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05000285/FIN-2013008-272013Q2Fort CalhounContinuous Monitoring Capability of Post Accident Main Steam Radiation Monitor RM-064The team identified an unresolved item associated with post accident radiation monitor RM-064. Specifically, the team is concerned about the capability of the monitor to provide representative measurements due to the system configuration, and this could represent a failure to ensure continuous effluent monitoring of the main steam lines following a steam generator tube rupture accident. Following the Three Mile Island Accident in March 25, 1979, licensees were required to ensure all potential effluent release points from nuclear power plants were equipped with high range radiation monitors. In particular, NUREG-0737, Section II.F.1.1 requires in part; that for pressurized water reactors such as FCS, Unit 1, steam release points be monitored for noble gases, and that indication of the activity must be monitored and recorded continuously. In addition Section II.F.1.1 requires the monitors shall be capable of functioning both during and following an accident. System designs shall accommodate a design-basis release and then be capable of following decreasing concentrations of noble gasses. In addition, the monitoring system shall be capable of obtaining readings at least every 15 minutes during and following an accident. . The team identified an unresolved item associated with post accident radiation monitor RM-064. Specifically, the team is concerned about the capability of the monitor to provide representative measurements due to the system configuration, and this could represent a failure to ensure continuous effluent monitoring of the main steam lines following a steam generator tube rupture accident. By application dated March 9, 1984, the licensee requested an amendment to the stations technical specifications in response to the Commissions Generic Letter 83-37, NUREG-0737 Technical Specifications. The generic letter, which was issued in November 1, 1983, advised licensees to submit new technical specifications for NUREG-0737 items, including Section II.F.1.1, Noble Gas Effluent Monitors (II.F.1.1). The stations potential post-accident steam release points include the main steam relief valves, the atmospheric dump valve, and the steam driven AFW pumps steam turbine. To comply with the high range radiation monitoring requirements, the licensee installed noble gas effluent monitors including, radiation monitor RM-064. Per USAR Section 11.2.3.11, RM-064, the post-accident main steam line monitor, is an off-line monitor designed to measure the steam activity by sampling steam from the two steam headers via two isolation valves HCV-921 and HVC-922. The monitor is placed in service in the event of a steam generator tube rupture. The monitor is capable of sampling steam from both steam headers and the recorded data from this monitor can then be utilized to quantify effluents released through the main atmospheric dump valve, the main steam safety valves, and the AFW pump turbine. Radiation monitor RM-064 is located in the turbine building next to Room 81. - 223 - The team noted that the design basis accident analysis contained in USAR, Section 14.14, Steam Generator Tube Rupture Accident, required the licensee to assume a coincident reactor trip and a loss of off-site power. Due to the assumed simultaneous loss of off-site power with the reactor trip, the reactor is cooled down by releasing steam via the main steam safety valves and atmospheric dump valve, creating a direct release path to the environment. In addition, due to the loss of off-site power, the normal condenser off-gas radiation monitor becomes un-available due to the loss of condenser vacuum. This leaves radiation monitor RM-064 as the only monitor available to measure radioactivity in the main steam lines. The analysis assumes all activity released from the faulted steam generator ceases when it is isolated by plant operators 2 hours after the event. The design of the FCS main steam line monitor is provided in MR-FC-79-190C, Post Accident Main Steam Line High Range Radiation Monitor RM-064, Revision 0, dated June 4, 1982. The station has two 28 inch diameter headers leading to the main turbine. Each main steam line is provided with six main steam safety valves each having different lift set-points. The pipe connecting these valves is 2.5 inches in diameter. The pipe connecting to the atmospheric dump valve is 3 inches in diameter. The sample line to radiation monitor RM-064 is 3/8 inch in diameter. This line is located upstream of the main steam isolation valves, in Room 81 of the auxiliary building. The distance from the main steam header to the actual location of radiation monitor RM-064 (outside Room 81) is over sixty feet long, while the main steam safeties and steam dump valve, are within 12 feet away from the main steam headers. The team reviewed the USAR, main steam drawings, applicable calculations, and interviewed engineers and operators to identify the design basis requirements for radiation monitor RM-064 and to verify it was capable of performing its intended functions. On The team also determined that for the B steam generator header the location of the 3/8 inch sample line leading to radiation monitor RM-064 was installed downstream of three of the main steam safety valves, including the lowest lift set-point valve. For the A steam generator header, the 3/8 inch sample line was located downstream of two of the safety valves but upstream of the lowest lift set point relief. Due to the location of the sample lines being downstream of the safety valves, the difference in pipe sizing between the lines to the monitor (3/8 inch), the main steam safety valves (2.5 inch), and the atmospheric dump valve (3 inch) and the distance from the main steam header to the monitor, the team questioned how the licensee assured a representative measurement would be obtained during and after a steam generator tube rupture accident. The team informed the licensee of their concerns and the licensee initiated Condition Reports CR 2013-04442, 2013-05515, and 2013-06267, to capture these concerns in the CAP. February 27, 2013, the team performed a walkdown of radiation monitor RM-064 and the steam lines. Because radiation monitor RM-064 is normally isolated, the team questioned how long it would take operators to put the monitor in service, and how the licensee met the requirement of continuous monitoring. During subsequent evaluations the licensee determined that there was not an established time requirement for operators to put radiation monitor RM-064 in service. - 224 - The licensee performed a simulator dry run with licensed operators to estimate the time required to place the monitor in service. During this simulated event, it took operators approximately 23 minutes to put the monitor in service, thus indicating that there could be an unmonitored release to the environment for at least 23 minutes following a steam generator tube rupture accident. Regarding the representative sample concern, engineers determined that without a sophisticated computer model it could not be definitely shown that the degree of turbulent mixing in the steam lines is sufficient to equalize the concentrations of radioactive gasses and entrained particulates downstream of the main steam safety valves where the lines connecting to radiation monitor RM-064 were located. The licensee issued Condition Report CR 2013-10507 requesting a detailed calculation to address this concern. The team determined this condition has existed since the time radiation monitor RM-064 was installed in February 1983, until February 27, 2013, when the issue was identified by the team. An engineering technical evaluation was then performed under Condition Report CR 2013-04442, based on existing radiological analysis Calculation FC06820 used for the steam generator accident analysis (USAR 14.14). This technical evaluation removed many of the conservative assumptions included in Calculation FC06820. Based on this basic evaluation and using engineering judgment, the licensee determined that there would be sufficient mixing and adequate concentration to provide a representative radiation measurement. The team concluded that further review is necessary in order to properly evaluate and disposition this issue. This issue is identified as URI 05000285/2013008-27, Continuous Monitoring Capability of Post Accident Main Steam Radiation Monitor RM- 064.
05000285/FIN-2016007-012016Q1Fort CalhounLicensee-Identified ViolationTechnical Specification 5.8.1 requires in part, that procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2. That appendix states, in part, that maintenance that can affect the performance of safety-related equipment should be performed in accordance with written procedures appropriate to the circumstances. Contrary to the above, maintenance that can affect the performance of safety-related equipment was not performed in accordance with written procedures appropriate to the circumstances. Specifically, maintenance that can affect the performance of safetyrelated valves was not performed in accordance with a procedure that required the licensee to review all diagnostic test results for compliance with the setpoint criteria for all diagnostic tests performed. As described in CR-2012-01601-017, the licensee restored compliance by writing and implementing procedure ER-FC- 410-AD-SETPOINT, Air-Operated Valve Setpoint Control, Revision 0. This procedure requires, in part, that the licensee review all diagnostic test results for compliance with the setpoint criteria for all diagnostic tests performed. The licensees failure to complete maintenance that can affect the performance of safety-related valves in accordance with written procedures appropriate to the circumstances was a performance deficiency that is more-than-minor because it adversely affected the Procedure Quality attribute of the Mitigating Cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, this performance deficiency resulted in valve HCV-2987, High Pressure Safety Injection Alternate Header Isolation, being not able to fulfill its design safety function from February, 2013, through July, 2013. Using Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, dated June 19, 2012, the inspectors determined that the finding should be processed through Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated July 1, 2012. Using Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that the finding was not a design or qualification deficiency but represented a loss of train function for greater than the outage time allowed by Technical Specifications. Therefore, a Region IV senior reactor analyst performed a detailed risk evaluation in accordance with Manual Chapter 0609, Appendix A, Section 6.0, Detailed Risk Evaluation. The analyst determined that the condition of valve HCV-2987 inoperability would affect only the plants response to a large-break loss-of-coolant accident followed by the failure of the instrument air system. The analyst calculated the initiating-event frequency to be 2.63 x 10-10 /year. Also, the analyst determined that the finding did not affect external initiator risk and would not involve a significant increase in the risk of a large, early release of radiation. Therefore, this violation has very low (Green) safety significance.
05000293/FIN-2016011-012017Q1PilgrimFailure to Identify All Root Causes of a Significant Condition Adverse to QualityThe NRC team identified a Green non-cited violation of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion XVI, Corrective Action, because Entergy did not adequately determine all root causes associated with a significant condition adverse to quality related to the failure to identify, evaluate, and correct the A SRVs failure to open upon manual actuation during a plant cooldown on February 9, 2013. Specifically, Entergy did not establish adequate measures to assure that the cause of a significant condition adverse to quality, inadequate shift manager operability determination rigor and its associated causes, were adequately determined and corrective action taken to preclude repetition. Entergys immediate corrective actions included planning to conduct operations management face-to-face conversations with shift manager qualified individuals to reinforce the shift managers responsibility for operability and functionality determination accuracy and rigor. Entergy entered this issue into the corrective action program as CRPNP-2017-00363 and CR-PNP-2017-00828. The performance deficiency was more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Specifically, if left uncorrected, the performance deficiency could have the potential to result in repetition of a failure to identify, evaluate, and correct an SRVs failure to open or a similar significant condition adverse to quality. The NRC team evaluated the finding using Exhibit 2, Mitigating Systems Screening Questions, of IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, and determined this finding did not affect the design or qualification of a mitigating structure, system, or component; represent a loss of system and/or function; involve an actual loss of function of at least a single train or two separate safety systems for greater than its technical specification-allowed outage time; or represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant. Therefore, the NRC team determined the finding was of very low safety significance (Green). The NRC team determined that the finding had a cross-cutting aspect in the area of Human Performance, Avoid Complacency, because individuals did not recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Specifically, Entergy incorrectly assumed that CR-PNP-2013-00825 contained inadequate information to determine that the A SRV had not opened, and this assumption ultimately impacted the root cause results documented in CR-PNP-2016-01621 (H.12).
05000293/FIN-2016011-022017Q1PilgrimFailure to Establish Corrective Actions to Preclude Repetition of a Significant Condition Adverse to QualityThe NRC team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, because Entergy did not implement CAPRs for a significant condition adverse to quality identified in root cause evaluation CR-PNP-2016-00716, Implementation of the Corrective Action Program, Revision 2. Specifically, the team identified that CAPRs for Entergys continued weaknesses in the implementation of the corrective action program were inadequate. Entergy entered this issue into their corrective action program for further evaluation as CR-PNP-2017-00053, CR-PNP-2017-00410, and CR-PNP-2017-01134. The performance deficiency was more than minor because if left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, the failure to preclude repetition of this significant condition adverse to quality could result in continuing weaknesses in implementation of the corrective action program, which was designated as a fundamental problem, and thus a contributing factor for PNPS Column 4 performance. Additionally, weaknesses with corrective action program implementation could result in equipment issues where operability is not maintained. The NRC team evaluated the finding using Exhibit 2, Mitigating Systems Screening Questions, of IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, and determined this finding did not affect the design or qualification of a mitigating structure, system, or component; represent a loss of system and/or function; involve an actual loss of function of at least a single train or two separate safety systems for greater than its technical specificationallowed outage time; or represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant. Therefore, the NRC team determined the finding was of very low safety significance (Green). The NRC team determined that the finding had a cross-cutting aspect in the area of Human Performance, Procedure Adherence, because individuals did not follow processes, procedures, and work instructions. Specifically, Entergy did not follow procedure EN-LI-102, which provides the station standards for crafting a corrective action and states, in part, that the corrective action descriptions must be worded to ensure that the adverse condition or cause/factor is addressed (H.8).
05000293/FIN-2016011-032017Q1PilgrimFailure to Issue Appropriate Corrective Actions to Preclude Repetition for the Causes of the September 2016 ScramThe NRC team identified a Green finding because Entergy did not issue appropriate CAPRs in accordance with Entergy procedure EN-LI-102, Corrective Action Process, Revision 28. Specifically, Entergy did not issue adequate CAPRs associated with Root Cause 1 of the feedwater regulating valve failure in September 2016 that resulted in a manual scram. As a result of the NRC teams questions, Entergy issued procedure 1.13.2, Vendor and Technical Information Reviews, Revision 0, as continuous use to ensure that planners will always have the checklist in-hand when planning work to ensure that appropriate vendor technical information is always included in applicable work instructions. Entergy entered the NRC teams concerns in the corrective action program as CR-PNP-2017-00687 and CR-PNP-2017-00936. The performance deficiency was more than minor because it is associated with the equipment performance attribute of the Initiating Events cornerstone and if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Specifically, if left uncorrected, the performance deficiency could have the potential to result in repetition of a significant condition adverse to quality, loss of control of feedwater regulating valve 642A and a manual scram. The NRC team evaluated the finding using Exhibit 1, Initiating Events Screening Questions, of IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, and determined this finding did not cause a reactor trip or the loss of mitigation equipment relied upon to transition the plant from the onset of a trip to a stable shutdown condition. Therefore, the NRC team determined the finding was of very low safety significance (Green). The NRC team determined that the finding had a cross-cutting aspect in the area of Human Performance, Procedure Adherence, because individuals did not follow processes, procedures, and work instructions. Specifically, Entergy did not follow procedure EN-LI-102, which provides the station standards for crafting a corrective action and states, in part, that the corrective action descriptions must be worded to ensure that the adverse condition or cause/factor is addressed (H.8).
05000293/FIN-2016011-042017Q1PilgrimProgrammatic Issue with Implementation of the Operability Determination ProcessThe NRC team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings. Specifically, the NRC team identified a programmatic issue because in some cases, Entergy did not enter the operability determination process when appropriate, and, when the process was entered, did not adequately document the basis for operability, in accordance with Procedure ENOP-104, Operability Determination Process, Revision 11. In each of the examples discussed, though the basis for operability was not adequate, all components were determined to be operable upon further evaluation. Entergy entered this issue into their corrective action program as CR-PNP-2017-00626. The performance deficiency was more than minor because if left uncorrected, could lead to a more significant safety issue. Specifically, the failure to enter and document a basis for operability could lead to not recognizing inoperable safety-related equipment, and place the reactor at a higher risk of core damage in a design basis accident. The NRC team evaluated the finding using Exhibit 2, Mitigating Systems Screening Questions, of IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, and determined this finding did not affect the design or qualification of a mitigating structure, system, or component; represent a loss of system and/or function; involve an actual loss of function of at least a single train or two separate safety systems for greater than its technical specification-allowed outage time; or represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant. Therefore, the NRC team determined the finding was of very low safety significance (Green). This finding had a cross-cutting aspect in the area of Human Performance, Teamwork. Specifically, the operations and engineering departments did not demonstrate a strong sense of collaboration and cooperation with respect to holding each other accountable when performing operability determinations to ensure nuclear safety is maintained (H.4).
05000293/FIN-2016011-052017Q1PilgrimFailure to Establish Corrective Actions to Address Scope of Procedure Quality IssuesThe NRC team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, because Entergy implemented inadequate corrective actions to address the procedure quality issues identified in CR-PNP-2016-02058. Specifically, Entergy inappropriately limited their corrective actions to those procedures that increased integrated risk above normal, and did not include other types of safety-related procedures that did not meet their procedure quality standards and resulted in procedure quality being a problem area. Entergy entered this issue into their corrective action program for further evaluation as CR-PNP-2017-00400. The performance deficiency was more than minor because it affected 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). Entergy limited corrective actions to procedures that increased integrated risk above normal or trip sensitive and failed to include other procedures associated with safety-related components that reflected the broader population reviewed during the collective evaluation. The NRC team evaluated the finding using Exhibit 2, Mitigating Systems Screening Questions, of IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, and determined this finding did not affect the design or qualification of a mitigating structure, system, or component; represent a loss of system and/or function; involve an actual loss of function of at least a single train or two separate safety systems for greater than its technical specificationallowed outage time; or represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant. Therefore, the NRC team determined the finding was of very low safety significance (Green). The NRC team determined that this finding had a cross-cutting aspect related to Human Performance, Resources, because the leaders failed to ensure that personnel, equipment, procedures, and other resources are available and adequate to support nuclear safety. Specifically, based on available resources, Entergy chose to limit the scope of safety-related procedures being revised to only those that resulted in high integrated risk or were trip sensitive (H.1).
05000293/FIN-2016011-062017Q1PilgrimDesign Change Not Appropriately Reviewed by EntergyThe NRC team identified a preliminary greater than Green finding and apparent violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, associated with Entergys failure to ensure that design changes were subject to design control measures commensurate with those applied to the original design and were approved by the designated responsible organization. Specifically, Entergy received a new style right angle drive for the A emergency diesel generator radiator blower fan from a vendor but failed to adequately review the differences in the design of the drives to identify potential new failure mechanisms for the part or the need for related preventive measures. Entergy entered this issue into the corrective action program as CR-PNP-2016-07443. The performance deficiency was more than minor because it was associated with the design control attribute of the Mitigating Systems cornerstone, and affected the associated cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, the team screened the finding for safety significance and determined that a detailed risk evaluation was required based on the A emergency diesel generator being inoperable for greater than the technical specification allowed outage time. Region I senior reactor analysts performed a detailed risk evaluation. The finding was preliminarily determined to be of greater than very low safety significance (greater than Green). The risk important sequences were dominated by external fire risk. Specifically, a postulated fire in the B 4 kilovolt (KV) switchgear room with a consequential loss of the unit auxiliary generator power supply, non-recoverable loss of off-site power (LOOP) to both safety buses A5 and A6, loss of the B emergency diesel generator with the conditional failure of the A emergency diesel generator, along with the loss of bus A8 feed (from the shutdown transformer or station blackout (SBO) diesel generator) to safety buses A5 and A6. The internal event risk was dominated by weather related LOOPs, failure of the A emergency diesel generator, with failure of the B emergency diesel generator and SBO diesel generator to run, along with failure to recover offsite power or the emergency diesel generators. See Attachment 1, A Emergency Diesel Generator Cooling Water System Degradation Detailed Risk Evaluation, for a detailed review of the quantitative criteria considered in the preliminary risk determination. The NRC team did not assign a cross-cutting aspect to this finding because the performance deficiency occurred in May 2000. Entergys program has undergone changes since May 2000, and the NRC team did not identify any recent examples of this performance deficiency. Other aspects of Entergys performance related to this issue are further discussed in Sections 5.10.3 and 6.3.4.
05000293/FIN-2016011-072017Q1PilgrimFailure to Report Condition Prohibited by Technical Specifications and a Safety System Functional FailureThe NRC team identified a Severity Level IV non-cited violation of 10 CFR 50.73, Licensee Event Report System, associated with Entergys failure to submit a licensee event report within 60 days following discovery of an event meeting the reportability criteria. Specifically, on September 28, 2016, Entergy identified the A emergency diesel generator was inoperable. The NRC team determined that the condition was prohibited by technical specifications and the inoperability of the A emergency diesel generator existed for a period of time longer than allowed by Technical Specification 3.5.F, Core and Containment Cooling Systems. This was also reportable as a safety system functional failure. Entergy entered this issue into the corrective action program as CR-PNP-2016-09552. Because this performance deficiency had the potential to impact the NRCs ability to perform its regulatory function, the NRC team evaluated the performance deficiency using traditional enforcement. The violation was evaluated using Section 2.3.11 of the NRC Enforcement Policy, because the failure to submit a required licensee event report may impact the ability of the NRC to perform its regulatory oversight function. In accordance with Section 6.9.d, Example 9, of the NRC Enforcement Policy, this violation was determined to be a Severity Level IV non-cited violation. Because this violation involves the traditional enforcement process and does not have an underlying technical violation, the NRC team did not assign a cross-cutting aspect to this violation, in accordance with IMC 0612, Appendix B.
05000293/FIN-2016011-082017Q1PilgrimFailure to Adequately Monitor the Performance of Maintenance Rule Scoped ComponentsThe NRC team identified a Green non-cited violation of 10 CFR 50.65(a)(2), Requirements for monitoring the effectiveness of maintenance at nuclear power plants. Specifically, Entergy did not demonstrate that the performance of 18 maintenance rule scoped components was effectively controlled through the performance of appropriate preventive maintenance, and did not establish goals and monitoring in accordance with 10 CFR 50.65(a)(1). Entergys immediate corrective action was to initiate a CR to evaluate moving the affected systems to 10 CFR 50.65(a)(1) monitoring requirements. Entergy entered this issue in the corrective action program as CR-PNP-2017-00401. The performance deficiency 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 availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, Entergy failed to demonstrate that the performance of the 18 maintenance rule scoped components was being effectively controlled through the performance of appropriate preventive maintenance which adversely impacts the reliability of those systems. The NRC team evaluated the finding using Exhibit 2, Mitigating Systems Screening Questions, of IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, and determined this finding did not affect the design or qualification of a mitigating structure, system, or component; represent a loss of system and/or function; involve an actual loss of function of at least a single train or two separate safety systems for greater than its technical specificationallowed outage time; or represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant. Therefore, the NRC team determined the finding was of very low safety significance (Green). The finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, in that Entergy failed to thoroughly evaluate and ensure that resolution of the identified issue, maintenance not being performed on maintenance rule scoped components, included reclassifying the components as necessary. Specifically, Entergy failed to demonstrate that the performance of Maintenance rule scoped components was effectively controlled through the performance of appropriate preventive maintenance, or through performance goals and monitoring. (P.2).
05000293/FIN-2016011-092017Q1PilgrimIneffective Corrective Actions to Address Conditions Adverse to Quality Regarding Components in Contact with or Close Proximity to the Drywell LinerThe NRC team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, associated with Entergys failure to correct a condition adverse to quality affecting safety-related equipment. Specifically, during a previous NRC inspection in August 2016, inspectors identified numerous locations in the drywell where non-seismic equipment was either in contact, or close proximity, with the drywell liner and had caused damage. Entergy initiated CRs and performed an operability evaluation for the identified issues. However, following a review of these CRs, the NRC team determined that Entergy failed to take corrective actions to address the condition adverse to quality. Entergy entered this issue into the corrective action program as CR-PNP-2016-09346 and CR-PNP-2016-09377 to perform an extent of condition review, secure the loose grating that had caused damage to the liner, and evaluate the need for a clearance criteria between components such as floor grating and support structures and the containment liner. The performance deficiency was more 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. Using IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 3, Barrier Integrity Screening Questions, the NRC team determined that this finding was of very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containment (valves, airlocks, etc.), containment isolation system (logic and instrumentation), and heat removal components. This finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, because the engineering evaluation of the degraded condition identified by the inspectors did not thoroughly evaluate the containment liner issues to ensure that resolutions address causes and extents of condition commensurate with their safety significance (P.2).
05000293/FIN-2016011-102017Q1PilgrimFailure to Promptly Correct a Condition Adverse to Quality for the Residual Heat Removal SystemThe NRC team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, because Entergy did not take timely corrective action for a previously identified condition adverse to quality. Specifically, Entergy failed to adequately resolve, through repair or adequate evaluation, gasket leakage on the B residual heat removal heat exchanger, which resulted in continued degradation and leakage for the heat exchanger gasket. Entergy did not consider this leakage as a degraded condition, with the potential to impact both the operability of the residual heat removal system, and PNPSs licensing basis with regards to leakage of a closed loop system outside of containment. After the NRC team raised the issue, Entergy performed an operability determination that established a reasonable expectation of operability pending implementation of corrective actions. Entergy entered this issue into their corrective action program as CR-PNP-2016-09725. The performance deficiency was more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to correct identified gasket leakage resulted in continued degradation and leakage of the heat exchanger gasket. The NRC team evaluated the finding using Exhibit 2, Mitigating Systems Screening Questions, of IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, and determined this finding did not affect the design or qualification of a mitigating structure, system, or component; represent a loss of system and/or function; involve an actual loss of function of at least a single train or two separate safety systems for greater than its technical specification-allowed outage time; or represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant. Therefore, the NRC team determined the finding was of very low safety significance (Green). The finding had a cross-cutting aspect in Human Performance, Conservative Bias, because Entergy failed to use decision making practices that emphasize prudent choices over those that are simply allowable (H.14).
05000293/FIN-2016011-112017Q1PilgrimFailure to Adequately Develop and Implement Targeted Performance Improvement PlansThe NRC team identified a Green finding because Entergy did not adequately develop and implement a CAPR of a root cause related to a Category A CR, as required by Entergy Procedure EN-LI-102, Corrective Action Program. Specifically, Entergy did not adequately develop and implement the Targeted Performance Improvement Plans, which were designated as a CAPR for the root cause for the Nuclear Safety Culture Fundamental Problem. Entergy documented this issue in the corrective action program for further evaluation as CR-PNP-2017-00406. The performance deficiency was more than minor because if left uncorrected, it could lead to a more significant safety concern. Specifically, inadequate implementation of the Targeted Performance Improvement Plans could result in recurrence of a culture in which leaders are not holding themselves and their subordinates accountable to high standards of performance, resulting in continuing performance issues at the station. The NRC team evaluated the finding using Exhibit 2, Mitigating Systems Screening Questions, of IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, and determined this finding did not affect the design or qualification of a mitigating structure, system, or component; represent a loss of system and/or function; involve an actual loss of function of at least a single train or two separate safety systems for greater than its technical specification-allowed outage time; or represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant. Therefore, the NRC team determined the finding was of very low safety significance (Green). This finding had a cross-cutting aspect in the area of Human Resources, Change Management, because leaders did not use a systematic process for evaluating and implementing change so that nuclear safety remains the overriding priority. In this case, PNPS leaders did not apply sufficient rigor in development and implementation of the Targeted Performance Improvement Plans such that they would be an adequate method to drive and sustain positive changes in the stations safety culture (H.3).
05000293/FIN-2016011-122017Q1PilgrimLicensee-Identified Violation10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, and shall be accomplished in accordance with those structures, procedures, and drawings. Entergy procedure EN-DC-148, Vendor Manuals and Vendor Re-Contact Process, Revision 6, requires, in part, that the station update vendor manuals every three years. Contrary to this, in July 2016, PNPS determined through a self-assessment that they had 13 vendor manuals that had not been evaluated for changes within 3 years. The NRC team determined that this finding did not affect the design or qualification of a mitigating structure, system or component; did not represent a loss of a system and/or function; did not result in loss of a train or two safety systems greater than any technical specification allowed outage time; did not result from an actual loss of safety function; and did not involve loss of any external event mitigating system. Consequently, the NRC team determined that this performance deficiency screened as having very low safety significance (Green). PNPS documented this issue in their corrective action program as CR-PNP-2016-05115.
05000293/FIN-2016011-132017Q1PilgrimLicensee-Identified Violation10 CFR 50.54(q)(2) requires, in part, that the licensee follow and maintain the effectiveness of an emergency plan to meet the planning standard of 10 CFR 50.47(b)(4). Specifically, the licensee was to maintain the necessary equipment to support the effectiveness of EALs. Contrary to these requirements, PNPS identified in CR-PNP-2016-01491 that on three past occasions (March 15 through August 8, 2012; September 4 through October 14, 2012; and June 4 through June 14, 2015) both trains of the H2O2 monitors and the Post-Accident Sampling System were unavailable to ensure the effectiveness of EAL 24, Deflagration concentrations exist inside PC, for the potential loss of the containment barrier within the Fission Product Barrier category of the EALs. This issue meets the criteria for very low safety significance (Green) because, due to other EALs, an appropriate emergency declaration could have been made in an accurate and timely manner.
05000305/FIN-2008004-012008Q3KewauneeOperability Evaluation for Degraded Pedestals Failed to Adequately Evaluate Degraded Conditions Per ProceduresOperability Evaluation for Degraded Pedestals Failed to Adequately Evaluate Degraded Conditions Per Procedure 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. (P.1(c)) (Section 1R15)
05000305/FIN-2008004-022008Q3KewauneeFailure to Perform a 10 CFR 50.59 Screening for Alteration During Maintenance That Existed for More Than 90 DaysA 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. (H.3(a)) (Section 1R15)
05000305/FIN-2008004-032008Q3KewauneeInadequate Procedure Results In Unplanned Control Rod MotionA 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 crosscutting 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. (H.1(a)) (Section 4OA3)
05000317/FIN-2012003-012012Q2Calvert CliffsFailure to Establish Testing Program for ESFAS SDSThe inspectors identified an NCV of 10 CFR 50, Appendix B, Criterion XI, Test Control, because Constellation did not establish an operational test program for the engineered safety features actuation system (ESFAS) shutdown sequencers (SDSs). Specifically, on May 4, 2012, the inspectors determined that the licensee had never performed an operational test on the SDSs. The SDS supports the Loss of Offsite Power (LOOP) event in chapter 14 of the Updated Final Safety Analysis Report (UFSAR). Constellations immediate corrective actions included entering the issue into their corrective action program (CAP), conducting an operability determination (OD), developing a procedure to test the SDSs online, and testing the SDSs. Planned corrective actions include submittal of a license amendment request to include the SDS testing in their technical specification (TS) requirements. The finding is more 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. Specifically, when tested, one of the SDSs did not perform as designed. The SDS logic for the No. 24 4kV bus initiated start of the auxiliary feedwater (AFW) pump on the incorrect step. In addition, if left uncorrected the performance deficiency had the potential to lead to a more safety significant concern, in that, an SDS failure would go undetected until an actual demand during an LOOP. The inspectors evaluated the finding using Phase 1, Initial Screening and Characterization, worksheet in Attachment 4 to IMC 0609, Significance Determination Process, and determined the finding is of very low safety significance (Green) because the performance deficiency was not a design or qualification deficiency, did not involve an actual loss of safety function, did not represent actual loss of safety function of a single train for greater than its TS allowed outage time, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The finding has a cross-cutting aspect in the area of problem identification and resolution, CAP, because Constellation did not identify this issue completely, accurately, and in a timely manner commensurate with its safety significance. Specifically, within the last 3 years, Constellation had several opportunities to completely and accurately identify the SDS test program deficiency as a result of multiple sequencer module replacements and through reviews of the emergency diesel generator (EDG) testing program
05000317/FIN-2012003-022012Q2Calvert CliffsFailure to Establish and Maintain Adequate Procedures for Maintenance on Pressurizer Power Operated Relief ValvesA self-revealing NCV of TS 5.4.1, Administrative Controls Procedures, was identified for the failure to establish and maintain adequate procedures for performing maintenance on pressurizer power operated relief valves (PORVs). Specifically, the maintenance procedure (purchase order) did not clearly prescribe acceptance criteria for the minimum acceptable clearances between the cage, guide, and the main disc. This resulted in the as left internal valves clearances being less than the minimum expected requirements. During disassembly, the valve disc of one of the PORVs (serial number BS07325) was stuck and had to be mechanically removed. Immediate corrective actions included entering this issue into the CAP, conducting an OD for the valves currently installed on both units, and conducting a past operability review of the PORVs that were removed. Planned corrective actions include updating the design specification and maintenance procedures to ensure that minimum allowable internal clearances are specified. This finding is more than minor because it is associated with the procedure quality attribute of the Mitigating System cornerstone and affects 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, when the valve was removed and disassembled, the valve disc was found stuck and had to be mechanically removed, thereby impacting the reliability and operability of the valve during operation at power the previous cycle. A detailed engineering analysis was performed which supported past operability of the valve. The inspectors evaluated the finding using Phase 1, Initial Screening and Characterization, worksheet in Attachment 4 to IMC 0609, Significance Determination Process, and determined the finding is of very low safety significance (Green) because the performance deficiency was not a design or qualification deficiency, did not involve an actual loss of safety function, did not represent actual loss of safety function of a single train for greater than its TS allowed outage time, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The finding has a cross-cutting aspect in the area of human performance, work practices, because personnel work practices did not support human performance. Specifically, Constellation did not ensure supervisory and management of oversight of work activities, including contractors, such that nuclear safety is supported. Critical dimensions affecting contractor work activities were not adequately captured in station processes, procedures, and work packages.
05000333/FIN-2010004-012010Q3FitzPatrickAppendix R Fire Door Blocked Open Without Establishing Required MeasuresThe inspectors identified a non-cited violation (NCV) of very low safety significance of license condition 2.C(3), \"Fire Protection,\" because Entergy personnel blocked a fire door in the open position, defeating its three hour fire barrier function, without establishing the required compensatory measures. Entergy personnel entered this issue into their corrective action program (CAP) as CR-JAF-2010-04825, issued a night order emphasizing the requirements associated with propping open fire doors, provided coaching, and submitted a procedure change request to further clarify procedural applicability requirements. This finding is more than minor because it is associated with the protection against external events attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability of systems that respond to initiating events to prevent undesirable consequences (Le., core damage). Specifically, the fire door being affixed open without the knowledge of the control room personnel and other operators and without an assigned fire watch resulted in a barrier to fire propagation that was less robust than required by the NRC approved fire protection program. The inspectors determined the significance of the finding using IMC 0609, Appendix F, \"Fire Protection Significance Determination Process,\" Phase 1. The finding was determined to be of very low safety significance (Green) because the deficiency represented a low degradation rating. Specifically, the individuals involved were members of the fire brigade, qualified in fire watch duties, and only blocked the door open during resin container transfers. The inspectors determined this finding had a cross-cutting aspect in the area of human performance within the work practices component because Entergy did not effectively communicate expectations to personnel regarding the applicable procedures and personnel did not follow the procedures (H.4(b)).
05000333/FIN-2010004-022010Q3FitzPatrickLicensee-Identified ViolationThe following violation of very low safety significance (Green) was identified by Entergy and is a violation of NRC requirements which meet the criteria of Section VI of the NRC Enforcement Policy, NUREG-1600, for being dispositioned as a NCV. TS 3.5.1 requires that with two or more low pressure ECCS injection/spray subsystems inoperable for reasons other than one low pressure injection pump in both low pressure injection subsystems inoperable that LCO 3.0.3 must be entered immediately. Contrary to this, on March 22, 2010, Entergy personnel identified that they had not complied with TS 3.5.1 for various periods of time since 1981, including most recently from February 5, 2010. Entergy personnel documented this condition in CR-JAF-201 0-01382 and CRJAF- 2010-01595. The inspectors evaluated this finding using IMC 0609.04, \"Phase 1 - Initial Screening and Characterization of Findings,\" and determined that the condition was of very low safety significance (Green) because it did not result in the loss of the overpressure relief safety function in either the RHR or CS systems.
05000333/FIN-2010005-012010Q4FitzPatrickInadequate Procedure for Refueling Water Level Control Resulted in Overflowing of Reactor Cavity Water in the Reactor BuildingA self-revealing NCV of very low safety significance of technical specification (TS) 5.4, Procedures, was identified because Entergy procedure OP-30A Refueling Water Level Control, did not provide adequate guidance to operators for filling the reactor cavity which resulted in the reactor building (RB) floor drains overflowing and water intrusion from higher to lower levels in the RB. Entergy personnel entered this issue into their corrective action program (CAP), (CR-JAF-2010-05406 and CR-JAF-2010-05407) and performed several actions to ensure proper water level control prior to the next drain down of the reactor cavity. These actions included revising OP-30A to provide sufficient detail, ensuring additional detail would be included in pre-job briefings to include potential drain paths from the reactor cavity and spent fuel pool, and installing a dedicated camera to monitor reactor cavity water level. This finding is more than minor because it is 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. Specifically, water spray throughout areas of the RB created a potential for water entering motors, valve operators, motor control centers, circuit breakers, and electricaljunction boxes, such that electrical components could have been compromised, which increased the likelihood of an event that would upset plant stability and challenge a critical safety function. The inspectors determined the significance of the finding using IMC 0609, Appendix G, Shutdown Operations Significance Determination Process, Phase 1. The finding was determined to be of very low safety significance because Entergy personnel maintained an adequate mitigation capability and there was there neither an inadvertent loss of two feet of RCS inventory nor an inadvertent reactor coolant system pressurization. The inspectors determined this finding had a cross-cutting aspect in the area of human performance within the resources component because the procedure used for filling the reactor cavity was not sufficiently complete to assure nuclear safety.
05000333/FIN-2010005-022010Q4FitzPatrickFailure to Maintain Equipment Status Control for a Manually Operated Normally Locked Open Residual Heat Removal Injection ValveA self-revealing NCV of very low safety significance of TS 5.4, Procedures, was identified because Entergy personnel did not implement AP-12.06, Equipment Status Control, as required. Specifically, Entergy personneldid not maintain status control and properly document the position of the residual heat removal (RHR) to reactor water recirculation loop B isolation valve (10RHR-818) as closed nor did operators restore the valve to its normal locked open position upon completion of a leak surveillance test. Entergy personnel entered this issue into their corrective action program (CAP), (CR-JAF-2010-06656) and promptly restored the valve to its required locked open position. This finding is more than minor because it is associated with the configuration control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Specifically, the operators did not maintain configuration control of the RHR isolation valve and restore the valve to a locked open position when the B RHR subsystem was credited for maintaining acceptable shutdown risk. The inspectors determined the significance of the finding using IMC 0609, Appendix G, Shutdown Operations Significance Determination Process. The issue was determined to screen as very low safety significance (Green) because the B RHR train could be considered available with respect to Appendix G, Section 4.0, and Attachment 3, Section 2,2.3. Specifically, the inspectors determined that operators had more than twice the time available (with a shortest time to boil of 5.8 hours) than would have been required to identify and take action to restore/open the RHR isolation valve in the event of a loss of shutdown cooling or RCS inventory. This finding had a cross-cutting aspect in the Human Performance cross-cutting area, Work Practices component, because Entergy personnel did not define and effectively com m unicate expectations regard ing procedural compliance, and personnel did not follow procedures.
05000333/FIN-2010005-032010Q4FitzPatrickLicensee-Identified ViolationTS 5.4 requires, in part, that the applicable procedures recommended in Regulatory Guide 1.33, Appendix A, November 1972,be established, implemented, and maintained, Regulatory Guide 1.33, Section D, Procedure for Startup, Operation, and Shutdown of Safety-Related BWR Systems, specifies, in part, that instructions for draining should be prepared, as appropriate, for the shutdown cooling, fuel storage pool cooling, and condensate systems. Entergy staff identified that, contrary to the above, they had not complied with TS 5.4.1 on October 3, 2010, when Entergy personnel did not adequately implement procedure OP-30A, Attachment 2, Checklist for Draining, step F.4.1. Specifically, upon performing additional reviews as a result of finding the B RHR valve mispositioned, Entergy staff identified that the B RHR auto control bypass and A and B CS auto actuation bypass switches had been in bypass when step F.4.1 was performed, which through verifying compliance with various technical specifications, required the verification that two low pressure emergency core cooling systems be operable prior to installing the spent fuel pool gates. However, with the three switches placed in bypass, only one emergency core cooling system, A RHR, was operable. Entergy personnel documented this condition in CR-JAF-2010-06659. The inspectors determined the significance of this finding using IMC 0609, Appendix G, Shutdown Operations Significance Determination Process. The issue was determined to screen as very low safety significance (Green) because the B RHR and A and B CS subsystems could be considered available with respect to Appendix G, Section 4.0, and Attachment 3, Section 2.2.3. Specifically, the inspectors determined that operators had more than twice the time available than would have been required to identify and take action to restore an additional injection source given an inadvertent RCS inventory loss.
05000333/FIN-2010005-042010Q4FitzPatrickLicensee-Identified Violation10 CFR 50, Appendix B, Criterion ll, Quality Assurance Program, requires, in part, that the licensee establish a quality assurance program which complies with Appendix B. This program shall be documented by written policies, procedures, or instructions and shall be carried out throughout plant life in accordance with those policies, procedures, or instructions. Procedure EN-QV-1Il, Training and Certification of Inspection/Verification and Examination Personnel, Section 4.0141(i), requires that the Entergy corporate ANSI Level lll inspector shall perform periodic (annual) surveillances of quality control inspection activities to ensure that the program is being adequately implemented and maintained. Contrary to the above, no surveillances of quality control inspection activities were performed for any Entergy site during calendar year 2008. The issue was not suitable for quantitative significance determination, so it was assessed using IMC 0609, Appendix M, and evaluated using the qualitative criteria listed in Table 4.1. This finding was determined to be of very low safety significance because other quality assurance program functions remained unaffected by this performance deficiency, so defense-in-depth continued to exist. This issue was entered into Entergy\'s CAP as CR-HQN-2009-00111.
05000336/FIN-2010002-012010Q1MillstoneLicensee-Identified ViolationTechnical Specification 6.8.1 requires, in part, that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Regulatory Guide1.33. Contrary to this, Dominion\'s procedure MP2704U, Containment Personnel Airlock, was inadequate because it was not established with sufficient detail to complete interlock restoration. As a result, the containment airlock interlock was not properly reset at the end of the refueling outage on November 13, 2009. On December 2, 2009, containment integrity was not maintained when the improperly reset interlock allowed both airlock doors to be briefly open during a containment entry while Unit 2 was at 100 percent power. Dominion immediately closed the door and entered the issue into their corrective action process, CR 360277. The finding is of very low safety significance because both airlock doors were only open briefly 1 minute) and there were no other degraded plant conditions that would have created a safety concern during the time the airlock doors were open
05000336/FIN-2010003-012010Q2MillstoneFailure to Properly Identify and Correct a Degraded Governor Condition in the Unit 2 \'A\' EDGA self-revealing, NCV of 1 0 CFR 50, Appendix B, Criterion XVI, Corrective Action, was identified for Dominion\'s failure to properly evaluate a condition adverse to quality involving the Unit 2 \'A\' emergency diesel generator (EDG). Dominion did not properly evaluate a degraded condition of the \'A\' EDG, which led to its inoperability from May 12,2010, to May 17, 2010. Dominion took immediate corrective action to replace the EDG govemor. The inspectors determined this finding was more than minor because it was associated with the Human 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, Dominion\'s inadequate evaluation of the degraded condition of the \'A\' EDG governor after the March 17, 2010, surveillance test did not result in effective corrective action to address the cause of the rapid load increase. As a result, the \'A\' EDG was declared inoperable when it again experienced a rapid load increase during its surveillance on May 12, 2010. The inspectors determined the significance of the finding using IMC 0609.04, Phase 1 - Initial Screening and Characterization of Findings, and determined that the finding was of very low safety significance (Green) because it did not represent an actual loss of safety function of a single train for greater than its Technical Specification allowed outage time. The inspectors determined that this finding had a cross-cutting aspect in the Human Performance cross-cutting area, Decision Making component, because Dominion did not use conservative assumptions in its decision making when they could not conclude that the EDG load fluctuations would not recur (H.1 (b )).
05000336/FIN-2010003-022010Q2MillstoneFailure to Properly Plan Work Activities for the Unit 2 \'D\' Circulating Water Bay Outage Results in Manual Reactor Trip.A self-revealing finding of very low safety significance (Green) was identified for Dominion\'s failure to properly plan work activities associated with the Unit 2 \'D\' circulating water (CW) bay outage in accordance with Dominion procedure WM-M- 3000, Managing Complex Work. The work plan failed to properly sequence work activities to prevent fouling of the \'C\' CW screens. The subsequent fouling of the \'C\' CW travelling screen resulted in an automatic trip of the \'C\' CW pump. Loss of the \'C\' CW pump, coupled with the unavailability of the \'D\' CW pump, required the operators to manually trip the reactor. Dominion entered this issue into their corrective action program (CR370363). This finding is more than minor because it was similar to NRC IMC 0612, Appendix E, Examples of Minor Issues, Example 4b, in that the implementation of the inadequate work plan caused the loss of the \'C\' CW pump, and required the operators to manually trip the reactor. The inspectors determined this finding was associated with the Human Performance attribute of the Initiating Events cornerstone, and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. Specifically, the work plan for the \'D\' CW bay outage did not properly sequence the work, which led to the loss of the \'C\' CW pump and required the operators to manually trip the reactor. The inspectors determined the significance of the finding using IMC 0609.04, Phase 1 - Initial Screening and Characterization of Findings, and determined that the finding was of very low safety significance (Green) because it did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available. The inspectors determined that this finding had a cross-cutting aspect in the Human Performance cross-cutting area, Work Control component, because Dominion did not appropriately plan the bay cleaning and demucking (removal of scraped material) work activity to address the risk of the activity to impact the other CW bays (H.3(a)).
05000336/FIN-2010003-032010Q2MillstoneFailure to Make a 10 CFR 50.72 (b )(3)(v) Report for an Inoperable Unit 3 Secondary ContainmentThe inspectors identified a Severity Level IV NCV of 10 CFR 50.72(b)(3)(v); in that, Dominion failed to make a timely 10 CFR 50.72 eighthour report to the NRC for a condition that, at the time of discovery, could have prevented secondary containment from fulfilling its safety function. On May 27, 2010, operations personnel found both sets of the auxiliary and service building tunnel exhaust dampers open which could have prevented secondary containment from fulfilling its safety function. Operations declared secondary containment inoperable, closed the auxiliary building tunnel exhaust dampers to restore operability, and initiated a 10 CFR 50.72 report. The inspectors determined that Dominion\\\'s failure to make a 10 CFR 50.72 eight-hour report to the NRC regarding the inoperable secondary containment as a condition that could have prevented it from fulfilling its safety function was a performance deficiency. The inspectors determined that traditional enforcement applied, since the failure to make a required report could adversely impact the NRC\\\'s ability to perform its regulatory function. In accordance with the NRC Enforcement Policy, Supplement I - Reactor Operations, Example D.4, a failure to make a required Licensee Event Report (LER) is categorized as a Severity Level IV violation. The inspectors determined that this finding had a cross-culling aspect in the Human Performance cross-culling area, Decision Making component, because Dominion did not use conservative assumptions in their decision-making when they could not demonstrate that secondary containment would have fulfilled its safety function (H.1(b)).
05000336/FIN-2010003-042010Q2MillstoneCharging Pump Overheating and Cavitation during RCS Loop Vacuum FillA self-revealing, NCVof 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified for Dominion\'s failure to have an adequate procedure for operating the Unit 3 charging pumps. Specifically, Dominion operating procedure (OP) 3304A, Charging and Letdown, did not require verification of Reactor Plant Closed Cooling Water (RPCCW) flow to the seal water heat exchanger, which resulted in overheating of the \'B\' charging pump during a reactor coolant system (RCS) vacuum fill on May 1, 2010. Dominion has created corrective actions to make procedural enhancements to OP-3304A, Charging and Letdown, and OP-3353.MB1C, Main Board Annunciator Response. The inspectors determined this finding was more than minor because it was associated with the Configuration 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. The inspectors performed an initial screening of the finding in accordance with IMC 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The inspectors then evaluated the significance of the finding using Inspection Manual Chapter 0609, Appendix G, Shutdown Operations - Significance Determination Process, Checklist 3, PWR Cold Shutdown and Refueling Operation; RCS Open and Refueling Cavity Level < 23\' Or RCS Closed and No Inventory in Pressurizer; Time to Boiling < 2 hours, and determined that the finding was of very low safety significance (Green) because all of the shutdown safety function guidelines were met. The inspectors determined that this finding had a cross-cutting aspect in the Human Performance cross-cutting area, Work Control component, because Dominion relied on the work control process to assure that the RPCCW cooling water was in service to the seal water heat exchanger at the time that the RCS vacuum fill was scheduled. Specifically, the work control process was insufficiently robust to ensure that cooling water was supplied to the seal water heat exchanger during charging pump operations (H.3(b)).
05000336/FIN-2010003-052010Q2MillstoneReactor Trip Caused by Loss of Positive Control of Steam Generator LevelA self-revealing finding of very low safety significance (Green) was identified for Dominion\'s failure to correct a long-standing stability problem with control of the Unit 3 feedwater regulating bypass valves (FRBVs). Operation at low power conditions has resulted in excessive steam generator (SG) level oscillations while in automatic control and unintended equipment response when attempting to control SG level in manual control. The inadequate design of the SG level control system for low power operations was identified by numerous condition reports dating back to 2002, but had not been corrected. Dominion entered this issue into their corrective action program (CR381435, CR384014). The finding is more than minor because it was similar to NRC Inspection Manual Chapter (IMC) 0612, Appendix E, Examples of Minor Issues, Example 4b, in that the failure to correct a condition adverse to quality resulted in a reactor trip. The inspectors determined that the finding was associated with the Equipment Performance attribute of the Initiating Events cornerstone, and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions. Specifically, the long standing condition of the FRBVs\' inability to control SG level at low power operations led to an automatic reactor trip. The inspectors performed an initial screening of the finding in accordance with IMC 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The inspectors determined the finding was of very low safety significance (Green) because it did not affect both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available. The inspectors determined that the finding had a cross-cutting aspect in the Problem Identification and Resolution cross-cutting area, Corrective Action Program component, because Dominion did not take appropriate corrective action to address the longstanding adverse conditions associated with control of the FRBVs (P.1 (d)).
05000354/FIN-2011008-012011Q1Hope CreekInadequate Corrective Actions for EHC Turbine Valve ProceduresThe inspectors identified a finding of very low safety significance (Green) because PSEG did not correct turbine valve test and maintenance procedure deficiencies. Specifically, PSEG closed out notification 2043100 within their corrective action program without performing the actions to resolve the procedure deficiencies as required by PSEG corrective action procedures. PSEG entered this issue into their corrective action program as notifications 20494248 and 20495156 to evaluate the corrective actions needed to address the issue. The finding was determined to be more than minor because the deficiency was associated with the procedure quality attribute of the Initiating Events cornerstone and adversely impacted the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. The inspectors evaluated the finding using IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Table 4a, for the Initiating Event cornerstone. Specifically, 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, the finding was determined to be of very low safety significance (Green). This finding had a cross-cutting aspect in the area of problem identification and resolution because PSEG did not take appropriate corrective actions to address safety issues in a timely manner, commensurate with their safety significance and complexity. Specifically, corrective actions outlined in notification20413100 to resolve procedural deficiencies were not completed.
05000354/FIN-2013003-012013Q3Hope CreekInadequate Preventive Maintenance Replacement Schedule for Agastat Control RelaysA self-revealing Green NCV of Technical Specifications (TS) 6.8.1, Procedures, was identified because PSEG failed to establish an appropriate preventive maintenance (PM) schedule for Tyco/Agastat General Purpose (GP) control relays. Specifically, the evaluation PSEG performed to revise the relay replacement periodicity from 22 years to 40 years neither adequately addressed available relay references nor all applicable failure mechanisms. As a result, high pressure coolant injection (HPCI) failed to respond to logic system actuation signals during surveillance testing on April 8, 2013. PSEGs immediate corrective actions included replacing failed relays and placing the issues in the corrective action program (CAP). Additionally, PSEG plans to revise the replacement frequency and to replace other Tyco/Agastat GP control relays of high safety significance, as identified in their extent of condition review. 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 of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure of a control relay caused the HPCI system to fail to automatically actuate during testing, and the HPCI system was unexpectedly declared inoperable. The inspectors evaluated the finding in accordance with IMC 0609, Significance Determination Process, issued June 2, 2011, and determined the finding is of very low safety significance (Green) following a detailed risk evaluation. No cross-cutting aspect was assigned to this finding because PSEG decisions made with regard to evaluating the PM replacement periodicity were made more than 3 years ago and a PM Ownership Committee has since been created to review PM change evaluations; therefore, this performance deficiency is not reflective of current plant performance.
05000354/FIN-2013004-012013Q3Hope CreekFailure to Follow PMT Procedure Prior to Returning the B FRVS Recirculation Fan to ServiceA self-revealing finding of very low safety significance (Green) and associated NCV of Title 10 of the Code of Federal Regulation (10 CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, was identified for PSEGs failure to properly test the B Filtration, Recirculation and Ventilation System (FRVS) recirculation fan following maintenance in accordance with site procedures. Specifically, on June 3, 2013, PSEG did not perform the required post-maintenance test (PMT) prior to returning the system to service. Consequently, when the fan failed during its surveillance on June 24, 2013, there was no reasonable assurance that the fan was operable since the last time maintenance was performed on it. Corrective actions included adding this event to the Licensed Operator Requalification training program to improve knowledge regarding PMT requirements. The performance deficiency (PD) was determined to be more than minor because it is associated with the system, structure, or component (SSC) and Barrier Performance attribute of the Barrier Integrity cornerstone, and affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Using IMC 0609, Appendix A, The SDP for Findings At-Power, dated June 19, 2012, the finding was determined to be of very low safety significance (Green) because it only represented a degradation of the radiological barrier function provided for the standby gas treatment system. This finding had a cross-cutting aspect in the area of Human Performance, Decision-Making, because PSEGs decisions did not demonstrate that nuclear safety was the overriding priority. Specifically, PSEG did not use conservative assumptions in decision-making when determining the proper PMT for the B FRVS recirculation fan prior to returning it to service.
05000354/FIN-2013004-022013Q3Hope CreekFailure to Perform Maintenance in Accordance with Station Procedures Led to RCS Pressure Boundary LeakageA self-revealing finding of very low safety-significance (Green) and associated NCV of Technical Specifications (TS) 3.4.3.2, Reactor Coolant System (RCS) Operational Leakage, was identified on June 12, 2013, when a through-wall flaw was discovered in the RCS pressure boundary. Specifically, because Hope Creek failed to perform maintenance on a B residual heat removal (RHR) shutdown cooling (SDC) vent line in accordance with PSEG maintenance procedures, the plant operated with RCS pressure boundary leakage for a period of time prohibited by TS. Immediate corrective actions included vent line assembly replacement and examination of additional vent line assemblies installed on RHR piping in the drywell under the same design change. Planned corrective actions include visual examination of other components that had work involving cutting on small bore piping in the drywell. The PD is more than minor because it is associated with the RCS Equipment and Barrier Performance attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, RCS, and containment) protect the public from radionuclide releases caused by accidents or events. PSEGs failure to perform maintenance in accordance with station procedures resulted in plant operation with a condition prohibited by TS and the degradation of a principal safety barrier. The inspectors determined that the finding is of very low safety significance (Green) because the piping flaw, after a reasonable assessment of degradation, could not result in exceeding the RCS leak rate for a small loss of coolant accident (LOCA) and could not likely affect other systems used to mitigate a LOCA resulting in a total loss of their function. This finding had a cross-cutting aspect in the Human Performance area, Work Practices component, because PSEG did not ensure supervisory and management oversight of work activities, including contractors, such that nuclear safety is ensured. Specifically, management did not appropriately supervise workers to ensure work was performed in accordance with site maintenance procedures. Even though the PD could have occurred any time between November 2007 and May 2012, the inspectors determined that the performance characteristic associated with ineffective work activity oversight could not be conclusively placed in the earlier portion of that time window. In addition, PSEG had not previously implemented actions to specifically correct or eliminate the potential for this PD. As a result, the inspectors concluded that the PD is indicative of current performance.
05000354/FIN-2013004-032013Q3Hope CreekLicensee-Identified ViolationOn August 27, 2013, a radiation protection technician transported four sources of radioactive material to an area of the owner controlled property for use in a maintenance activity. At the end of shift, the technician brought the sources offsite on the public highways, rather than returning them to the protected area, without the knowledge of radiation protection management. Once radiation protection management became aware that the sources were offsite, the technician was contacted and promptly returned the sources to Hope Creek. 10 CFR 71.5, and 49 CFR Parts 172 and 173 require that such sources be properly shipped when on public highways. Contrary to these requirements, these sources were transported on public highways without the required specification packaging, marking, labeling, and shipping papers. This finding was evaluated using IMC 0609, Appendix D for public radiation safety. The finding involved radioactive material control in the area of transportation, but did not involve exceeding a radiation limit, package breach, certificate of compliance, burial ground nonconformance, or failure to make a required notification; therefore, the finding is Green. This issue was documented in PSEGs CAP as Notification 20619627.
05000387/FIN-2011004-012011Q3SusquehannaViolation of 10CFR55.25, Failure to Notify NRC of a Change in Medical Status and Request a Conditional LicenseThe inspectors identified a SL lV NOV of 10 CFR 55.25, lncapacitation Because of Disability or lllness, for PPL failing to notify the NRC of a known permanent change in medical status of a licensed operator, and 10 CFR 55.3, License Requirements, for failing to ensure that an individual license holder, in the capacity of a reactor operator (RO), met the medical prerequisites prior to performing licensed operator duties. Specifically, an RO failed a medical examination in both 2009 and 2011 which identified a disqualifying condition and performed licensed duties without an NRC-approved, amended license. He performed the function of an RO while on watch from April 2009 through August 2011, when the NRC identified this issue. However, the operator did wear corrective lenses while standing watch since April 2009. Upon notification PPL submitted, and the NRC approved, a conditional license to address the disqualifying medical condition. PPL entered this issue into their corrective action program (CAP) as condition report (CR) 1 4501 38. The inspectors determined that PPL\\\'s failure to notify the NRC of a known permanent change in a licensed operator\\\'s medical status and request an amended license in order to assume licensed duties was a performance deficiency. This finding was evaluated using the traditional enforcement process because the issue had the potential to impact or impede the regulatory process. Specifically, there was a potentialfor license termination or the issuance of a conditional license to accommodate for a medical condition. The RO performed licensed duties from April 2009 through August 2011 with a disqualifying condition that required his license to be amended. Using the NRC Enforcement Policy, this violation was characterized at SL lV, in accordance with Section 6.4. This violation is being cited in the enclosed Notice in accordance with NRC Enforcement Manual Section 3.1.2, because the violation was determined to be repetitive of NRC Enforcement Action (EA) 09-248 dated January 28,2Q1Q, an SLlll Notice of Violation related to a Senior Reactor Operator (SRO) standing watch without meeting a medical qualification requirement. The medical conditions in both the former and current cases were similar: therefore, it was reasonable that an adequate extent of condition review for EA-09-24g should have identified the additional discrepancy. This significance of the associated performance deficiency was screened against the Reactor oversight Process (Rop) per the guidance of tMC 0612, Appendix B. No associated ROP finding was identified and no cross-cutting aspect was assigned
05000387/FIN-2013003-012013Q2SusquehannaInadequate Operability Assessment of Synchroscope SwitchInspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, when PPL performed an inadequate operability determination for a synchroscope switch failure that rendered offsite power and the four emergency diesel generators (EDGs) inoperable. This resulted in PPL being in violation of Unit 1 TSs 3.8.1, 3.8.2, and 3.0.3, and Unit 2 TSs 3.6.4.1 and 3.8.2. PPL entered the issue in their CAP as CR 1703293, re-evaluated past operability and submitted a licensee event report (LER) for the associated condition prohibited by plant Technical Specifications (TS) on July 8, 2013 (ADAMS Accession No. ML13190A104). The performance deficiency was determined to be more than minor since it was associated with the equipment performance attribute of the Mitigating Systems cornerstone and adversely affected its objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was evaluated using the SDP of IMC 0609.04. The finding was evaluated under both the Mitigating Systems Exhibit of IMC 0609 Appendix A when Unit 1 was at power and Appendix G for the times when one or both units were in a shutdown condition. Under IMC 0609, Appendix A, the finding screened to Green since it was not a design or qualification deficiency and was not a potential or actual loss of system or safety function. Under IMC 0609, Appendix G, Attachment 1, Checklists 5 through 7, the inspectors screened the issue to Green since it affected the requirement for operable DGs under TS 3.8.1 and TS 3.8.2. The inspectors determined that a Phase 2 analysis was not warranted since it did not match those criteria listed for further analysis in these checklists. Specifically, since all automatic transfer functions of off-site power and the EDGs remained functional, inspectors determined that none of the functions evaluated under the SDPs were affected. The finding had a cross-cutting aspect in Problem Identification and Resolution (PI&R), corrective action program (CAP), because PPL staff did not thoroughly evaluate problems such that the resolutions address the causes and extent of conditions, to include properly classifying, prioritizing and evaluating for operability. Specifically, PPL staff did not appropriately evaluate the effect that the synchroscope switch failure had on offsite power and emergency diesel generator operability.
05000387/FIN-2013003-022013Q2SusquehannaUnacceptable Preconditioning of RPS and EOC-RPT Time Response TestInspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion XI, Test Control, because PPL staff performed unacceptable preconditioning by performing corrective maintenance prior to recording the as-found time response of the reactor protection system (RPS) and end-of-cycle recirculation pump trip (EOC-RPT) for the turbine control valve (TCV) fast closure function. Specifically, corrective maintenance was performed with the potential to improve the time response of the system without verifying that the as-found condition was within the acceptance criteria assumed in the accident analysis. PPL entered the issue into their CAP as CR 1712564 and verified as-left data was verified to be within acceptance criteria which provided reasonable assurance that the SSC would perform satisfactorily during the subsequent operational period. Inspectors determined the performance deficiency is more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone and affected the 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 collect as-found data could result in the inability to verify the operability of structures, systems, and components (SSC). Additionally, in this case, the test had exhibited low margin and unreliable performance during its previous surveillance test. The inspectors determined, through a review of IMC 0609, Appendix A, Exhibit 2, that the finding was Green because the finding was not related to a design or qualification deficiency, did not represent a loss of a mitigating system safety function, and did not screen as potentially risk significant due to external initiating events. The finding is related to the cross-cutting area of PI&R, CAP, in that PPL did not take appropriate corrective actions to address safety issues and adverse trends in a timely manner, commensurate with their safety significance and complexity. Specifically, though degraded performance was identified during previous testing, PPL staff did not take timely and effective corrective actions to ensure the required maintenance did not unacceptably precondition the following 24-month surveillance test.
05000387/FIN-2013003-032013Q2SusquehannaInadequate Procedure to Control and Monitor RCS Heatup RateInspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because PPL did not adequately incorporate the acceptance criteria for heatup rate specified in the plant TSs, as amplified in its basis, into the surveillance test implementing procedure for monitoring adherence to pressure and temperature requirements during plant heatup and cooldown. Based on this procedure inadequacy, operators exceeded the TS limit during a plant startup on May 28, 2013. PPL entered the issue into their CAP as CR 1709058 and revised plant procedures to appropriately incorporate the acceptance criteria. This performance deficiency is more than minor because it was associated with the human performance and procedure quality attribute of the Barrier Integrity cornerstone and affected the objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system (RCS), and containment) protect the public from radionuclide releases caused by accidents or events. Using IMC 0609, Significance Determination Process, Appendix A, Exhibit 3, Barrier Integrity Screening Questions, the inspectors determined that this issue required a detailed risk evaluation. In consultation with a Region I Senior Reactor Analyst, the inspectors completed a qualitative risk assessment and determined this issue is of very low safety significance (Green). Specifically, there was no impact on the integrity of the reactor vessel due to the short duration temperature gradient imposed by exceeding the TS heatup rate. Consistent with PPLs evaluation, the observed heatup rate minimally exceeded the specified limit during plant startup and remained within the acceptable bounds of the current plant pressure and temperature analysis. The finding is related to the cross-cutting area of PI&R, Corrective Actions, in that PPL did not take appropriate corrective actions to address safety issues and adverse trends in a timely manner, commensurate with their safety significance and complexity. Specifically, PPL did not take effective corrective actions to correct an inadequate procedure for monitoring adherence to pressure/temperature (P/T) limits after it was identified by inspectors.
05000387/FIN-2013003-042013Q2SusquehannaImplementation of Enforcement Guidance Memorandum (EGM) 11-003, Revision 1From April 17 through May 17, PPL performed OPDRVs without establishing secondary containment integrity. An OPDRV is an activity that could result in the draining or siphoning of the RPV water level below the top of fuel, without crediting the use of mitigating measures to terminate the uncovering of fuel. TS 3.6.4.1, Secondary Containment requires that secondary containment be operable and is applicable during OPDRVs. The required action for this specification if secondary containment is inoperable in this condition of applicability is to initiate actions to suspend OPDRVs immediately. Therefore, failing to maintain secondary containment operability during OPDRVs without initiating actions to suspend the operation was considered a condition prohibited by TSs as defined by 10 CFR 50.73(a)(2)(i)(B). As reported in LER 05000388/2013-001, PPL conducted the following OPDRVs during the period of secondary containment inoperability: Recirculation system drain and maintenance; RWCU system removal from service, maintenance, testing and restoration; RHR system LLRT, drain, maintenance and testing; Hydraulic Control Unit (HCU) replacement; Local Power Range Monitor (LPRM) replacement; Scram Discharge Volume (SDV) maintenance and testing; CRD mechanism replacements; Scram Pilot Solenoid Valve testing; Reset of Reactor Scram; and Dynamic vent of CRD headers. NRC EGM 11-03, Enforcement Guidance Memorandum On Dispositioning BWR Licensee Noncompliance With TS Containment Requirements During Operations With A Potential For Draining The Reactor Vessel, provides, in part, for the exercise of enforcement discretion only if the licensee demonstrates that it has met four specific criteria during an OPDRV activity. The inspectors assessments of PPLs implementation of these four criteria during the LPRM replacement activity are described below: 1) The inspectors observed that, as required by the EGM, the OPDRV activities were logged in the control room narrative logs and that the log entries appropriately documented actions being taken to ensure water inventory was maintained and defense-in-depth criteria were in place. 2) The inspectors noted that the reactor vessel water level was maintained above the RHR high water level setpoint of 22 feet. The inspectors also noted that at least one safety-related pump was the standby source of makeup designated in the control room narrative logs for the evolutions. PPL logged that the worst case estimated time to drain the reactor cavity to the RPV flange was greater than the EGM criteria of 24 hours. 3) The inspectors verified that the OPDRVs were not conducted in Mode 4 and that PPL maintained secondary containment operability for the refueling floor while moving irradiated fuel during OPDRVs. The inspectors noted that PPL had contingency plans in place for isolating the potential leakage paths, should difficulty arise during the LPRM replacement activities. Additionally, the inspectors verified that two independent means of measuring RPV water level (one alarming) were available for identifying the onset of loss of inventory events. 4) Inspectors verified that, for the periods of April 17 through May 7 and May 10 through May 17, all other TSs were met during OPDRVs with secondary containment inoperable. For the period of May 7 through May 10, inspectors identified that the requirements of TS 3.8.2, AC Sources- Shutdown was not met when operators failed to adequately assess the operability of the EDGs and offsite power. Enforcement associated with performance deficiency is discussed in Section 1R15 of this report. TS 3.6.4.1 is applicable during OPDRVs and requires that secondary containment be operable. TS 3.6.4.1, action C.3, requires operators to initiate actions to suspend OPDRVs immediately upon discovery that secondary containment is inoperable. Contrary to the above, between 12:02 a.m. on April 17, 2013, and 12:53 a.m. on May 7, 2013, and between 3:32 a.m. on May 10, 2013, and 10:00 p.m. on May 17, 2013, PPL did not maintain secondary containment operable while performing OPDRVs. Because the violation was identified during the discretion period described in EGM 11-003, the NRC is exercising enforcement discretion in accordance with Section 3.5, Violations Involving Special Circumstances, of the NRC Enforcement Policy and, therefore, will not issue enforcement action for this violation. In accordance with EGM 11-003, each licensee that receives discretion must submit a license amendment request within 4 months of the NRC staffs publication in the Federal Register of the notice of availability for a generic change to the Standard TSs to provide more clarity to the term OPDRV. The inspectors observed that PPL is tracking the need to submit a license amendment request in its CAP as CR 1707662. This LER is closed.
05000387/FIN-2013008-012013Q2SusquehannaFailure to Restrict Operators from Performing Licensed Duties with Medically Disqualifying Conditions and Failure to Notify the NRC within 30 Days of Discovering Changes in Medical ConditionsThe inspectors identified: 1) an apparent violation (AV) of Title 10 of the Code of Federal Regulations (10 CFR) 55.21, Medical Examination; Part 55.25 Incapacitation because of disability or illness; Part 55.33, Disposition of an Initial Application, for the failure of the licensee to restrict operators from performing licensed duties when they had disqualifying medical conditions; and 10 CFR 50.74, Notification of change in operator or senior operator status, for PPLs failure to notify the NRC within 30 days of changes in licensed operators medical conditions; and, 2) a related finding of very low safety significance (Green) for PPLs failure to implement effective corrective actions to prevent this recurring AV. Specifically, the inspectors identified that four licensed operators developed disqualifying medical conditions that were not properly evaluated by PPL staff in accordance with ANSI/ANS-3.4-1983, American National Standard Medical Certification and Monitoring of Personnel Requiring Operator Licenses for Nuclear Power Plants. Additionally, PPL did not restrict the operators from performing licensed duties or obtain NRC approval (by requesting conditioned licenses) to continue to perform licensed duties, which caused the operators to not meet the requirements of 10 CFR 55.33(a)(1). Additionally, the inspectors identified eight instances in which PPL failed to notify the NRC within 30 days of learning of changes in licensed operator medical conditions that involved permanent disabilities/illnesses as required by 10 CFR 50.74. This resulted in the operators performing licensed operator duties without properly restricted licenses. PPL has taken actions to correct these issues by formally notifying the NRC and requesting conditioned licenses, as necessary, training the licensed operators and medical staff in the applicable requirements, and revising related procedures to provide additional guidance and require annual training. PPL entered this issue into their corrective action program. (CR-1709539) The inspectors reviewed this issue in accordance with NRC IMC 0612, Appendix B, Issue Screening for traditional enforcement and as part of the Reactor Oversight process (ROP). Under the ROP, the inspectors also identified a related finding of very low safety significance (Green) involving PPLs failure to prevent this recurring AV. Traditional Enforcement Basis for AV: The inspectors determined that PPLs failure to ensure that licensed operators met the license conditions associated with minimum medical qualifications prior to performing license activities and failure to notify the NRC of the changes within 30 days was a performance deficiency that was within PPL\\\'s ability to foresee and correct and should have been prevented. Specifically, the inspectors determined that four operators with disqualifying medical conditions continued to stand watch in the control room. Additionally, the NRC identified eight instances where changes occurred in licensee medical conditions that involved permanent disabilities and/or illnesses but were not reported within 30 days as required. The inspectors determined that Traditional Enforcement (TE) applies because the issue impacted the NRC\\\'s ability to perform its regulatory function. Namely, the NRC relies upon PPL to ensure all licensed operators meet the medical conditions of their licenses. 10 CFR 55.25 requires that if, during the term of the individual operator license, the operator develops a permanent physical or mental disability that causes the operator to fail to meet the requirements of 55.21 of this part, the facility licensee shall notify the Commission, within 30 days of learning of the diagnosis, in accordance with 10 CFR 50.74(c). PPL failed to properly identify that operators had disqualifying medical conditions and also did not restrict the operators from performing licensed duties. These issues are being characterized as an apparent violation in accordance with the NRCs Enforcement Policy, and its final significance will be dispositioned in separate future correspondence. (Section 4OA2.1) ROP Green Finding Basis: Since 2008, PPL has been issued three Severity Level (SL) IV violations and one SL III violation related to the medical qualifications of their licensed operators. PPL procedure, NDAP-QA-0702, ACTION REQUEST AND CONDITION REPORT PROCESS, Revision 39, defines a Significant Condition Adverse to Quality (SCAQ) as a condition determined to be significant enough to warrant a root cause analysis and actions to prevent recurrence. (For significant conditions adverse to quality, the causes of the deficiency shall be determined and corrective actions shall be taken to preclude recurrence). NDAP-QA-0702 Attachment M, lists a Severity Level III or greater NRC Notice of Violation (NOV), as an example of a SCAQ. Following the 2009 SL III NOV, PPL failed to identify the causes of the condition that led to the SCAQ, and the extent of cause and condition reviews were ineffective to identify additional issues. The inspectors determined that PPLs failure to implement adequate corrective actions to prevent this recurrence in response to these previous violations was an associated performance deficiency that was within PPL\\\'s ability to foresee and correct and should have been prevented. The performance deficiency is more than minor because, if left uncorrected, having operators stand watch with disqualifying medical conditions has the potential to lead to a more significant safety concern. The inspectors evaluated this finding using IMC 0609, Attachment 4, Initial Characterization of Findings, and IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions. All questions in Exhibit 2 were answered no and therefore this associated finding was determined to be of very low safety significance (Green). This issue is indicative of current performance and is determined to have a cross-cutting issue in the area of Problem Identification and Resolution - Evaluation P.1(c), related to thorough evaluation of problems. Specifically, PPLs reviews following the issuance of similar violations in 2009 and 2011, did not identify the additional cases discovered in 2012, and PPLs root cause evaluation completed in 2012, did not identify a root and several contributing causes, which were subsequently identified by the NRC inspectors.
05000387/FIN-2013008-022013Q2SusquehannaFailure to Establish and Implement Written Procedures for Operating Plant Equipment Failure to Provide Complete and Accurate Medical Information for Licensed Operator ApplicationsThe inspectors identified an AV of 10 CFR 50.9, Completeness and Accuracy of Information, related to PPL's failure to provide information to the NRC regarding medical examinations of licensed operators that was complete and accurate in all material respects. Specifically, PPL submitted three NRC licensed operator renewal applications and one initial license application, each of which certified the medical fitness of the applicants and that no restricting license conditions were necessary. However, the applicants, in fact, each had medical conditions that did not meet the minimum standards of 10 CFR 55.33(a)(1) and required license conditions to be in place in order for the operators to perform licensed activities. PPL entered this issue into their corrective action program. (CR-1709540) The inspectors determined that PPL?s failure to provide complete and accurate information to the NRC was a performance deficiency that was within PPL?s ability to foresee and correct and should have been prevented. The inspectors determined that TE applies, as the issue impacted the NRC's ability to perform its regulatory function. Specifically, the NRC issued new and/or renewed licenses to the operators based on information that was not complete and accurate in all material respects. The performance deficiency was screened against the ROP per the guidance of IMC 0612, Appendix B, lssue Screening. No associated ROP finding was identified and no crosscutting aspect was assigned. This issue constitutes an apparent violation in accordance with the NRC?s Enforcement Policy, and its final significance will be dispositioned in separate future correspondence.
05000387/FIN-2014002-012014Q1SusquehannaAdequacy of Compensatory Measures to Restore Technical Specification OperabilityAn Unresolved Item (URI) was identified because additional NRC review and evaluation is needed to determine whether implementation of a compensatory measure to restore TS operability required NRC approval prior to implementation and to subsequently determine whether a violation of 10 CFR 50.59, Changes, Tests and Experiments was more than minor. During a review of a prompt operability determination addressing the inadvertent closure of a main turbine CV, inspectors questioned whether a compensatory measure specified to maintain compliance with TS required NRC approval prior to implementation. Specifically, to address the degraded condition, PPL implemented a compensatory measure of crediting plant equipment not previously credited in the UFSAR to restore and maintain operability in accordance with TSs 3.2.2, Minimum Critical Power Ratio and 3.2.3, Linear Heat Generation Rate. PPL did not perform an evaluation of this change as required by 10 CFR 50.59(d)(1). On July 10, the number 3 main turbine CV on Unit 1, XV-10150C, slowly drifted close while operating at 100 percent rated thermal power (RTP). In response to the issue, the electro-hydraulic control system opened CVs 1, 2 and 4 to maintain reactor pressure stable. Operators reduced power to approximately 96 percent RTP. Operators also generated CR 1724394 and assessed the condition for operability. PPL performed a prompt operability determination and assessed, in part, the potential affect the degraded condition had on the power distribution limits. Specifically, PPL determined, during discussions with the fuel vendor, that the thermal limits were affected by the number 3 CV being closed. Specifically, with the number 3 CV closed, the steam relieving capacity of the main steam system was below assumed values in the transient analysis for a Recirculation Flow Controller Failure (RFCF). The RFCF is one of the limiting events used to develop the flow-based Minimum Critical Power Ratio and Linear Heat Generation Rate thermal limits. To compensate for this and restore operability per TSs 3.2.2 and 3.2.3, PPL specified crediting the reactor recirculation motor-generator set high speed electrical and mechanical stops to limit the maximum flow assumed in the transient analysis. Consistent with Inspection Manual Part 9900 Technical Guidance, Operability Determinations & Functionality Assessments for Resolution of Degraded or Nonconforming Conditions Adverse to Quality or Safety and the PPL 50.59 Resource Manual, Revision 6, PPL considered this compensatory measure a change to the facility and assessed whether the change required prior NRC approval in accordance with NDAP-QA-0726, 10 CFR 50.59 and 10 CFR 72.48 Implementation, and PPLs 50.59 Resource Manual. PPL determined the change did not require evaluation under 10 CFR 50.59 and documented this on a 50.59 Screening Determination. In part, this was based on answering no to the question does the proposed activity involve a change to an SSC that adversely affects an FSAR described design function. The basis for this determination was that the reactor recirculation motor-generator set high speed electrical and mechanical stops are not credited in the FSAR transient analysis and, therefore, have no design function. PPL considered the effect on the design function of the fuel assemblies to not fail during normal operation and anticipated operational occurrences and determined that the compensatory measure ensures that the requirement of the design function is met. PPL concluded that the change did not adversely affect any of the design functions for the fuel. Inspectors reviewed the 50.59 screening determination and questioned the basis of PPLs conclusion that an evaluation of the change was not required. Specifically, the change had the effect of creating a new design function for the reactor recirculation motor-generator set high speed electrical and mechanical stops to limit flow during a RFCF event. Additionally, a failure of these components could preclude the design function of the fuel from being met. The resource manual provides a definition of adverse effects which states, in part: Changes that would introduce a new type of accident or malfunction with a different result would screen in. If a proposed change would reduce the reliability of a design function, this change should be screened in because there is an adverse effect on a design function. Changes to SSCs that are not explicitly described in the FSAR can have the potential to affect SSCs that are explicitly described in the FSAR and thus may require a 10 CFR 50.59 Evaluation. If for the larger FSAR described SSC, the change affects a FSAR described design function or an evaluation demonstrating that intended design functions will be accomplished, then a 10 CFR 50.59 Evaluation is required. In this case, the introduction of a new design function for the components and reliance on these components to function to ensure a design function of the fuel was met had an adverse effect by introducing a new potential malfunction that could result in the design function not being met. Therefore, inspectors determined that PPL should have answered Yes to the screening question Does the proposed activity involve a change to an SSC that adversely affects an FSAR described design function. PPL also should have evaluated whether the change needed prior NRC approval in accordance with 10CFR 50.59(d)(1). Inspectors determined that the issue was a performance deficiency, however, could not determine whether the change would ever have ultimately required NRC approval. Therefore, in accordance with the NRC Enforcement Policy, inspectors could not determine whether the performance deficiency was more than minor. PPL entered the issue into the CAP as CR-2014-09397 and initiated actions to evaluate the change in accordance with 10 CFR 50.59(d)(1). Pending completion of PPLs 50.59 evaluation and review by inspectors, this is a URI.
05000387/FIN-2014002-032014Q1SusquehannaLicensee-Identified Violation10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires in part, that activities affecting quality shall be prescribed by documented instructions and procedures of a type appropriate to the circumstances. Contrary to the above, prior to July 6, 2012, it was identified that PPL had not incorporated adequate written guidance in TP-264-032, Core Flow Calibration, Revision 5, to require iteration of the procedural steps used for the calibration check if flow instrumentation summer gains were adjusted. This resulted in core flow for the A recirculation loop being adjusted to approximately 2.4 Mlb/hr below actual loop flow and the B recirculation loop being adjusted to approximately 0.2 Mlb/hr below actual loop flow. PPL entered the issue into the corrective action program as CR 1708878. Inspectors determined this finding to be of very low safety significance (Green) in accordance with IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process (SDP) for Findings At-Power, because none of the logic questions under the barrier integrity cornerstone applied, indicating the issue screened to Green. Inspectors reviewed PPLs technical evaluation and determined that there was adequate margin in the thermal limit calculations to ensure that no safety or operating limits were exceeded. This issue was discussed in further detail within Section 4OA3 of this report.
05000388/FIN-2014002-022014Q1SusquehannaReactor Scram due to Loss of Reactor Feed PumpsA finding of very low safety significance (Green) for failure to implement work instructions for an engineering change to the Integrated Control System (ICS) was self-revealed when Unit 2 lost control of reactor vessel level on September 14, 2013, requiring insertion of a manual scram. The cause of the loss of level control was determined to be a coding error in the ICS that resulted in the improper transition of feedwater control modes during a reactor shutdown. PPLs immediate corrective actions included entering the issue into their corrective action program (CAP) as condition report 1746169, correcting the coding error, and performing and extent of condition review of the ICS code to ensure no additional errors were present. The performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Initiating Events cornerstone and affected its objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the failure to implement work instructions associated with the engineering change resulted in an ICS logic code error which caused a loss of reactor feed requiring a manual reactor scram. The inspectors evaluated the finding in accordance with IMC 0609, Appendix A, The SDP for Findings At-Power, Exhibit 1 for the Initiating Events cornerstone. The inspectors determined the finding was of very low safety significance (Green) because it did not cause both a reactor trip and the loss of mitigation equipment. This finding was determined to have a cross-cutting aspect in the area of Human Performance, Work Management because PPL did not implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority, including the identification and management of risk commensurate to the work. Specifically, the work instructions associated with the engineering change lacked the specificity commensurate with the complexity of the work such that it could be accomplished without error.
05000388/FIN-2014008-012014Q3SusquehannaInadequately Maintained Procedures for Plant Shutdown to Hot Standby led to Reactor Water Level and Power TransientsA self-revealing Green NCV of Technical Specification (TS) 5.4.1, Procedures, was identified because PPL did not adequately maintain operating procedures for plant shutdown to hot standby. Specifically, the general operating procedure for plant shutdown to minimum power and the reactor feed pump (RFP) operating procedure were revised, and the technical reviews did not adequately verify the functional and technical adequacy of the procedures. The technical reviews did not identify a valve lineup conflict existed between the two procedures. The conflict resulted in an improper feed lineup to the reactor pressure vessel (RPV) causing two level transients and corresponding power transients of approximately five percent on March 20, 2014. PPLs corrective actions included mitigating the pressure vessel level transients, collecting personnel statements, revising the general operating procedure to remove the valve conflict, initiating an apparent cause evaluation to determine the cause of the level transients, resetting the Operations Department human performance clock due to operator performance issues during the event, reviewing the event with every Operations Department shift crew, performing a standdown for the Operations Department to compare operator performance issues between this event and the December 19, 2012 scram event, as well as entering the events of March 1, 2014, and March 20, 2014, into the corrective action program as condition report (CR)-2014-08941 and CR-2014-10388. The inspectors determined that PPLs inadequate maintenance of procedures for plant shutdown to hot standby was more than minor, because it is associated with the procedure quality attribute of the Initiating Events cornerstone 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. Specifically, the inadequate technical reviews associated with the revision of procedures for placing standby RFPs into service in startup level control and valve control (manual), and procedures for placing standby RFPs into service in the flow control mode (FCM) and valve control (manual) mode, resulted in two reactor power transients up to five percent and two significant reactor vessel water level transients which challenged the stability of the plant. Additionally, this issue is similar to Example 4b described in IMC 0612, Appendix E, Examples of Minor Issues, which states that issues are not minor if procedure issues cause a reactor trip or other transient. The inspectors evaluated the finding using Attachment 0609.04, "Initial Characterization of Findings," worksheet to IMC 0609, Significance Determination Process, issued June 2, 2011. The attachment instructs the inspectors to utilize IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, issued June 19, 2012. The inspectors determined this finding did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition (e.g. loss of condenser, loss of feedwater) and is therefore of very low safety significance (Green). A cross-cutting aspect was assigned in the area of Human Performance, Change Management, because leaders did not use a systematic process for evaluating and implementing change so that nuclear safety remained the overriding priority. Specifically, PPL did not maintain a clear focus on nuclear safety when implementing changes to the general operating procedure for shutdown to minimum power and this resulted in an unintended procedure discrepancy (H.3).
05000482/FIN-2014005-012014Q4Wolf CreekFailure to Conduct and Evaluate Simulator Testing In Accordance with ANSI/ANS 3.5-2009 and ANSI/ANS 3.5-1998The inspectors identified a Green finding for the inadequate conduct and evaluation of simulator performance testing in accordance with the standards of ANSI/ANS 3.5-2009 and ANSI/ANS 3.5-1998, Nuclear Power Plant Simulators for Use in Operator Training and Examination. Specifically, Wolf Creek Nuclear Operating Corporation (WCNOC) did not adequately identify that the simulator responses during 2008 through 2014 tests of Transient 3, Simultaneous Closure of All Main Steam Isolation Valves, did not meet the acceptance criteria described in Section 4.1.4 of ANSI/ANS 3.5-2009 (or the 1998 edition), which if left uncorrected, could have resulted in negative training of licensed operators and call into question Wolf Creeks ability to conduct valid licensing examinations with the simulator. WCNOC initiated condition reports 90179 and 90417 and simulator discrepancy report A14-154. WCNOC also plans to conduct benchmarking at other sites to compare simulator responses during applicable testing, and is evaluating the need for additional procedure revisions or other corrective actions. The performance deficiency is more than minor because it adversely impacted the human performance attribute of the mitigating systems cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Additionally, if left uncorrected, the performance deficiency could have become more significant in that not correcting noticeable differences between the simulator and the reference plant could cause negative training of licensed operators and call into question WCNOCs ability to conduct valid licensing examinations with the simulator. Utilizing Inspection Manual Chapter 0609, Significance Determination Process, Attachment 4, Tables 1 and 2 worksheets, issued June 19, 2012, and flowchart block 14 of Appendix I, Licensed Operator Requalification Significance Determination Process (SDP), issued December 6, 2011, the finding was determined to have very low safety significance (Green), because the deficiencies were associated with simulator testing, modifications, and maintenance, and there was no evidence that the plant-referenced simulator does not demonstrate the expected plant response or have uncorrected modeling and hardware deficiencies. This finding has a cross-cutting aspect in the area of problem identification and resolution, Identification, because WCNOC personnel did not implement a corrective action program with a low threshold for identifying issues. Specifically, this issue was first identified when the RETRAN-3D code analysis was first used in 2008 transient testing, and additional tests performed in 2008, 2009, 2010, and 2012 were opportunities to identify the performance deficiency; however, the issue was not entered into the corrective action program, a noticeable difference was not evaluated, a training needs assessment was not performed, and the process used to conduct simulator transient testing, as described in Procedure Al 30C-006, was not updated to include all of the minimum acceptance criteria described in the ANSI/ANS 3.5 standard. Hence, simulator issues expected to be identified during the testing process could potentially be missed by implementing the AI 30C-006 procedure, which did not include all of the minimum acceptance criteria described in the ANSI standard (P.1).
05000482/FIN-2014005-022014Q4Wolf CreekNotice of Enforcement Discretion 14-4-02 for Emergency Diesel Generator B Exciter Cabinet FireAn unresolved item (URI) is being opened to assess whether the cause for the request for enforcement discretion associated with the fire in the exciter circuit of emergency diesel generator B on October 6, 2014, involved a violation of NRC requirements. On October 6, 2014, at 1:26 p.m., emergency diesel generator B was declared inoperable when it tripped during a 24-hour surveillance test and operators identified a fire in an associated exciter cabinet. An Alert was declared and operators entered Technical Specification 3.8.1, AC Sources Operating, Required Action B.4.1, which required emergency diesel generator B be restored to operable status within 72 hours. Actions in response to the fire were completed, the fire was quickly suppressed, and WCNOC exited the Alert. Following the completion of repairs, WCNOC identified that postmaintenance testing required to demonstrate system operability included completing a 24-hour run. Since the postmaintenance testing and subsequent system restoration was expected to exceed the time remaining in the 72-hour action statement, WCNOC requested that the NRC exercise discretion to not enforce compliance with the actions required in Wolf Creek Generating Station Technical Specification 3.8.1, Required Action B.4.1, and approve an additional 8 hours to restore the system. NOED NO. 14-4-02, documents this request and the NRCs approval. Following postmaintenance testing, emergency diesel generator B was restored to operable status at 5:17 p.m. on October 9, 2014. WCNOC concluded that the most likely cause of the event was the failure of the power current transformers power rectifier bridge. WCNOC postulated that when the bridge failed, power from the power current transformers to the generator field was lost. As a result, the voltage regulator attempted to maintain the field current using only the power potential transformer. Since the power potential transformer is not rated to sustain full field current, the transformer was overloaded, which caused it to overheat and catch fire. Troubleshooting also indicated that the emergency diesel generator B tripped on phase differential current for the same reasons. WCNOC removed the failed rectifier bridge for further analysis in December 2014; at the end of the inspection period, WCNOC personnel were awaiting additional failure analyses of the failed rectified bridge to determine the specific direct causes of the fire and unplanned emergency diesel generator B inoperability. The root cause is being evaluated by Condition Report 88665. When an NOED is issued, Inspection Manual Chapter 0410, Notice of Enforcement Discretion, requires that a URI will be opened to assess whether the cause(s) of the events leading up to the request for the Notice of Enforcement Discretion involved violations of NRC requirements. This issue will be tracked as a URI in order to review and evaluate WCNOCs additional rectifier bridge failure analyses, root cause analysis, and other supporting documentation to determine if a violation exists: URI 05000482/2014005-02, Notice of Enforcement Discretion 14-4-02 for Emergency Diesel Generator B Exciter Cabinet Fire. These activities constitute completion of two event follow-up samples, as defined in Inspection Procedure 71153.
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