended Voiding Of The Reactor Vessel Closure Head A finding of very low safety-significance and associated non-cited violation (NCV) of Technical Specification 5.4.1, Procedures, was identified by the inspectors for the failure to establish, implement, and maintain procedures for shutdown operations involving the draining of reactor coolant system (RCS) inventory. Specifically, on March 21, 2011, during a pressurizer draindown evolution, licensed operators unknowingly created a gas void in the reactor vessel closure head (RVCH) that displaced water to a level near the RVCH flange. Subsequent evaluation determined that the procedure for draining the RCS did not contain adequate guidance to ensure that an unacceptable void in the RVCH was not present prior to or formed during operations draindown activities. The licensee subsequently entered the issue into its corrective action program as CR418537 and performed a remedial corrective action of removing the gas void that accumulated in the RVCH. At the end of the inspection period, the licensee was performing an apparent cause evaluation to determine the causes of the event and develop additional corrective actions.
The finding was determined to be more than minor because the finding was associated with the Mitigating Systems Cornerstone attribute of operating procedure quality and adversely 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). Specifically, the formation of the gas void in the RVCH displaced RCS inventory and could have challenged the ability to remove decay heat in the event of a loss of shutdown cooling. The Region III senior reactor analyst determined that this issue is best characterized as a finding of very low safety-significance (Green). The finding has a cross-cutting aspect in the areas of human performance, work practices, because operations personnel
 
did not follow or implement the guidance contained in plant procedures. Specifically, procedure OP KW AOP RC 002 prescribed actions to take if a gas void formed in the RVCH that resulted in RVLIS level readings less than 88 percent, which had occurred several hours prior to the start of a pressurizer draining evolution (H.4(b)).
Inspection Report# : 2011002 (pdf)
Significance:        Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Incorrect Containment Fan Coil Unit Acceptance Criteria A finding of very low safety significance and associated non cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors for the failure to correctly translate the applicable regulatory requirements and the design basis into procedures and instructions. Specifically, the licensee failed to adequately translate the containment fan coil unit (CFCU) service water flow acceptance criteria from the current design basis calculations into the CFCU performance monitoring procedures, which resulted in the incorrect acceptance criteria in plant test procedures. The licensee took immediate corrective actions to correct the acceptance criteria in the test procedures and to perform an operability determination on CFCU C, the only one of the four CFCUs that showed a recent decrease in flow. At the end of the inspection period, the licensee was completing an apparent cause evaluation and developing additional long term corrective actions.
The finding was determined to be more than minor in accordance with Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated December 24, 2009, because it was associated with the Mitigating Systems Cornerstone attribute of procedure quality and adversely affected the cornerstone objective of ensuring the reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to ensure that procedure PMP 18 13, "Containment Fan Coil Unit Performance Monitoring (AQ-1)," contained the correct acceptance criteria for testing the CFCUs. The inspectors determined the finding could be evaluated using the Significance Determination Process (SDP) in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Tables 3b and 4a for the Mitigating Systems Cornerstone, dated January 10, 2008.
The inspectors answered "no" to the Mitigating Systems questions and screened the finding as having very low significance (Green). This finding has a cross cutting aspect in the area of human performance, Resources, because the licensee did not maintain complete, accurate, and up to date procedures. Specifically, the correct acceptance criteria for testing the CFCUs from the design basis calculations were not specified in the CFCU testing procedure (H.2(c)).
Inspection Report# : 2010005 (pdf)
Significance:        Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Unacceptable Preconditioning of Safety-Related Pressure Switches A finding of very low significance and associated non cited violation (NCV) of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the failure to develop and implement an adequate surveillance test procedure to accurately assess the as found trip setpoint for the pressure switches associated with the turbine building service water isolation function and various other safety related functions. The licensee initiated condition report CR401813, performed an apparent cause evaluation, and initiated corrective actions to address the issues identified in the casual evaluation.
The finding was determined to be more than minor in accordance with Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated December 24, 2009, because it was associated with the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences and affected the cornerstone attribute of Equipment Performance. The inspectors determined the finding could be evaluated using the Significance Determination Process (SDP) in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Tables 3b and 4a for the Mitigating Systems Cornerstone, dated January 10, 2008. The inspectors answered no to the Mitigating Systems questions and screened the finding as having very low significance (Green). This finding has a cross cutting aspect in the area of problem identification and resolution, Operating Experience, because the licensee did not evaluate and communicate external
 
operating experience to internal stakeholders in a timely manner (P.2(a)).
Inspection Report# : 2010005 (pdf)
Significance:        Dec 31, 2010 Identified By: NRC Item Type: NCV NonCited Violation Inappropriate Isolation of the Safety Injection Pump Minimum Flow Recirculation Lines A finding of very low safety significance and associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for multiple inadequate procedures, which directed closing the common train safety injection minimum flow recirculation line valves, an activity affecting quality. Specifically, station procedures directed operators to close the safety injection pump minimum flow recirculation valves in order to complete valve timing tests, and to engage an interlock that allowed closure of the containment sump recirculation valves. However, the procedures and licensed operators failed to recognize that closure of either minimum flow recirculation valve affected the operability and availability of both safety injection pumps for certain design basis accidents because the minimum flow recirculation path was isolated.
The licensee subsequently entered the issue into its corrective action program as CR393930. The licensee corrected the procedure inadequacies and completed a root cause evaluation that recommended several corrective actions to prevent recurrence.
The finding was determined to be more than minor in accordance with Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated December 24, 2009, because it was associated with the Mitigating System Cornerstone attribute of procedure quality, 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 ensure that procedures implemented during power operations ensured the operability of both trains of safety injection. The inspectors determined the finding could be evaluated using the Significance Determination Process (SDP) in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Tables 3b and 4a for the Mitigating Systems Cornerstone, dated January 10, 2008. The inspectors answered "yes" to the Mitigating Systems question that confirmed the finding represented a loss of system safety function. The Region III Senior Reactor Analyst (SRA) performed an SDP Phase 2 analysis and a Phase 3 analysis. The Phase 3 analysis determined that the resultant delta core damage frequency (CDF) was less than 1E 6 and delta large early release frequency (LERF) was less than 1E 7, which represented a Green finding. The dominant scenario involved a small break loss of coolant accident with operator failure to perform a rapid cool down. The finding has a cross cutting aspect in the area of human performance, Decision Making, because although the licensee procedures cautioned that starting a safety injection pump following the closure of a minimum flow recirculation valve would result in damage to the pump, the licensee staff failed to use conservative decision making to question the adequacy of the prescribed procedure actions (H.1(b)).
Inspection Report# : 2010005 (pdf)
Barrier Integrity Significance:        Jun 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failed Standoffs Result In An Inoperable Train of Shield Building Ventilation A finding of very low safety significance and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by inspectors for the failure to have and follow adequate procedures for evaluation and installation of components in shield building ventilation (SBV) train A. Specifically, the licensee failed to have adequate procedures to direct the completion of a subcomponent classification evaluation (SCE) and prevent non safety-related parts from being installed in safety-related applications; have torque specifications for the standoffs (spacers for circuit cards) in the work instructions; and properly accomplish the SCE procedure when evaluating the standoffs. The licensees initial short-term corrective actions removed the installed
 
standoffs from both trains. The licensee also performed an extent of condition looking at previously completed item equivalency evaluations to determine if an SCE was needed or missing for newly installed components.
The finding was determined to be more than minor because the finding was associated with the Barrier Integrity Cornerstone attribute of procedure quality, 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 licensee failed to have and follow adequate procedures which led to the failure of SBV train A. The inspectors determined that this was a type B containment finding since it was related to a degraded condition that had potential important implications for the integrity of the containment, without affecting the likelihood of core damage.
The inspector evaluated the finding using the significance determination process (SDP) in accordance with Inspection Manual Chapter 0609, Appendix H, Containment Integrity SDP, Table 4.1, and determined that the finding did not relate to a containment structure, system, and component, nor containment status that had an impact on large early release frequency. Because of this, the issue screened as Green, using the flowchart in Figure 4.1. The finding has a cross-cutting aspect in the area of problem identification and resolution, corrective action program, because the licensee failed to thoroughly evaluate problems such that the resolutions address causes and extent of conditions, as necessary. This includes properly classifying, prioritizing, and evaluating for operability and reportability conditions adverse to quality. This also includes, for significant problems, conducting effectiveness reviews of corrective actions to ensure that the problems are resolved. Specifically, the licensee failed to properly evaluate and identify the cause of the SBV train A failure and produce a resolution that addressed the cause (P.1(c)).
Inspection Report# : 2011003 (pdf)
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.
Miscellaneous Significance: SL-IV Jun 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure To Submit LER Per 10 CFR 50.73 A Severity Level IV non-cited violation of 10 CFR Part 50.73(a)(2)(i)(B) and 50.73(a)(2)(v)(C) was identified by the inspectors for the failure of the licensee to report an event or condition that was prohibited by Technical Specifications, and an event or condition that could have prevented the fulfillment of a safety function, that is relied upon to control the release of radioactive material. Specifically, the licensee failed to report the inoperability of shield building ventilation train A from December 3, 2010, through January 26, 2011, a condition prohibited by Technical Specification 3.6.c.1, which allowed a single train outage time of seven days. Additionally, shield building ventilation
 
train B was inoperable on multiple occasions during the same time period, requiring the licensee to also report an event or condition that could have prevented the fulfillment of a safety function, which is relied upon to control the release of radioactive material. At the end of the inspection period, the licensee was completing an apparent cause evaluation to determine the cause and develop corrective actions.
Because violations of 10 CFR 50.73 are considered to be violations that potentially impact the regulatory process, they are dispositioned using the traditional enforcement process instead of the Reactor Oversight Process Significance Determination Process. A cross-cutting aspect was not assigned to this violation. Per the NRC Enforcement Policy, Section 6.0, Violation Examples, a failure to submit a required licensee event report is categorized as a Severity Level IV violation.
Inspection Report# : 2011003 (pdf)
Last modified : January 04, 2012
 
Kewaunee 4Q/2011 Plant Inspection Findings Initiating Events Significance:        Dec 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Procedures For Reduced Inventory Operations Were Not Appropriate To Preclude Air Entrainment The inspectors identified a 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 failure to establish procedures for reduced inventory operations that were appropriate to manage gas accumulation. Specifically, the procedures did not preclude air entrainment into the residual heat removal (RHR) and reactor coolant systems (RCSs). This finding was entered into the licensees corrective action program. The licensee's immediate corrective actions included calculating the instrument inaccuracies for RHR flow and refueling level instrument loops, referencing the level inaccuracies based on inactive flow in RCS loops in the associated procedures., evaluating levels, and updating the procedures with a new graph.
The performance deficiency was determined to be more than minor because it was associated with the Initiating Event Cornerstone attribute of procedure quality, and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown operations. Specifically, the failure to establish procedures for reduced inventory operations that were appropriate to preclude air entrainment did not limit the likelihood of events that result from adverse air entrainment into the RHR and RCSs. The finding screened as having very low safety significance (Green) because the Region III Senior Reactor Analysts determined that it reasonably met the safety functions of core heat removal, RCS inventory control, power availability, containment control, and reactivity control; and there had been no actual air entrainment problems that had occurred using the procedures. The finding has a cross-cutting aspect in the area of Problem Identification and Resolution because the licensee did not thoroughly evaluate relevant operating experience. Specifically, the licensees evaluation of gas related issues in response to NRC Generic Letter (GL) 2008 01 was deficient in that it did not consider vortexing during reduced inventory operations. (P.2(a))
Inspection Report# : 2011005 (pdf)
Significance:        Oct 07, 2011 Identified By: NRC Item Type: NCV NonCited Violation Flammable Gas Bottles Installed and/or Stored in the Auxiliary Building The inspectors identified a finding of very low safety significance and associated NCV of Title 10, Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion III, Design Control, for the failure to check the adequacy of design for flammable gas bottles installed and/or stored in fire areas and fire zones located within the auxiliary building and their impact on safe shutdown cables, safety-related cables and safety-related equipment. Specifically, the licensee failed to evaluate how a failure of the flammable gas bottles and a resulting fire or explosion at the installed and/or stored locations could impact nearby safety-related structures, systems, or components. The licensee entered this issue into their corrective action program to review the placement of the flammable gas bottles.
The inspectors determined that the finding was more than minor because the finding was associated with the Initiating Events cornerstone attribute of Protection against External Factors (Fire) and affected the cornerstones objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The finding was of very low safety significance due to the low fire initiating frequency and the availability of remaining mitigating systems. This finding did not have a cross-cutting aspect because the finding was not representative of current performance.
Inspection Report# : 2011008 (pdf)
 
Significance:        Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Repetitive Molded Case Circuit Breaker Failures A finding of very low significance and associated non-cited violation of Title 10 of the Code of Federal Regulations (CFR) 50.65(a)(3) was identified by the inspectors for the failure to incorporate industry operating experience into preventive maintenance activities when practical to do so. Specifically, the failure to incorporate the industry operating experience resulted in multiple molded case circuit breaker (MCCB) failures that could have been prevented by implementing an MCCB cycling program. The need to cycle MCCBs was identified in industry operating experience as well as the vendor's instructions for the breakers. The licensee was performing an apparent cause evaluation which was still in progress at the conclusion of the inspection period. Initial corrective actions included scheduling the MCCBs for the breaker cycling maintenance activity.
This finding was determined to be of greater than minor significance because it was associated with the Protection Against External Factors attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events, such as fire, that challenge critical safety functions during shutdown as well as power operations. Specifically, the lack of a cycling program for safety related MCCBs resulted in breakers remaining in the on position after an overcurrent condition. The inspectors determined the finding had very low safety significance (Green) because the breakers and associated cabling did not significantly affect safe shutdown defense in depth strategies and the finding did not involve a design or qualification deficiency, did not represent a loss of system safety function, did not represent a loss of Technical Specification equipment for greater than its allowed outage time, and did not affect risk significant equipment per 10 CFR 50.65. This finding has a cross-cutting aspect in the area of human performance, work control, because the licensee did not emphasize the need for work groups to communicate, coordinate, and cooperate with each other during activities in which interdepartmental coordination is necessary to assure plant and human performance.
Inspection Report# : 2011004 (pdf)
Significance:        Mar 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Misapplication Of Code Acceptance Criteria For Weld Flaws A finding of very low safety-significance and associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion IX, Control of Special Processes, was identified by the inspectors on March 3, 2011, for the licensees failure to establish a procedure that incorporated the American Society of Mechanical Engineers Code acceptance criteria for evaluation of flaws detected during ultrasonic examinations. Consequently, the licensee applied incorrect acceptance criteria to the flaws identified during ultrasonic examination of a weld on the chemical and volume control system seal water injection filter 1A housing. Licensee corrective actions included: evaluation of weld flaws to ensure they met applicable Code criteria and revision of a site procedure to incorporate appropriate Code acceptance criteria.
The finding was determined to be more than minor because the finding, if left uncorrected, would become a more significant safety concern. Absent NRC identification, the failure to provide Code acceptance criteria could have allowed components with unacceptable cracks to be returned to service. Cracks in components returned to service would place safety related piping systems at increased risk for through wall leakage and/or failure. The licensee promptly corrected this issue before components with unacceptable flaws were returned to service. The inspectors answered No to the Significance Determination Process Phase I screening question, Assuming worst case degradation, would the finding result in exceeding the Technical Specification (TS) limit for any reactor coolant system leakage or could the finding have likely affected other mitigation systems resulting in a total loss of their safety function assuming the worst case degradation? Therefore, this finding screened as having very low safety-significance (Green). This finding has a cross-cutting aspect in the area of human performance, work practices, because the licensee did not effectively implement human error prevention techniques. Specifically, the lack of procedure acceptance criteria was caused by inadequate peer checking during the licensees review and approval of the procedure for evaluation of non destructive examination data (H.4(a)).
Inspection Report# : 2011002 (pdf)
 
Significance:      Mar 31, 2011 Identified By: Self-Revealing Item Type: FIN Finding Partial Loss Of Offsite Power Caused By Less Than Adequate Interface And Oversight Of Switchyard Modification Work A finding of very low safety-significance was self-revealed for the failure to adequately control relay testing for switchyard breaker installations under Design Change WO KW100691871. Specifically, on March 10, 2011, Dominion Electrical Transmission technicians deviated from standard work practices to test a relay via an internal corporate server, which caused a partial loss of offsite power to the plant through the loss of the main auxiliary transformer backfeed to safety-related bus 6. Licensee corrective actions included a human performance and safety stand down for substation personnel on the day of the event, the development of a mitigating strategy that outlined expectations and implemented increased direct supervision on critical tasks, and the development of a formal memo describing expectations related to the restricted use of the server for performing remote testing of control functions.
The finding was determined to be more than minor because, if left uncorrected, the finding had the potential to lead to a more significant safety concern. Specifically, had a different breaker been inappropriately tripped, the station could have experienced a total loss of offsite power. The inspectors concluded that the finding could be evaluated using Inspection Manual Chapter 0609, Appendix M, Significance Determination Process Using Qualitative Criteria.
Specifically, the inspectors qualitatively evaluated the finding by applying the spent fuel pool questions in the Fuel Barrier column of Table 4a, Attachment 4. The inspectors answered "No" to all three questions and determined that the finding was of very low safety-significance (Green). The finding has a cross-cutting aspect in the areas of human performance, work practices, because supervisory and management oversight of work activities, including contractors, was not implemented for this evolution (H.4(c)).
Inspection Report# : 2011002 (pdf)
Mitigating Systems Significance:      Dec 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Unanalyzed Flood Source From Technical Support Center Building A finding of very low safety significance and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors for the failure to identify and analyze a potential flooding source that was within the Kewaunee licensing basis. Specifically, during the internal flood basis reconstitution in 2005, the licensee failed to realize and assess the potential for fire main piping in the technical support center (TSC) building to be ruptured during a tornado or seismic event. Water from a ruptured fire main had the potential to accumulate in the basement of the TSC building, flow into the attached auxiliary building, and potentially affect safety related (SR) equipment. The licensee initiated a condition report (CR) and completed calculations and analyses to demonstrate the existing barriers, although not credited at the time, were adequate to support this internal flood scenario. In addition, the licensee performed an extent of condition analysis to determine if any additional internal flood scenarios were missed.
The inspectors determined that the finding was more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated December 24, 2009, because it was associated with the Mitigating Systems (MS) 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 failure to analyze all potentially credible internal flood sources could affect the availability of SR systems. The inspectors determined that the finding could be evaluated using the significance determination process in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings, Table 4a, for the MS Cornerstone, dated January 10, 2008. The inspectors answered "No" to the MS questions and screened the finding as having very low safety significance (Green). The inspectors did not identify a cross-cutting aspect associated with this finding because the internal flood design basis reconstitution occurred in 2005 and the inspectors determined that there was not an
 
opportunity to identify this deficiency in the past three years.
Inspection Report# : 2011005 (pdf)
Significance:      Dec 31, 2011 Identified By: Self-Revealing Item Type: FIN Finding Two Of Six Operating Crew Failures On The Simulator Operational Evaluation Portion Of The 2011 Annual Requalification Operating Test A self-revealed finding associated with operating crew performance on the simulator during a licensee-administered requalification examination was identified. Two of the six crews evaluated during the annual operating tests failed to pass their simulator examinations. As immediate corrective action, the failed operating crews were remediated (i.e.,
the operating crews were re-trained and successfully re-tested) prior to returning to shift. The licensee entered this issue into the CAP as CR456328.
The inspectors determined that the crew failures constituted a performance deficiency based on the fact that licensed operators are expected to operate the plant with acceptable standards of knowledge and abilities demonstrated through periodic testing as required by 10 CFR 55.59(a)(2). Two out of six crews of licensed operators failed to demonstrate a satisfactory understanding of the required actions and mitigating strategies required to safely operate the facility under normal, abnormal, and emergency conditions. The finding was more than minor because the performance deficiency potentially affects the Human Performance attribute of the MS Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.
Specifically, the finding reflected the potential inability of the crews to take appropriate SR action in response to actual abnormal and emergency conditions. The perceived risk associated with the number of crews failing the annual operating test is provided in the Simulator Operational Evaluation matrix of IMC 0609, Appendix I, Licensed Operator Requalification SDP. The finding was of very low safety significance (Green) because only two of six of the operating crews failed; the failed operating crews were remediated (i.e., the operating crews were re-trained and successfully re-tested) prior to returning to shift; and because there was not a finding associated with operating crew failures during calendar year 2010. The cause of this finding was directly related to the cross-cutting aspect of personnel training and qualifications in the area of Human Performance - Resources, in that the licensee failed to ensure the adequacy of the training provided to operators to assure nuclear safety. (H.2(b))
Inspection Report# : 2011005 (pdf)
Significance:      Dec 31, 2011 Identified By: NRC Item Type: FIN Finding Inadequate Operability Determination Of Control Room Air Conditioning System Components A finding of very low safety significance was identified by the inspectors for the failure to adequately assess operability of the Control Room Air Conditioning Component (ACC) dampers, ACC-15 and ACC-16, in Operability Determination (OD) 456, Revision 0, ACC-15 and ACC-16 QA Classification, in accordance with site Procedure OP AA 102 1001, Development of Technical Basis to Support Operability Determinations, Revision 4. The licensee entered the issue into their CAP and was completing an apparent cause evaluation at the conclusion of the inspection period.
The finding was determined to be more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated December 24, 2009, because the finding, if left uncorrected, had the potential to become a more significant safety concern. Specifically, the failure to give the operators written instructions to manually reposition the SR dampers could have lead to situations where the operators would not have been able to rapidly and correctly manually reposition the SR dampers to perform their required safety functions necessary to mitigate design basis accidents. The inspectors determined the finding could be evaluated using IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings, Table 4a, for the MS Cornerstone, dated January 10, 2008. The inspectors answered "No" to the MS questions and screened the finding as having very low significance (Green). The finding has a cross-cutting aspect in the area of Human Performance - Decision Making, because the licensee failed to communicate decisions and the bases for decisions to personnel who had a need to know the information in order to perform work safely, in a timely manner. Specifically, the licensee failed to communicate in a timely manner to the reactor operators the written
 
instructions in the standing order necessary to manually reposition the dampers to their SR positions after a design basis accident. (H.1(c))
Inspection Report# : 2011005 (pdf)
Significance:        Dec 31, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Unanticipated Closure Of Emergency Diesel Generator B Output Breaker A finding of very low safety significance and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed for the failure to implement a procedure for an activity affecting quality. Procedure OP-KW-OSP-DGE-003B, Diesel Generator B Semi-Annual, required electrical maintenance personnel to check only the voltage of the emergency diesel generator (EDG) B output breaker Relay 52C/1-603; however, the electricians checked voltage and then attempted to check resistance of the relay. Specifically, after successfully testing for voltage, an electrician then selected a resistance setting for the volt-ohm meter (VOM) in an attempt to perform a continuity check of the relay, which was not prescribed by the procedure. The electricians actions resulted in the closure of the EDG output Breaker 1 603, and EDG B was paralleled to the grid out-of-phase.
The licensee initiated a condition report and took remedial corrective actions that included additional testing and inspections of EDG B to ensure that no damage occurred to the equipment as a result of the system transient, followed by the successful completion of post maintenance testing. At the end of the inspection period, the licensee was performing a root cause evaluation to determine the cause of the event and to develop additional corrective actions related to the organizational performance issues.
The inspectors determined that the finding was more than minor in accordance with IMC 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," because the finding was associated with the MS Cornerstone attribute of Equipment Performance, and adversely impacted the objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the attribute of equipment performance impacted the availability and reliability of EDG B and could have resulted in the catastrophic failure of the generator. The inspectors determined that 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 MS Cornerstone, dated January 10, 2008. The inspectors answered "No" to the MS questions and screened the finding as having very low safety significance (Green). The inspectors determined that this finding has a cross -cutting aspect in the area of Human Performance - Work Practices, because the maintenance personnel and supervision failed to communicate and ensure human error prevention techniques were used, such as holding formal pre job briefings, and self and peer checking. The licensee also failed to ensure that these techniques were used commensurate with the potential risk of the assigned task, such that work activities were performed safely. Finally, during these maintenance activities, the inspectors concluded that licensee personnel proceeded in the face of uncertainty. (H.4(a))
Inspection Report# : 2011005 (pdf)
Significance:        Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure To Maintain Fire Barrier And Automatic Fire Suppression A finding of very low safety significance and associated non-cited violation of license condition 2.C(3) of the Kewaunee Power Station Renewed Operating License was identified by inspectors for the failure to have a self-closing fire door that closed and latched each time it was open. License condition 2.C(3) requires, in part, that the licensee implement and maintain, in effect, all provisions of the approved fire protection program as described in the licensees fire plan. Appendix B of the Kewaunee Power Station Fire Protection Program Plan lists the 1975 edition of NFPA 80 [National Fire Protection Association], Fire Doors and Windows, as an applicable NFPA code. NFPA 80 states, in part, that a self closing door shall be equipped with a closing device to cause the door to close and latch each time it is opened. The licensee entered the issue into its corrective action program and adjusted the door closing device to ensure the door properly closed when the train A screenhouse ventilation fan was operating.
The inspectors determined that the failure of the door to close and latch was contrary to the requirements of NFPA 80 and was a performance deficiency. The finding was determined to be more than minor because it was associated with
 
the Mitigating Systems Cornerstone attribute of protection against external factors (Fire) and adversely 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). The inspectors screened the finding as having very low safety significance (Green). The 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 that the resolutions address causes and extent of conditions. This includes properly classifying, prioritizing, and evaluating for operability and reportability conditions adverse to quality.
Inspection Report# : 2011004 (pdf)
Significance:        Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Misapplication Of Technical Specification 3.1.6 Applicability Note A finding of very low safety significance and associated non-cited violation of Technical Specification (TS) 3.1.6, Control Bank Insertion Limits, was identified by the inspectors for the failure to comply with TS action condition 3.1.6.A due to incorrect use of the applicability note. Specifically, on August 30, 2011, during the performance of SP-49-075, Control Rod Exercise, operators received a rod control urgent failure while inserting control bank A group 1 control rods. The test was suspended for troubleshooting for approximately 20 hours with control bank A group 2 control rods, inserted one step below the control rod insertion limit in violation of TS 3.1.6.A action condition. The inspectors concluded that, once the test was suspended for troubleshooting activities, use of the applicability note was not appropriate; therefore, the operators should have complied with the TS 3.1.6.A action condition for control bank A group 2 control rods at that time. On August 31, operators withdrew control bank A group 2 rods one step, which restored the rods to within the limit specified in the core operating limits report. At the end of this inspection period, the licensee was still performing an apparent cause evaluation to determine the causes of the event and to develop corrective actions.
The finding was determined to be more than minor because the finding adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent desirable consequences. Specifically, the human performance attributes of the licensees failure to recognize the misapplication of the applicability note of the TS affected the capability of systems that respond to initiating events. The inspectors screened the finding as having very low safety significance (Green) because an actual loss of safety function did not occur. The finding has a cross-cutting aspect in the area of human performance, decision making, because the licensee failed to use conservative assumptions and adopt a requirement to demonstrate that the proposed action is safe in order to proceed rather than a requirement to demonstrate that it is unsafe in order to disapprove the action.
Inspection Report# : 2011004 (pdf)
Significance:        Sep 30, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Incorrect Transformer Load Tap Changer Setting Causes Inoperable Offsite Power A finding of very low safety significance and associated non-cited violation of Technical Specification 3.8.1 was self revealed for the failure to maintain a switchyard transformer load tap changer (LTC) at the appropriate setting for the predicted post trip voltage of offsite power. The incorrect setting resulted in the inoperability of the Reserve Auxiliary Transformer (RAT) offsite power source. The licensees corrective actions included restoring the RAT supply transformer (RST) LTC to an appropriate setting, creating a short term standing order to prevent operation of the RST LTC outside settings that were supported by the existing interface agreement with the transmission system operator.
The licensee performed an apparent cause evaluation, a root cause analysis and also, as a long-term corrective action, modified procedure OP-KW-NOP SUB 003 to prevent operation of the RST LTC outside settings that were supported by the existing interface agreement with the transmission system operator.
The finding was determined to be more than minor because the finding was associated with the Mitigating Systems Cornerstone attribute of configuration control and adversely 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). The inspectors screened the finding as having very low safety significance (Green). This finding has a cross-cutting aspect in the area of human performance, resources, because the licensee did not maintain
 
complete, accurate, and up to date procedures for the use of the RST LTC following its installation during the spring 2011 outage.
Inspection Report# : 2011004 (pdf)
Significance:      Jun 30, 2011 Identified By: Self-Revealing Item Type: FIN Finding Technical Support Center Diesel Generator Output Breaker Fails To Close A finding of very low safety significance was self revealed for the failure to perform adequate preventive maintenance on latching relay VR1/B46, a relay required for closure of the technical support center (TSC) diesel generators (DG's) output breaker and automatic restoration of bus 1-46, which powers the TSC DGs cooling system. Specifically, on March 20, 2011, during a partial loss of offsite power event, the TSC DG started but failed to load onto bus 1-46.
After approximately 43 minutes of operation, the DG automatically shut down from an over-temperature condition, as designed. The licensee initiated condition report 417289 and performed apparent cause evaluation 018573. The licensees short-term corrective actions included troubleshooting the initial failure, repairing relay VR1/B46, and restoring the TSC DG to functional status. The licensees long-term corrective actions were in-progress at the completion of this inspection period.
The finding 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.
Specifically, the failure of the output breaker to close and energize bus 1-46 caused the TSC DG to overheat and automatically shut down during a partial loss of offsite power. The inspectors concluded the finding could be evaluated in accordance with Inspection Manual Chapter 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Table 4a, for the Mitigating Systems Cornerstone. The inspectors answered Yes to questions 2 and 4 of the Mitigating Systems Cornerstone column and determined that the finding required a Phase 2 analysis. The Region III senior reactor analyst completed a Phase 2 analysis and determined the risk significance of the issue to be very low (Green). The finding has a cross-cutting aspect in the area of human performance, resources, because a licensee effort to review various plant components for possible inclusion in a preventive maintenance optimization project had assigned a low priority to this relay (H.2(a)).
Inspection Report# : 2011003 (pdf)
Significance:      Jun 30, 2011 Identified By: NRC Item Type: FIN Finding Inadequate Operability Determination Of A Heat Exchanger Leak On Emergency Diesel Generator A A finding of very low safety significance was identified by the inspectors for the failure to adequately assess operability of the service water system in operability determination 413, EDG A Jacket Water Expansion Tank Overflow, in accordance with site Procedure OP-AA-102-1001, Development of Technical Basis to Support Operability Determinations. At the end of the inspection period, the licensee was completing an apparent cause evaluation to determine the cause and develop corrective actions.
The finding was determined to be more than minor because the finding, if left uncorrected, had the potential to become a more significant safety concern. Specifically, the failure to perform operability evaluations on degraded safety-related systems could lead to situations where systems needed to mitigate design basis accidents were not capable of performing their required safety functions. The inspectors determined the finding could be evaluated using Inspection Manual Chapter 0609, Significance Determination Process, Attachment 0609.04, Phase 1- Initial Screening and Characterization of Findings, Table 4a, for the Mitigating Systems Cornerstone. The inspectors answered "No" to the Mitigating Systems questions and screened the finding as having very low safety significance (Green). The finding has a cross-cutting aspect in the area of human performance, work practices, because the licensee failed to communicate decisions and the bases for decisions to personnel who had a need to know the information in order to perform work safely. Specifically, the licensee failed to effectively communicate the expectation to assess operability of the service water system in the pre-job brief and peer review (H.1(c)).
Inspection Report# : 2011003 (pdf)
 
Significance:        Jun 30, 2011 Identified By: NRC Item Type: FIN Finding Failure To Review And Update Severe Accident Management Guidelines In Accordance With An Established Program A finding of very low safety significance was identified by the inspectors for the licensees failure to perform reviews and update the Severe Accident Management Guidelines (SAMGs) in accordance with the licensees nuclear administrative directives (NADs). Specifically, Procedure NAD 14.06 required that the engineering group review industry correspondence related to SAMGs and implement appropriate changes, and that the emergency preparedness group conduct biennial reviews of the SAMGs. The inspectors identified that neither group had performed the reviews. As a result, the SAMGs were not adequately updated. The licensee entered this issue into their corrective action program as condition reports 424681, 424855, 424865, 424866, 425092, 426999, and 427092, and was still evaluating the cause for this condition at the end of this inspection period. The licensee scheduled the revision of the SAMGs for completion by December 2011.
The finding was determined to be more than minor because, if left uncorrected, the finding had the potential to lead to a more significant safety concern. Specifically, the failure to review and update the SAMGs would have hampered the licensees response in the unlikely event of a severe accident, because the SAMGs were not current. The inspectors, in consultation with the Region III senior reactor analyst, determined that the finding could be evaluated using the Significance Determination Process in accordance with Inspection Manual Chapter 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Table 4a, for the Mitigating Systems Cornerstone. The inspectors answered "No" to the Mitigating Systems questions and screened the finding as having very low safety significance (Green). The finding has a cross-cutting aspect in the area of problem identification and resolution, corrective action program, because the licensee failed to take appropriate corrective actions to address safety issues and adverse trends in a timely manner, commensurate with their safety significance and complexity. Specifically, the licensee identified in an apparent cause evaluation initiated in April 2010 that the emergency preparedness organization had not performed the required reviews and updates of emergency preparedness procedures, and the SAMGs were identified in the licensees extent of condition. However, the inspectors identified that the corrective actions issued for this extent of condition did not address the SAMGs and, therefore, no corrective actions were taken (P.1(d)).
Inspection Report# : 2011003 (pdf)
Significance:        Mar 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Work Instructions Results In Potential Orange Path A finding of very low safety-significance and associated non-cited violation (NCV) of Technical Specification 5.4.1, Procedures, was identified by the inspectors for the failure to implement procedures for shutdown operations involving shutdown operations safety assessments. Specifically, OU KW 201, Shutdown Safety Assessment Checklist, step 3.3.1, stated, in part, that a shutdown safety assessment was required to be completed in accordance with the procedure for core cooling; however, the inspectors noted that the February 28, 2011, 6:00 p.m. analysis credited the safety injection system feed and bleed as an available alternate decay heat removal system when the system was not available as described in Section 5.3.2, Available/Availability, for work scheduled at that time on the emergency core cooling system (ECCS) sump. The licensee initiated condition report CR415539, and at the end of the inspection period, the licensee was performing a causal evaluation to determine the causes of the event and develop corrective actions. On February 28, as a remedial corrective action prior to the start of work, additional steps to the work instructions were added to ensure the equipment would meet the intended function, operators were designated to perform the local manual operations and a pre job brief was conducted that provided training for using the equipment in the given situation.
The finding was determined to be more than minor because the finding was associated with the Mitigating Systems Cornerstone attribute of human error (pre event) and adversely 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). Specifically, the availability of the ECCS sump was integral to ensuring that the plant was not in an orange risk path for the evolutions completed on February 28. The inspectors screened the finding as of very low
 
safety-significance (Green) because the finding did not degrade the licensees ability to establish an alternate core cooling path if decay heat removal could not be re established and, therefore, did not require a Significance Determination Process phase 2 or phase 3 analysis. The finding has a cross-cutting aspect in the areas of human performance, work control, because the licensee failed to plan the work activities by incorporating the need for planned contingencies and compensatory actions to ensure the ECCS sump was available to ensure an orange risk path for core cooling was not entered (H.3(a)).
Inspection Report# : 2011002 (pdf)
Significance:        Mar 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Unintended Voiding Of The Reactor Vessel Closure Head A finding of very low safety-significance and associated non-cited violation (NCV) of Technical Specification 5.4.1, Procedures, was identified by the inspectors for the failure to establish, implement, and maintain procedures for shutdown operations involving the draining of reactor coolant system (RCS) inventory. Specifically, on March 21, 2011, during a pressurizer draindown evolution, licensed operators unknowingly created a gas void in the reactor vessel closure head (RVCH) that displaced water to a level near the RVCH flange. Subsequent evaluation determined that the procedure for draining the RCS did not contain adequate guidance to ensure that an unacceptable void in the RVCH was not present prior to or formed during operations draindown activities. The licensee subsequently entered the issue into its corrective action program as CR418537 and performed a remedial corrective action of removing the gas void that accumulated in the RVCH. At the end of the inspection period, the licensee was performing an apparent cause evaluation to determine the causes of the event and develop additional corrective actions.
The finding was determined to be more than minor because the finding was associated with the Mitigating Systems Cornerstone attribute of operating procedure quality and adversely 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). Specifically, the formation of the gas void in the RVCH displaced RCS inventory and could have challenged the ability to remove decay heat in the event of a loss of shutdown cooling. The Region III senior reactor analyst determined that this issue is best characterized as a finding of very low safety-significance (Green). The finding has a cross-cutting aspect in the areas of human performance, work practices, because operations personnel did not follow or implement the guidance contained in plant procedures. Specifically, procedure OP KW AOP RC 002 prescribed actions to take if a gas void formed in the RVCH that resulted in RVLIS level readings less than 88 percent, which had occurred several hours prior to the start of a pressurizer draining evolution (H.4(b)).
Inspection Report# : 2011002 (pdf)
Barrier Integrity Significance:        Jun 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failed Standoffs Result In An Inoperable Train of Shield Building Ventilation A finding of very low safety significance and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by inspectors for the failure to have and follow adequate procedures for evaluation and installation of components in shield building ventilation (SBV) train A. Specifically, the licensee failed to have adequate procedures to direct the completion of a subcomponent classification evaluation (SCE) and prevent non safety-related parts from being installed in safety-related applications; have torque specifications for the standoffs (spacers for circuit cards) in the work instructions; and properly accomplish the SCE procedure when evaluating the standoffs. The licensees initial short-term corrective actions removed the installed standoffs from both trains. The licensee also performed an extent of condition looking at previously completed item equivalency evaluations to determine if an SCE was needed or missing for newly installed components.
The finding was determined to be more than minor because the finding was associated with the Barrier Integrity Cornerstone attribute of procedure quality, 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 licensee failed to have and follow adequate procedures which led to the failure of SBV train A. The inspectors determined that this was a type B containment finding since it was related to a degraded condition that had potential important implications for the integrity of the containment, without affecting the likelihood of core damage.
The inspector evaluated the finding using the significance determination process (SDP) in accordance with Inspection Manual Chapter 0609, Appendix H, Containment Integrity SDP, Table 4.1, and determined that the finding did not relate to a containment structure, system, and component, nor containment status that had an impact on large early release frequency. Because of this, the issue screened as Green, using the flowchart in Figure 4.1. The finding has a cross-cutting aspect in the area of problem identification and resolution, corrective action program, because the licensee failed to thoroughly evaluate problems such that the resolutions address causes and extent of conditions, as necessary. This includes properly classifying, prioritizing, and evaluating for operability and reportability conditions adverse to quality. This also includes, for significant problems, conducting effectiveness reviews of corrective actions to ensure that the problems are resolved. Specifically, the licensee failed to properly evaluate and identify the cause of the SBV train A failure and produce a resolution that addressed the cause (P.1(c)).
Inspection Report# : 2011003 (pdf)
Emergency Preparedness Significance:        Dec 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Incorrect Assumptions Used In The Development Of Emergency Action Level Thresholds A finding of very low safety significance and associated non-cited violation of 10 CFR 50.54(q) was identified by the NRC for failing to maintain emergency plans that meet the requirements of emergency planning standard 10 CFR 50.47(b)(4). The inspectors determined that a performance deficiency existed in that incorrect assumptions were used in the development of Emergency Action Level (EAL) thresholds associated with containment gas (R12) and containment ventilation (R21) radiation monitors. The licensee entered this issue into its CAP as CR356229 and corrected the errant EAL thresholds in its emergency classification and action level scheme.
This finding was determined to be more than minor because the deficiency, if left uncorrected, could have the potential to lead to a more significant safety concern. Specifically, in the event of a radiological emergency, the deficiency has the potential to increase the risk to the public through a premature and/or unnecessary general emergency declaration and subsequent protective action recommendation of evacuation. This finding was evaluated using IMC 0609, Appendix B, Emergency Preparedness Significance Determination Process, Sheet 1, Failure To Comply. This finding is associated with a failure to meet or implement a regulatory requirement. The deficiency is not greater than Green because it did not result in the Risk-Significant Planning Standard Function being lost or degraded. No cross-cutting aspect is assigned to this finding because it is not indicative of current plant performance.
Inspection Report# : 2011005 (pdf)
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.
Miscellaneous Significance: SL-IV Jun 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure To Submit LER Per 10 CFR 50.73 A Severity Level IV non-cited violation of 10 CFR Part 50.73(a)(2)(i)(B) and 50.73(a)(2)(v)(C) was identified by the inspectors for the failure of the licensee to report an event or condition that was prohibited by Technical Specifications, and an event or condition that could have prevented the fulfillment of a safety function, that is relied upon to control the release of radioactive material. Specifically, the licensee failed to report the inoperability of shield building ventilation train A from December 3, 2010, through January 26, 2011, a condition prohibited by Technical Specification 3.6.c.1, which allowed a single train outage time of seven days. Additionally, shield building ventilation train B was inoperable on multiple occasions during the same time period, requiring the licensee to also report an event or condition that could have prevented the fulfillment of a safety function, which is relied upon to control the release of radioactive material. At the end of the inspection period, the licensee was completing an apparent cause evaluation to determine the cause and develop corrective actions.
Because violations of 10 CFR 50.73 are considered to be violations that potentially impact the regulatory process, they are dispositioned using the traditional enforcement process instead of the Reactor Oversight Process Significance Determination Process. A cross-cutting aspect was not assigned to this violation. Per the NRC Enforcement Policy, Section 6.0, Violation Examples, a failure to submit a required licensee event report is categorized as a Severity Level IV violation.
Inspection Report# : 2011003 (pdf)
Last modified : March 02, 2012
 
Kewaunee 1Q/2012 Plant Inspection Findings Initiating Events Significance:        Dec 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Procedures For Reduced Inventory Operations Were Not Appropriate To Preclude Air Entrainment The inspectors identified a 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 failure to establish procedures for reduced inventory operations that were appropriate to manage gas accumulation. Specifically, the procedures did not preclude air entrainment into the residual heat removal (RHR) and reactor coolant systems (RCSs). This finding was entered into the licensees corrective action program. The licensee's immediate corrective actions included calculating the instrument inaccuracies for RHR flow and refueling level instrument loops, referencing the level inaccuracies based on inactive flow in RCS loops in the associated procedures., evaluating levels, and updating the procedures with a new graph.
The performance deficiency was determined to be more than minor because it was associated with the Initiating Event Cornerstone attribute of procedure quality, and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown operations. Specifically, the failure to establish procedures for reduced inventory operations that were appropriate to preclude air entrainment did not limit the likelihood of events that result from adverse air entrainment into the RHR and RCSs. The finding screened as having very low safety significance (Green) because the Region III Senior Reactor Analysts determined that it reasonably met the safety functions of core heat removal, RCS inventory control, power availability, containment control, and reactivity control; and there had been no actual air entrainment problems that had occurred using the procedures. The finding has a cross-cutting aspect in the area of Problem Identification and Resolution because the licensee did not thoroughly evaluate relevant operating experience. Specifically, the licensees evaluation of gas related issues in response to NRC Generic Letter (GL) 2008 01 was deficient in that it did not consider vortexing during reduced inventory operations. (P.2(a))
Inspection Report# : 2011005 (pdf)
Significance:        Oct 07, 2011 Identified By: NRC Item Type: NCV NonCited Violation Flammable Gas Bottles Installed and/or Stored in the Auxiliary Building The inspectors identified a finding of very low safety significance and associated NCV of Title 10, Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion III, Design Control, for the failure to check the adequacy of design for flammable gas bottles installed and/or stored in fire areas and fire zones located within the auxiliary building and their impact on safe shutdown cables, safety-related cables and safety-related equipment. Specifically, the licensee failed to evaluate how a failure of the flammable gas bottles and a resulting fire or explosion at the installed and/or stored locations could impact nearby safety-related structures, systems, or components. The licensee entered this issue into their corrective action program to review the placement of the flammable gas bottles.
The inspectors determined that the finding was more than minor because the finding was associated with the Initiating Events cornerstone attribute of Protection against External Factors (Fire) and affected the cornerstones objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The finding was of very low safety significance due to the low fire initiating frequency and the availability of remaining mitigating systems. This finding did not have a cross-cutting aspect because the finding was not representative of current performance.
Inspection Report# : 2011008 (pdf)
 
Significance:      Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Repetitive Molded Case Circuit Breaker Failures A finding of very low significance and associated non-cited violation of Title 10 of the Code of Federal Regulations (CFR) 50.65(a)(3) was identified by the inspectors for the failure to incorporate industry operating experience into preventive maintenance activities when practical to do so. Specifically, the failure to incorporate the industry operating experience resulted in multiple molded case circuit breaker (MCCB) failures that could have been prevented by implementing an MCCB cycling program. The need to cycle MCCBs was identified in industry operating experience as well as the vendor's instructions for the breakers. The licensee was performing an apparent cause evaluation which was still in progress at the conclusion of the inspection period. Initial corrective actions included scheduling the MCCBs for the breaker cycling maintenance activity.
This finding was determined to be of greater than minor significance because it was associated with the Protection Against External Factors attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events, such as fire, that challenge critical safety functions during shutdown as well as power operations. Specifically, the lack of a cycling program for safety related MCCBs resulted in breakers remaining in the on position after an overcurrent condition. The inspectors determined the finding had very low safety significance (Green) because the breakers and associated cabling did not significantly affect safe shutdown defense in depth strategies and the finding did not involve a design or qualification deficiency, did not represent a loss of system safety function, did not represent a loss of Technical Specification equipment for greater than its allowed outage time, and did not affect risk significant equipment per 10 CFR 50.65. This finding has a cross-cutting aspect in the area of human performance, work control, because the licensee did not emphasize the need for work groups to communicate, coordinate, and cooperate with each other during activities in which interdepartmental coordination is necessary to assure plant and human performance.
Inspection Report# : 2011004 (pdf)
Mitigating Systems Significance:      Dec 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Unanalyzed Flood Source From Technical Support Center Building A finding of very low safety significance and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors for the failure to identify and analyze a potential flooding source that was within the Kewaunee licensing basis. Specifically, during the internal flood basis reconstitution in 2005, the licensee failed to realize and assess the potential for fire main piping in the technical support center (TSC) building to be ruptured during a tornado or seismic event. Water from a ruptured fire main had the potential to accumulate in the basement of the TSC building, flow into the attached auxiliary building, and potentially affect safety related (SR) equipment. The licensee initiated a condition report (CR) and completed calculations and analyses to demonstrate the existing barriers, although not credited at the time, were adequate to support this internal flood scenario. In addition, the licensee performed an extent of condition analysis to determine if any additional internal flood scenarios were missed.
The inspectors determined that the finding was more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated December 24, 2009, because it was associated with the Mitigating Systems (MS) 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 failure to analyze all potentially credible internal flood sources could affect the availability of SR systems. The inspectors determined that the finding could be evaluated using the significance determination process in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings, Table 4a, for the MS Cornerstone, dated January 10, 2008. The inspectors answered "No" to the MS questions and screened the finding as having very low safety
 
significance (Green). The inspectors did not identify a cross-cutting aspect associated with this finding because the internal flood design basis reconstitution occurred in 2005 and the inspectors determined that there was not an opportunity to identify this deficiency in the past three years.
Inspection Report# : 2011005 (pdf)
Significance:      Dec 31, 2011 Identified By: Self-Revealing Item Type: FIN Finding Two Of Six Operating Crew Failures On The Simulator Operational Evaluation Portion Of The 2011 Annual Requalification Operating Test A self-revealed finding associated with operating crew performance on the simulator during a licensee-administered requalification examination was identified. Two of the six crews evaluated during the annual operating tests failed to pass their simulator examinations. As immediate corrective action, the failed operating crews were remediated (i.e.,
the operating crews were re-trained and successfully re-tested) prior to returning to shift. The licensee entered this issue into the CAP as CR456328.
The inspectors determined that the crew failures constituted a performance deficiency based on the fact that licensed operators are expected to operate the plant with acceptable standards of knowledge and abilities demonstrated through periodic testing as required by 10 CFR 55.59(a)(2). Two out of six crews of licensed operators failed to demonstrate a satisfactory understanding of the required actions and mitigating strategies required to safely operate the facility under normal, abnormal, and emergency conditions. The finding was more than minor because the performance deficiency potentially affects the Human Performance attribute of the MS Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.
Specifically, the finding reflected the potential inability of the crews to take appropriate SR action in response to actual abnormal and emergency conditions. The perceived risk associated with the number of crews failing the annual operating test is provided in the Simulator Operational Evaluation matrix of IMC 0609, Appendix I, Licensed Operator Requalification SDP. The finding was of very low safety significance (Green) because only two of six of the operating crews failed; the failed operating crews were remediated (i.e., the operating crews were re-trained and successfully re-tested) prior to returning to shift; and because there was not a finding associated with operating crew failures during calendar year 2010. The cause of this finding was directly related to the cross-cutting aspect of personnel training and qualifications in the area of Human Performance - Resources, in that the licensee failed to ensure the adequacy of the training provided to operators to assure nuclear safety. (H.2(b))
Inspection Report# : 2011005 (pdf)
Significance:      Dec 31, 2011 Identified By: NRC Item Type: FIN Finding Inadequate Operability Determination Of Control Room Air Conditioning System Components A finding of very low safety significance was identified by the inspectors for the failure to adequately assess operability of the Control Room Air Conditioning Component (ACC) dampers, ACC-15 and ACC-16, in Operability Determination (OD) 456, Revision 0, ACC-15 and ACC-16 QA Classification, in accordance with site Procedure OP AA 102 1001, Development of Technical Basis to Support Operability Determinations, Revision 4. The licensee entered the issue into their CAP and was completing an apparent cause evaluation at the conclusion of the inspection period.
The finding was determined to be more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated December 24, 2009, because the finding, if left uncorrected, had the potential to become a more significant safety concern. Specifically, the failure to give the operators written instructions to manually reposition the SR dampers could have lead to situations where the operators would not have been able to rapidly and correctly manually reposition the SR dampers to perform their required safety functions necessary to mitigate design basis accidents. The inspectors determined the finding could be evaluated using IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings, Table 4a, for the MS Cornerstone, dated January 10, 2008. The inspectors answered "No" to the MS questions and screened the finding as having very low significance (Green). The finding has a cross-cutting aspect in the area of Human Performance - Decision Making, because the licensee failed to communicate decisions and the
 
bases for decisions to personnel who had a need to know the information in order to perform work safely, in a timely manner. Specifically, the licensee failed to communicate in a timely manner to the reactor operators the written instructions in the standing order necessary to manually reposition the dampers to their SR positions after a design basis accident. (H.1(c))
Inspection Report# : 2011005 (pdf)
Significance:        Dec 31, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Unanticipated Closure Of Emergency Diesel Generator B Output Breaker A finding of very low safety significance and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed for the failure to implement a procedure for an activity affecting quality. Procedure OP-KW-OSP-DGE-003B, Diesel Generator B Semi-Annual, required electrical maintenance personnel to check only the voltage of the emergency diesel generator (EDG) B output breaker Relay 52C/1-603; however, the electricians checked voltage and then attempted to check resistance of the relay. Specifically, after successfully testing for voltage, an electrician then selected a resistance setting for the volt-ohm meter (VOM) in an attempt to perform a continuity check of the relay, which was not prescribed by the procedure. The electricians actions resulted in the closure of the EDG output Breaker 1 603, and EDG B was paralleled to the grid out-of-phase.
The licensee initiated a condition report and took remedial corrective actions that included additional testing and inspections of EDG B to ensure that no damage occurred to the equipment as a result of the system transient, followed by the successful completion of post maintenance testing. At the end of the inspection period, the licensee was performing a root cause evaluation to determine the cause of the event and to develop additional corrective actions related to the organizational performance issues.
The inspectors determined that the finding was more than minor in accordance with IMC 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," because the finding was associated with the MS Cornerstone attribute of Equipment Performance, and adversely impacted the objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the attribute of equipment performance impacted the availability and reliability of EDG B and could have resulted in the catastrophic failure of the generator. The inspectors determined that 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 MS Cornerstone, dated January 10, 2008. The inspectors answered "No" to the MS questions and screened the finding as having very low safety significance (Green). The inspectors determined that this finding has a cross -cutting aspect in the area of Human Performance - Work Practices, because the maintenance personnel and supervision failed to communicate and ensure human error prevention techniques were used, such as holding formal pre job briefings, and self and peer checking. The licensee also failed to ensure that these techniques were used commensurate with the potential risk of the assigned task, such that work activities were performed safely. Finally, during these maintenance activities, the inspectors concluded that licensee personnel proceeded in the face of uncertainty. (H.4(a))
Inspection Report# : 2011005 (pdf)
Significance:        Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure To Maintain Fire Barrier And Automatic Fire Suppression A finding of very low safety significance and associated non-cited violation of license condition 2.C(3) of the Kewaunee Power Station Renewed Operating License was identified by inspectors for the failure to have a self-closing fire door that closed and latched each time it was open. License condition 2.C(3) requires, in part, that the licensee implement and maintain, in effect, all provisions of the approved fire protection program as described in the licensees fire plan. Appendix B of the Kewaunee Power Station Fire Protection Program Plan lists the 1975 edition of NFPA 80 [National Fire Protection Association], Fire Doors and Windows, as an applicable NFPA code. NFPA 80 states, in part, that a self closing door shall be equipped with a closing device to cause the door to close and latch each time it is opened. The licensee entered the issue into its corrective action program and adjusted the door closing device to ensure the door properly closed when the train A screenhouse ventilation fan was operating.
 
The inspectors determined that the failure of the door to close and latch was contrary to the requirements of NFPA 80 and was a performance deficiency. The finding was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of protection against external factors (Fire) and adversely 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). The inspectors screened the finding as having very low safety significance (Green). The 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 that the resolutions address causes and extent of conditions. This includes properly classifying, prioritizing, and evaluating for operability and reportability conditions adverse to quality.
Inspection Report# : 2011004 (pdf)
Significance:        Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Misapplication Of Technical Specification 3.1.6 Applicability Note A finding of very low safety significance and associated non-cited violation of Technical Specification (TS) 3.1.6, Control Bank Insertion Limits, was identified by the inspectors for the failure to comply with TS action condition 3.1.6.A due to incorrect use of the applicability note. Specifically, on August 30, 2011, during the performance of SP-49-075, Control Rod Exercise, operators received a rod control urgent failure while inserting control bank A group 1 control rods. The test was suspended for troubleshooting for approximately 20 hours with control bank A group 2 control rods, inserted one step below the control rod insertion limit in violation of TS 3.1.6.A action condition. The inspectors concluded that, once the test was suspended for troubleshooting activities, use of the applicability note was not appropriate; therefore, the operators should have complied with the TS 3.1.6.A action condition for control bank A group 2 control rods at that time. On August 31, operators withdrew control bank A group 2 rods one step, which restored the rods to within the limit specified in the core operating limits report. At the end of this inspection period, the licensee was still performing an apparent cause evaluation to determine the causes of the event and to develop corrective actions.
The finding was determined to be more than minor because the finding adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent desirable consequences. Specifically, the human performance attributes of the licensees failure to recognize the misapplication of the applicability note of the TS affected the capability of systems that respond to initiating events. The inspectors screened the finding as having very low safety significance (Green) because an actual loss of safety function did not occur. The finding has a cross-cutting aspect in the area of human performance, decision making, because the licensee failed to use conservative assumptions and adopt a requirement to demonstrate that the proposed action is safe in order to proceed rather than a requirement to demonstrate that it is unsafe in order to disapprove the action.
Inspection Report# : 2011004 (pdf)
Significance:        Sep 30, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Incorrect Transformer Load Tap Changer Setting Causes Inoperable Offsite Power A finding of very low safety significance and associated non-cited violation of Technical Specification 3.8.1 was self revealed for the failure to maintain a switchyard transformer load tap changer (LTC) at the appropriate setting for the predicted post trip voltage of offsite power. The incorrect setting resulted in the inoperability of the Reserve Auxiliary Transformer (RAT) offsite power source. The licensees corrective actions included restoring the RAT supply transformer (RST) LTC to an appropriate setting, creating a short term standing order to prevent operation of the RST LTC outside settings that were supported by the existing interface agreement with the transmission system operator.
The licensee performed an apparent cause evaluation, a root cause analysis and also, as a long-term corrective action, modified procedure OP-KW-NOP SUB 003 to prevent operation of the RST LTC outside settings that were supported by the existing interface agreement with the transmission system operator.
The finding was determined to be more than minor because the finding was associated with the Mitigating Systems Cornerstone attribute of configuration control and adversely 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). The inspectors screened the finding as having very low safety significance (Green). This finding has a cross-cutting aspect in the area of human performance, resources, because the licensee did not maintain complete, accurate, and up to date procedures for the use of the RST LTC following its installation during the spring 2011 outage.
Inspection Report# : 2011004 (pdf)
Significance:      Jun 30, 2011 Identified By: Self-Revealing Item Type: FIN Finding Technical Support Center Diesel Generator Output Breaker Fails To Close A finding of very low safety significance was self revealed for the failure to perform adequate preventive maintenance on latching relay VR1/B46, a relay required for closure of the technical support center (TSC) diesel generators (DG's) output breaker and automatic restoration of bus 1-46, which powers the TSC DGs cooling system. Specifically, on March 20, 2011, during a partial loss of offsite power event, the TSC DG started but failed to load onto bus 1-46.
After approximately 43 minutes of operation, the DG automatically shut down from an over-temperature condition, as designed. The licensee initiated condition report 417289 and performed apparent cause evaluation 018573. The licensees short-term corrective actions included troubleshooting the initial failure, repairing relay VR1/B46, and restoring the TSC DG to functional status. The licensees long-term corrective actions were in-progress at the completion of this inspection period.
The finding 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.
Specifically, the failure of the output breaker to close and energize bus 1-46 caused the TSC DG to overheat and automatically shut down during a partial loss of offsite power. The inspectors concluded the finding could be evaluated in accordance with Inspection Manual Chapter 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Table 4a, for the Mitigating Systems Cornerstone. The inspectors answered Yes to questions 2 and 4 of the Mitigating Systems Cornerstone column and determined that the finding required a Phase 2 analysis. The Region III senior reactor analyst completed a Phase 2 analysis and determined the risk significance of the issue to be very low (Green). The finding has a cross-cutting aspect in the area of human performance, resources, because a licensee effort to review various plant components for possible inclusion in a preventive maintenance optimization project had assigned a low priority to this relay (H.2(a)).
Inspection Report# : 2011003 (pdf)
Significance:      Jun 30, 2011 Identified By: NRC Item Type: FIN Finding Inadequate Operability Determination Of A Heat Exchanger Leak On Emergency Diesel Generator A A finding of very low safety significance was identified by the inspectors for the failure to adequately assess operability of the service water system in operability determination 413, EDG A Jacket Water Expansion Tank Overflow, in accordance with site Procedure OP-AA-102-1001, Development of Technical Basis to Support Operability Determinations. At the end of the inspection period, the licensee was completing an apparent cause evaluation to determine the cause and develop corrective actions.
The finding was determined to be more than minor because the finding, if left uncorrected, had the potential to become a more significant safety concern. Specifically, the failure to perform operability evaluations on degraded safety-related systems could lead to situations where systems needed to mitigate design basis accidents were not capable of performing their required safety functions. The inspectors determined the finding could be evaluated using Inspection Manual Chapter 0609, Significance Determination Process, Attachment 0609.04, Phase 1- Initial Screening and Characterization of Findings, Table 4a, for the Mitigating Systems Cornerstone. The inspectors answered "No" to the Mitigating Systems questions and screened the finding as having very low safety significance (Green). The finding has a cross-cutting aspect in the area of human performance, work practices, because the licensee failed to communicate decisions and the bases for decisions to personnel who had a need to know the information in order to perform work safely. Specifically, the licensee failed to effectively communicate the expectation to assess operability of the service water system in the pre-job brief and peer review (H.1(c)).
 
Inspection Report# : 2011003 (pdf)
Significance:      Jun 30, 2011 Identified By: NRC Item Type: FIN Finding Failure To Review And Update Severe Accident Management Guidelines In Accordance With An Established Program A finding of very low safety significance was identified by the inspectors for the licensees failure to perform reviews and update the Severe Accident Management Guidelines (SAMGs) in accordance with the licensees nuclear administrative directives (NADs). Specifically, Procedure NAD 14.06 required that the engineering group review industry correspondence related to SAMGs and implement appropriate changes, and that the emergency preparedness group conduct biennial reviews of the SAMGs. The inspectors identified that neither group had performed the reviews. As a result, the SAMGs were not adequately updated. The licensee entered this issue into their corrective action program as condition reports 424681, 424855, 424865, 424866, 425092, 426999, and 427092, and was still evaluating the cause for this condition at the end of this inspection period. The licensee scheduled the revision of the SAMGs for completion by December 2011.
The finding was determined to be more than minor because, if left uncorrected, the finding had the potential to lead to a more significant safety concern. Specifically, the failure to review and update the SAMGs would have hampered the licensees response in the unlikely event of a severe accident, because the SAMGs were not current. The inspectors, in consultation with the Region III senior reactor analyst, determined that the finding could be evaluated using the Significance Determination Process in accordance with Inspection Manual Chapter 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Table 4a, for the Mitigating Systems Cornerstone. The inspectors answered "No" to the Mitigating Systems questions and screened the finding as having very low safety significance (Green). The finding has a cross-cutting aspect in the area of problem identification and resolution, corrective action program, because the licensee failed to take appropriate corrective actions to address safety issues and adverse trends in a timely manner, commensurate with their safety significance and complexity. Specifically, the licensee identified in an apparent cause evaluation initiated in April 2010 that the emergency preparedness organization had not performed the required reviews and updates of emergency preparedness procedures, and the SAMGs were identified in the licensees extent of condition. However, the inspectors identified that the corrective actions issued for this extent of condition did not address the SAMGs and, therefore, no corrective actions were taken (P.1(d)).
Inspection Report# : 2011003 (pdf)
Barrier Integrity Significance:      Mar 31, 2012 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Post-Maintenance Test Of Motor Replacements The inspectors identified a finding of very low safety significance and associated non-cited violation (NCV) of Technical Specification (TS) 5.4.1, Procedures, which required, in part, that written procedures shall be implemented covering the applicable procedures recommended in Regulatory Guide (RG) 1.33, Revision 2, Appendix A. Specifically, Procedure GNP 08.02.12, Post-Maintenance Testing/Operations Retest, stated, in part, that the post-maintenance tests (PMTs) were performed upon completion of maintenance activities, and demonstrated that the identified deficiency was repaired, and that no new deficiency was created. On July 4, 2011, the licensee replaced the spent fuel pool (SFP) pump motor B, and failed to conduct an adequate PMT, which demonstrated no new deficiency was created. The PMT only tested the replaced motor and failed to include testing of the pump to ensure that no new deficiency was created. The licensee entered the issue into its corrective action program (CAP) as condition reports (CRs) 464645, 466183, and 466215, and planned to perform an apparent cause evaluation (ACE) and take corrective actions.
The inspectors determined that the finding was more than minor in accordance with IMC 0612, "Power Reactor
 
Inspection Reports," Appendix B, "Issue Screening," because, if left uncorrected, the failure to perform adequate PMT on motor replacements would have the potential to lead to a more significant safety concern. The inspectors determined that 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, dated January 10, 2008. The inspectors answered "No" to the Reactor Coolant System or Fuel Barrier Questions related to Spent Fuel Pool Issues, and screened the finding as having very low safety significance (Green). The inspectors also determined that this finding had a cross-cutting aspect in the area of human performance, resources, because the licensee did not ensure the PMT procedure guidance related to motor replacements was adequate and accurate to assure nuclear safety. (H.2(c))
Inspection Report# : 2012002 (pdf)
Significance:      Mar 31, 2012 Identified By: NRC Item Type: NCV NonCited Violation Inadequate Procedure For Technical Specification Surveillance The inspectors identified a finding of very low safety significance (Green) and an associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to have appropriate procedures to complete TS required surveillances. Specifically, OSP CCI 004, Containment Isolation Valve Verification, did not contain adequate steps to complete a TS required airlock door check and the procedure did not include six manual containment isolation valves (CIVs) that should have been included in the procedure for position verification. The licensee corrected the procedure and entered the issue into its corrective action program as condition reports (CRs) 464355, 464494, and 467560, and planned to perform an apparent cause evaluation (ACE).
The inspectors determined that the finding was more than minor in accordance with IMC 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," because the finding was associated with the Barrier Integrity Cornerstone attribute of procedure quality 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, inspectors found seven examples in OSP-CCI-004 where either the procedure steps were not adequate or CIVs were missing that should have been included in the procedure for position verification. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, 609.04, Phase 1- Initial Screening and Characterization of Findings, Table 4a for the Barrier Integrity Cornerstone, dated January 10, 2008. The inspectors answered No to the Containment Barrier questions and screened the finding as having very low safety significance (Green). This finding has a cross-cutting aspect in the area of human performance, resources, because the licensee did not ensure supervisory and management oversight of work activities, including contractors, such that nuclear safety was supported. Specifically, OSP-CCI-004 did not get an approval review during the procedure review process and the supervisory review that was conducted did not identify the procedural errors. (H.4(c))
Inspection Report# : 2012002 (pdf)
Significance:      Jun 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failed Standoffs Result In An Inoperable Train of Shield Building Ventilation A finding of very low safety significance and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by inspectors for the failure to have and follow adequate procedures for evaluation and installation of components in shield building ventilation (SBV) train A. Specifically, the licensee failed to have adequate procedures to direct the completion of a subcomponent classification evaluation (SCE) and prevent non safety-related parts from being installed in safety-related applications; have torque specifications for the standoffs (spacers for circuit cards) in the work instructions; and properly accomplish the SCE procedure when evaluating the standoffs. The licensees initial short-term corrective actions removed the installed standoffs from both trains. The licensee also performed an extent of condition looking at previously completed item equivalency evaluations to determine if an SCE was needed or missing for newly installed components.
The finding was determined to be more than minor because the finding was associated with the Barrier Integrity Cornerstone attribute of procedure quality, 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 licensee failed to have and follow adequate procedures which led to the failure of SBV train A. The inspectors determined that this was a type B containment finding since it was related to a degraded condition that had potential important implications for the integrity of the containment, without affecting the likelihood of core damage.
The inspector evaluated the finding using the significance determination process (SDP) in accordance with Inspection Manual Chapter 0609, Appendix H, Containment Integrity SDP, Table 4.1, and determined that the finding did not relate to a containment structure, system, and component, nor containment status that had an impact on large early release frequency. Because of this, the issue screened as Green, using the flowchart in Figure 4.1. The finding has a cross-cutting aspect in the area of problem identification and resolution, corrective action program, because the licensee failed to thoroughly evaluate problems such that the resolutions address causes and extent of conditions, as necessary. This includes properly classifying, prioritizing, and evaluating for operability and reportability conditions adverse to quality. This also includes, for significant problems, conducting effectiveness reviews of corrective actions to ensure that the problems are resolved. Specifically, the licensee failed to properly evaluate and identify the cause of the SBV train A failure and produce a resolution that addressed the cause (P.1(c)).
Inspection Report# : 2011003 (pdf)
Emergency Preparedness Significance:        Dec 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Incorrect Assumptions Used In The Development Of Emergency Action Level Thresholds A finding of very low safety significance and associated non-cited violation of 10 CFR 50.54(q) was identified by the NRC for failing to maintain emergency plans that meet the requirements of emergency planning standard 10 CFR 50.47(b)(4). The inspectors determined that a performance deficiency existed in that incorrect assumptions were used in the development of Emergency Action Level (EAL) thresholds associated with containment gas (R12) and containment ventilation (R21) radiation monitors. The licensee entered this issue into its CAP as CR356229 and corrected the errant EAL thresholds in its emergency classification and action level scheme.
This finding was determined to be more than minor because the deficiency, if left uncorrected, could have the potential to lead to a more significant safety concern. Specifically, in the event of a radiological emergency, the deficiency has the potential to increase the risk to the public through a premature and/or unnecessary general emergency declaration and subsequent protective action recommendation of evacuation. This finding was evaluated using IMC 0609, Appendix B, Emergency Preparedness Significance Determination Process, Sheet 1, Failure To Comply. This finding is associated with a failure to meet or implement a regulatory requirement. The deficiency is not greater than Green because it did not result in the Risk-Significant Planning Standard Function being lost or degraded. No cross-cutting aspect is assigned to this finding because it is not indicative of current plant performance.
Inspection Report# : 2011005 (pdf)
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.
Miscellaneous Significance: N/A Mar 31, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure To Submit LER Per 10 CFR 50.73 The inspectors identified a Severity Level (SL) IV NCV of 10 CFR 50.73(a)(2)(vii) for the failure of the licensee to report an event where a single cause or condition caused two independent trains to become inoperable in a single system designed to control the release of radioactive material. Specifically, the licensee failed to report that both trains of shield building ventilation (SBV) were inoperable due to a single cause, because both trains contained unqualified control card standoffs that were needed to maintain the seismic qualification and operability of the system. The licensee entered this into their CAP as CR429469, planned to perform an ACE, and was drafting an update to Licensee Event Report (LER) 05000305/2011-005.
The inspectors determined that the failure to report the event in accordance with 10 CFR 50.73 was a performance deficiency. Because violations of 10 CFR 50.73 are considered to be violations that potentially impact the regulatory process, they are dispositioned using the traditional enforcement process instead of the Reactor Oversight Process (ROP) SDP. Because the performance deficiency, a failure to report, was not an ROP finding per IMC 0612, Appendix B, Issue Screening, a cross-cutting aspect was not assigned to this violation. Per the NRC Enforcement Policy, Section 6.0, Violation Examples, a failure to submit a required LER is categorized as an SL IV violation.
Inspection Report# : 2012002 (pdf)
Significance: SL-IV Jun 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure To Submit LER Per 10 CFR 50.73 A Severity Level IV non-cited violation of 10 CFR Part 50.73(a)(2)(i)(B) and 50.73(a)(2)(v)(C) was identified by the inspectors for the failure of the licensee to report an event or condition that was prohibited by Technical Specifications, and an event or condition that could have prevented the fulfillment of a safety function, that is relied upon to control the release of radioactive material. Specifically, the licensee failed to report the inoperability of shield building ventilation train A from December 3, 2010, through January 26, 2011, a condition prohibited by Technical Specification 3.6.c.1, which allowed a single train outage time of seven days. Additionally, shield building ventilation train B was inoperable on multiple occasions during the same time period, requiring the licensee to also report an event or condition that could have prevented the fulfillment of a safety function, which is relied upon to control the release of radioactive material. At the end of the inspection period, the licensee was completing an apparent cause evaluation to determine the cause and develop corrective actions.
Because violations of 10 CFR 50.73 are considered to be violations that potentially impact the regulatory process, they are dispositioned using the traditional enforcement process instead of the Reactor Oversight Process Significance Determination Process. A cross-cutting aspect was not assigned to this violation. Per the NRC Enforcement Policy, Section 6.0, Violation Examples, a failure to submit a required licensee event report is categorized as a Severity Level IV violation.
Inspection Report# : 2011003 (pdf)
Last modified : May 29, 2012
 
Kewaunee 2Q/2012 Plant Inspection Findings Initiating Events Significance:        Dec 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Procedures For Reduced Inventory Operations Were Not Appropriate To Preclude Air Entrainment The inspectors identified a 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 failure to establish procedures for reduced inventory operations that were appropriate to manage gas accumulation. Specifically, the procedures did not preclude air entrainment into the residual heat removal (RHR) and reactor coolant systems (RCSs). This finding was entered into the licensees corrective action program. The licensee's immediate corrective actions included calculating the instrument inaccuracies for RHR flow and refueling level instrument loops, referencing the level inaccuracies based on inactive flow in RCS loops in the associated procedures., evaluating levels, and updating the procedures with a new graph.
The performance deficiency was determined to be more than minor because it was associated with the Initiating Event Cornerstone attribute of procedure quality, and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown operations. Specifically, the failure to establish procedures for reduced inventory operations that were appropriate to preclude air entrainment did not limit the likelihood of events that result from adverse air entrainment into the RHR and RCSs. The finding screened as having very low safety significance (Green) because the Region III Senior Reactor Analysts determined that it reasonably met the safety functions of core heat removal, RCS inventory control, power availability, containment control, and reactivity control; and there had been no actual air entrainment problems that had occurred using the procedures. The finding has a cross-cutting aspect in the area of Problem Identification and Resolution because the licensee did not thoroughly evaluate relevant operating experience. Specifically, the licensees evaluation of gas related issues in response to NRC Generic Letter (GL) 2008 01 was deficient in that it did not consider vortexing during reduced inventory operations. (P.2(a))
Inspection Report# : 2011005 (pdf)
Significance:        Oct 07, 2011 Identified By: NRC Item Type: NCV NonCited Violation Flammable Gas Bottles Installed and/or Stored in the Auxiliary Building The inspectors identified a finding of very low safety significance and associated NCV of Title 10, Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion III, Design Control, for the failure to check the adequacy of design for flammable gas bottles installed and/or stored in fire areas and fire zones located within the auxiliary building and their impact on safe shutdown cables, safety-related cables and safety-related equipment. Specifically, the licensee failed to evaluate how a failure of the flammable gas bottles and a resulting fire or explosion at the installed and/or stored locations could impact nearby safety-related structures, systems, or components. The licensee entered this issue into their corrective action program to review the placement of the flammable gas bottles.
The inspectors determined that the finding was more than minor because the finding was associated with the Initiating Events cornerstone attribute of Protection against External Factors (Fire) and affected the cornerstones objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The finding was of very low safety significance due to the low fire initiating frequency and the availability of remaining mitigating systems. This finding did not have a cross-cutting aspect because the finding was not representative of current performance.
Inspection Report# : 2011008 (pdf)
 
Significance:      Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Repetitive Molded Case Circuit Breaker Failures A finding of very low significance and associated non-cited violation of Title 10 of the Code of Federal Regulations (CFR) 50.65(a)(3) was identified by the inspectors for the failure to incorporate industry operating experience into preventive maintenance activities when practical to do so. Specifically, the failure to incorporate the industry operating experience resulted in multiple molded case circuit breaker (MCCB) failures that could have been prevented by implementing an MCCB cycling program. The need to cycle MCCBs was identified in industry operating experience as well as the vendor's instructions for the breakers. The licensee was performing an apparent cause evaluation which was still in progress at the conclusion of the inspection period. Initial corrective actions included scheduling the MCCBs for the breaker cycling maintenance activity.
This finding was determined to be of greater than minor significance because it was associated with the Protection Against External Factors attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events, such as fire, that challenge critical safety functions during shutdown as well as power operations. Specifically, the lack of a cycling program for safety related MCCBs resulted in breakers remaining in the on position after an overcurrent condition. The inspectors determined the finding had very low safety significance (Green) because the breakers and associated cabling did not significantly affect safe shutdown defense in depth strategies and the finding did not involve a design or qualification deficiency, did not represent a loss of system safety function, did not represent a loss of Technical Specification equipment for greater than its allowed outage time, and did not affect risk significant equipment per 10 CFR 50.65. This finding has a cross-cutting aspect in the area of human performance, work control, because the licensee did not emphasize the need for work groups to communicate, coordinate, and cooperate with each other during activities in which interdepartmental coordination is necessary to assure plant and human performance.
Inspection Report# : 2011004 (pdf)
Mitigating Systems Significance:      Jun 30, 2012 Identified By: Self-Revealing Item Type: NCV NonCited Violation Failure To Utilize Work Order For Temporary Weld Repair On ASME Code, Class 2 Piping A finding of very low safety significance and associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed for the failure to accomplish Temporary Modification (TMOD) 2012-11 in accordance with Work Order (WO) KW100894696 and the associated weld data sheet and map. Specifically, licensee personnel failed to utilize the WO instructions, weld data sheet and weld map when welding a temporary NRC-approved clamp on American Society of Mechanical Engineers (ASME) Code Class 2 residual heat removal (RHR) piping. The failure to use the required documentation to perform the work resulted in the worker creating a second through wall leak on the ASME Code, Class 2 RHR piping upstream of valve RHR 600.
The licensee entered the issue into its corrective action program (CAP) as condition report (CR) 472915 and permanently corrected both through wall leaks on the RHR system piping following the approval of a second proposed alternative, without incident on May 5, 2012. At the end of the inspection period, the licensee continued to perform an apparent cause evaluation (ACE) to determine the causes for the organizational failures that occurred.
The finding was determined to be more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated December 24, 2009, because the finding was associated with the Mitigating Systems Cornerstone attribute of human error (pre-event) and adversely affected the cornerstone objective to ensure the reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). The inspectors determined that the finding could be evaluated in accordance with IMC 0609, Appendix G, Shutdown Operations SDP, dated February 28, 2005. The inspectors used Checklist 1, PWR Hot Shutdown Operation: Time to Core Boiling <2 Hours, contained in Attachment 1 and
 
determined that the finding affected core heat removal guidelines I.B(1), Procedures, and I.C(2), Equipment. The inspectors screened the finding as very low safety significance (Green) because it did not degrade the licensees ability to establish an alternate core cooling path if decay heat removal could not be re established and, therefore, did not require a phase 2 or phase 3 analysis. This finding has a cross-cutting aspect in the area of human performance, resources, because the licensee did not ensure supervisory and management oversight of work activities, including contractors, such that nuclear safety was supported. Specifically, the inspectors identified that the pre-job brief conducted by supervision and management for this work did not include a review of the WO, weld sheet, or weld map and did not convey accurate information regarding the significance of the activity, the type of weld to be performed and the system conditions where the weld was performed. (Section 1R18)
Inspection Report# : 2012003 (pdf)
Significance:      Jun 30, 2012 Identified By: NRC Item Type: NCV NonCited Violation Failure To Provide Adequate Suppression In Cable Spreading Area The inspectors identified a finding of very low safety significance (Green) and associated non-cited violation (NCV) of 10 CFR Part 50, Appendix R, Section III.G.3, for the licensees failure to provide adequate fire suppression coverage for fire zone AX-32. Specifically, the licensee failed to provide required fire suppression coverage for safe shutdown functions of source range monitoring, isolation of a steam generator (SG) blowdown line, and pressurizer level instrumentation in the cable spreading area. The licensee entered the issue into the CAP, designated manual backup from hose stations, and implemented an hourly fire watch for the radiation protection office (RP) in fire zone AX-32.
The inspectors determined that the finding was more than minor because the failure to provide suppression for redundant trains of safe shutdown equipment increased the likelihood that alternative shutdown methods would have to be used in the event of a fire. The finding was of very low safety significance based on a Phase 3 significance determination analysis. The finding has a cross-cutting aspect in the area of problem identification, corrective action program, because the licensee did not take appropriate corrective actions to address the inadequate suppression system in fire zone AX 32. (Section 4OA2.4)
Inspection Report# : 2012003 (pdf)
Significance:      Dec 31, 2011 Identified By: NRC Item Type: NCV NonCited Violation Unanalyzed Flood Source From Technical Support Center Building A finding of very low safety significance and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors for the failure to identify and analyze a potential flooding source that was within the Kewaunee licensing basis. Specifically, during the internal flood basis reconstitution in 2005, the licensee failed to realize and assess the potential for fire main piping in the technical support center (TSC) building to be ruptured during a tornado or seismic event. Water from a ruptured fire main had the potential to accumulate in the basement of the TSC building, flow into the attached auxiliary building, and potentially affect safety related (SR) equipment. The licensee initiated a condition report (CR) and completed calculations and analyses to demonstrate the existing barriers, although not credited at the time, were adequate to support this internal flood scenario. In addition, the licensee performed an extent of condition analysis to determine if any additional internal flood scenarios were missed.
The inspectors determined that the finding was more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated December 24, 2009, because it was associated with the Mitigating Systems (MS) 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 failure to analyze all potentially credible internal flood sources could affect the availability of SR systems. The inspectors determined that the finding could be evaluated using the significance determination process in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings, Table 4a, for the MS Cornerstone, dated January 10,
 
2008. The inspectors answered "No" to the MS questions and screened the finding as having very low safety significance (Green). The inspectors did not identify a cross-cutting aspect associated with this finding because the internal flood design basis reconstitution occurred in 2005 and the inspectors determined that there was not an opportunity to identify this deficiency in the past three years.
Inspection Report# : 2011005 (pdf)
Significance:      Dec 31, 2011 Identified By: Self-Revealing Item Type: FIN Finding Two Of Six Operating Crew Failures On The Simulator Operational Evaluation Portion Of The 2011 Annual Requalification Operating Test A self-revealed finding associated with operating crew performance on the simulator during a licensee-administered requalification examination was identified. Two of the six crews evaluated during the annual operating tests failed to pass their simulator examinations. As immediate corrective action, the failed operating crews were remediated (i.e.,
the operating crews were re-trained and successfully re-tested) prior to returning to shift. The licensee entered this issue into the CAP as CR456328.
The inspectors determined that the crew failures constituted a performance deficiency based on the fact that licensed operators are expected to operate the plant with acceptable standards of knowledge and abilities demonstrated through periodic testing as required by 10 CFR 55.59(a)(2). Two out of six crews of licensed operators failed to demonstrate a satisfactory understanding of the required actions and mitigating strategies required to safely operate the facility under normal, abnormal, and emergency conditions. The finding was more than minor because the performance deficiency potentially affects the Human Performance attribute of the MS Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.
Specifically, the finding reflected the potential inability of the crews to take appropriate SR action in response to actual abnormal and emergency conditions. The perceived risk associated with the number of crews failing the annual operating test is provided in the Simulator Operational Evaluation matrix of IMC 0609, Appendix I, Licensed Operator Requalification SDP. The finding was of very low safety significance (Green) because only two of six of the operating crews failed; the failed operating crews were remediated (i.e., the operating crews were re-trained and successfully re-tested) prior to returning to shift; and because there was not a finding associated with operating crew failures during calendar year 2010. The cause of this finding was directly related to the cross-cutting aspect of personnel training and qualifications in the area of Human Performance - Resources, in that the licensee failed to ensure the adequacy of the training provided to operators to assure nuclear safety. (H.2(b))
Inspection Report# : 2011005 (pdf)
Significance:      Dec 31, 2011 Identified By: NRC Item Type: FIN Finding Inadequate Operability Determination Of Control Room Air Conditioning System Components A finding of very low safety significance was identified by the inspectors for the failure to adequately assess operability of the Control Room Air Conditioning Component (ACC) dampers, ACC-15 and ACC-16, in Operability Determination (OD) 456, Revision 0, ACC-15 and ACC-16 QA Classification, in accordance with site Procedure OP AA 102 1001, Development of Technical Basis to Support Operability Determinations, Revision 4. The licensee entered the issue into their CAP and was completing an apparent cause evaluation at the conclusion of the inspection period.
The finding was determined to be more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated December 24, 2009, because the finding, if left uncorrected, had the potential to become a more significant safety concern. Specifically, the failure to give the operators written instructions to manually reposition the SR dampers could have lead to situations where the operators would not have been able to rapidly and correctly manually reposition the SR dampers to perform their required safety functions necessary to mitigate design basis accidents. The inspectors determined the finding could be evaluated using IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings, Table 4a, for the MS Cornerstone, dated January 10, 2008. The inspectors answered "No" to the MS
 
questions and screened the finding as having very low significance (Green). The finding has a cross-cutting aspect in the area of Human Performance - Decision Making, because the licensee failed to communicate decisions and the bases for decisions to personnel who had a need to know the information in order to perform work safely, in a timely manner. Specifically, the licensee failed to communicate in a timely manner to the reactor operators the written instructions in the standing order necessary to manually reposition the dampers to their SR positions after a design basis accident. (H.1(c))
Inspection Report# : 2011005 (pdf)
Significance:        Dec 31, 2011 Identified By: Self-Revealing Item Type: NCV NonCited Violation Unanticipated Closure Of Emergency Diesel Generator B Output Breaker A finding of very low safety significance and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed for the failure to implement a procedure for an activity affecting quality. Procedure OP-KW-OSP-DGE-003B, Diesel Generator B Semi-Annual, required electrical maintenance personnel to check only the voltage of the emergency diesel generator (EDG) B output breaker Relay 52C/1-603; however, the electricians checked voltage and then attempted to check resistance of the relay. Specifically, after successfully testing for voltage, an electrician then selected a resistance setting for the volt-ohm meter (VOM) in an attempt to perform a continuity check of the relay, which was not prescribed by the procedure. The electricians actions resulted in the closure of the EDG output Breaker 1 603, and EDG B was paralleled to the grid out-of-phase.
The licensee initiated a condition report and took remedial corrective actions that included additional testing and inspections of EDG B to ensure that no damage occurred to the equipment as a result of the system transient, followed by the successful completion of post maintenance testing. At the end of the inspection period, the licensee was performing a root cause evaluation to determine the cause of the event and to develop additional corrective actions related to the organizational performance issues.
The inspectors determined that the finding was more than minor in accordance with IMC 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," because the finding was associated with the MS Cornerstone attribute of Equipment Performance, and adversely impacted the objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the attribute of equipment performance impacted the availability and reliability of EDG B and could have resulted in the catastrophic failure of the generator. The inspectors determined that 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 MS Cornerstone, dated January 10, 2008. The inspectors answered "No" to the MS questions and screened the finding as having very low safety significance (Green). The inspectors determined that this finding has a cross -cutting aspect in the area of Human Performance - Work Practices, because the maintenance personnel and supervision failed to communicate and ensure human error prevention techniques were used, such as holding formal pre job briefings, and self and peer checking. The licensee also failed to ensure that these techniques were used commensurate with the potential risk of the assigned task, such that work activities were performed safely. Finally, during these maintenance activities, the inspectors concluded that licensee personnel proceeded in the face of uncertainty. (H.4(a))
Inspection Report# : 2011005 (pdf)
Significance:        Sep 30, 2011 Identified By: NRC Item Type: NCV NonCited Violation Failure To Maintain Fire Barrier And Automatic Fire Suppression A finding of very low safety significance and associated non-cited violation of license condition 2.C(3) of the Kewaunee Power Station Renewed Operating License was identified by inspectors for the failure to have a self-clo}}

Latest revision as of 13:53, 29 November 2024

2017 Q1-Q4 ROP Inspection Findings
ML20311A529
Person / Time
Site: Kewaunee Dominion icon.png
Issue date: 06/04/2013
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Office of Nuclear Reactor Regulation
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References
Download: ML20311A529 (758)
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