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| Title | Quarter | Description | |
|---|---|---|---|
| 05000293/FIN-2005003-02 | Inattentive Control Room Supervisor with Willful Inappropriate Response by Other Control Room Licensed Staff | 2005Q2 | In a letter dated July 14, 2005, the NRC issued a Severity Level III Notice of Violation and Proposed Imposition of Civil Penalty to Entergy in the base amount of $60,000 associated with a Severity Level III problem. The Severity Level III problem involved four violations of NRC requirements related to Technical Specification 5.4.1, 10 CFR Part 50 Appendix B, and 10 CFR Part 26. The specific violations involved: (1) a Pilgrim control room supervisor sleeping for approximately four minutes in the control room; (2) a reactor operator observing the sleeping control room supervisor, but deliberately not taking immediate actions to awaken the control room supervisor, inform appropriate site personnel and initiate a condition report; (3) a Shift Manager, in careless disregard of requirements, although taking some actions, not informing appropriate site personnel and initiating a condition report; and (4) the sleeping control room supervisor not being relieved of duty and for-cause Fitness-for-Duty tested. There were no actual safety consequences resulting from this event because there were no plant conditions that warranted immediate action. |
| 05000293/FIN-2006003-01 | the Inspectiors Identified a Severity Level Iv NON-CITED Violation Associated with the Failure to Perform an Adequate Safety Evaluation as Required by 10 CFR 50.59 for Changes Made to the Facility as Described in the UFSAR | 2006Q2 | Severity Level IV. The inspectors identified a Severity Level IV Non-Cited Violation associated with the licensees failure to perform an adequate safety evaluation per 10 CFR 50.59. Contrary to 10 CFR 50.59, a screening safety evaluation for handling of a 35 ton cask in the Reactor Building did not provide an adequate basis to demonstrate that the evaluation for use of a heavier cask did not change the evaluation methods approved by the NRC staff in 1985 for the control of heavy loads per NUREG 0612 commitments, as described in the UFSAR and the Pilgrim licensing basis. The licensee made significant enhancements to the original 50.59 safety evaluation and entered this issue into the corrective action program. The finding was determined to be more than minor because the inspectors could not reasonably determine that the methodology used to evaluate the use of a heavier cask did not constitute a change that would have required NRC approval. The conditions associated with the finding (i.e., the potential drop of a loaded cask) were determined to be of very low safety significance because they did not result in the loss of operability of a safety system. Because the issue affected the NRCs ability to perform its regulatory function, this finding was evaluated using the traditional enforcement process and was classified at Severity Level IV because the violation of 10 CFR 50.59 involved conditions evaluated as having very low safety significance by the SDP. This finding has a cross-cutting aspect in the area of human performance because Entergy did not fully evaluate the licensing basis to develop the 50.59 safety evaluation, and thereby failed to assure a design document was complete and accurate. |
| 05000293/FIN-2007003-04 | Application of TS 4.0.3 When it is Discovered that a Surveillance Has Never Been Performed | 2007Q2 | The inspectors questioned Entergy regarding the applicability of TS 4.0.3 given that the time response test had never been performed on four of the RPS scram contactors, as compared to missing a surveillance test following satisfactory initial system baseline testing that originally showed system operability. Entergy identified that industry guidance had been issued on this question in the form of Technical Specification Task Force (TSTF)-IG-06-01 dated May 2006. In the TSTF, the following question is asked, If it is discovered that a Surveillance has never been performed or has never been completely performed, can SR 3.0.3 be used? (Note: SR 3.0.3 is the standard TS equivalent of Pilgrim TS 4.0.3) The TSTF response states, Yes, SR 3.0.3 applies in conditions in which a Surveillance has never been performed or has never been performed completely provided that there is an expectation that the Surveillance will pass when it is performed and that the associated system is OPERABLE. An Unresolved Item (URI) will track NRC evaluation of this industry guidance to determine if additional NRC guidance is necessary to specify when TS 4.0.3 applies in the case of a missed surveillance where it is determined that the surveillance was never originally performed to establish initial system operability. (URI 05000293/2007003-04, Application of TS 4.0.3 When it is Discovered that a Surveillance Has Never Been Performed) Entergy modified the applicable surveillance procedures and successfully response time response tested all RPS scram contactors |
| 05000293/FIN-2008002-01 | Inadequate Risk Assessment for Emerency Maintenance on A5 Emergency Bus Undervoltage Relays | 2008Q1 | The inspectors identified a Green non-cited violation (NCV) of 10 CFR 50.65(a)(4) for Entergy\'s failure to conduct an adequate risk assessment for emergent maintenance on the A5 Emergency Bus undervoltage relays. Specifically, the inspectors noted that Entergy had downgraded an on-line risk assessment from Red to Green without a valid technical basis and did not recognize the unavailability of the automatic function of the Emergency Diesel Generator (EDG); as a result, Entergy did not evaluate or specify risk management actions. This finding is more than minor because the risk assessment had incorrect assumptions that changed the outcome of the assessment. The inspectors conducted a screening in accordance with IMC 0609, Significance Determination Process, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process. The finding was determined to be of very low safety significance (Green) because the Incremental Core Damage Probability Deficit for the timeframe that the relays were removed from service was significantly less than 1E-6 due to the short amount of time the EDG was unavailable in the automatic mode and the reasonable assurance that operators could manually tie the EDG to the bus in the event of a Loss Of Offsite Power. This finding has a cross-cutting aspect in the area of Human Performance, Decision Making, because Entergy did not use a systematic process to make a risk-significant decision, when faced with an unexpected plant condition. (H.1(a)) (Section 1R13 |
| 05000293/FIN-2008003-01 | Licensee-Identified Violation | 2008Q2 | TS 3.7.C, Secondary Containment, requires, in part, that whenever the reactor is critical, secondary containment integrity must be maintained. Contrary to this, on January 10, 2008, damper AO-N-78, the Reactor Building Isolation Control System ventilation damper, failed to go fully closed on demand during testing, resulting in secondary containment integrity not being maintained. TS actions to restore operability or to secure the damper were not taken for approximately 96 hours. This was identified in Entergys CAP as CR-PNP-2008-140 and CR-PNP-2008-143. This violation is of very low safety significance (Green) because it did not represent an open pathway in secondary containment due to the availability of a backup isolation damper |
| 05000293/FIN-2008003-02 | Licensee-Identified Violation | 2008Q2 | TS 3.2.B, Instrumentation and Control of Core and Containment Cooling Systems, requires undervoltage relays 127-A5/1 and 127-A5/2 trip settings to be within a range of 20-25 percent of rated bus voltage. Contrary to this, on January 11, 2008, the as-found set points were 27.17 percent and 30.29 percent, respectively, of rated bus voltage. This was identified in Entergys CAP as CR-PNP-2008-00117. This finding is of very low safety significance (Green) because it did not represent a loss of function of the 4kV safety related bus or associated safety related loads because the relay trip settings were within the acceptable limit per industry standard Handbook on Electric Motors |
| 05000293/FIN-2008005-01 | Failure to Conduct a Risk Assessment for Emergent Maintenance on the High Pressure Coolant Injection System | 2008Q4 | The inspectors identified a Green non-cited violation (NCV) of 10 CFR 50.65(a)(4) for Entergys failure to conduct a risk assessment for emergent maintenance on the High Pressure Coolant Injection (HPCI) system injection valve. Specifically, the failure to conduct a risk assessment resulted in Entergy not recognizing an increase in risk to a Yellow condition, and therefore no risk management actions were taken. Entergy entered this issue into their corrective action program. Corrective actions will include revising attachments in Entergys Technical Specification requirements procedure to perform a risk review as a result of emergent maintenance activities. This finding was more than minor because Entergy failed to consider the unavailability of a risk significant system where the outcome of the risk assessment would have been a change in a risk management category. The inspectors conducted an evaluation in accordance with IMC 0609, Significance Determination Process, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process. The finding was determined to be of very low safety significance (Green) because the Incremental Core Damage Probability Deficit for the timeframe that HPCI was removed from service was significantly less than 1E-6. The inspectors determined that this finding had a cross-cutting aspect in the area of Human Performance, Decision Making, because Entergy did not use a systematic process to make a risk-significant decision when faced with an unexpected plant condition |
| 05000293/FIN-2008005-02 | Procedural Error Resulting in Unplanned RCIC Isolation | 2008Q4 | A self-revealing Green non-cited violation (NCV) of TS 5.4.1, Procedures , was identified for a procedure which resulted in an inadvertent isolation of the Reactor Core Isolation Cooling (RCIC) system. Specifically, the procedure was previously revised and a step was inadvertently placed out-of-order. The procedure incorrectly instructed technicians to remove relay contact blockers, or boots , before clearing an isolation signal which resulted in the system isolation. Entergy entered this issue into their corrective action program. Corrective actions will include revising this procedure and reviewing other surveillance procedures that had been revised at the same time. This finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone. Isolating the RCIC system affected the cornerstone objective of ensuring the availability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated this finding using IMC 0609.04, Phase 1 Initial Screening and Characterization of Findings . This finding was of very low safety significance because it was not a design or qualification deficiency, did not represent a loss of system safety function, did not represent an actual loss of a single train system for greater than the Technical Specification allowed outage time, and was not made risk-significant because of external events. The inspectors determined that this finding had a cross-cutting aspect in the area of Human Performance, Resources, because Entergy did not ensure that the procedure was complete and accurate |
| 05000293/FIN-2008005-03 | Application of TS 4.0.3 When It Was Discovered That a Surveillance Had Never Been Performed | 2008Q4 | On June 25, 2007, Entergy informed the Nuclear Regulatory Commission (NRC) staff that it had missed a Technical Specification (TS) surveillance requirement to perform time response testing of four Reactor Protection System (RPS) scram contactors. During their review, Entergy identified that the four RPS scram contactors had never been tested. Entergy evaluated the operability of the RPS system and determined that the system remained operable and that TS 4.0.3, Surveillance Requirement Applicability, would allow a delay period up to the limit of the specified surveillance frequency. The inspectors questioned Entergy regarding the applicability of TS 4.0.3 given that the time response test had never been performed on the RPS scram contactors, as compared to missing a surveillance test following satisfactory initial system baseline testing that originally showed system operability. As a result of this implementation of TS 4.0.3, Entergy failed to take action in accordance with TS 3.1, Reactor Protective System, which constituted a violation of NRC requirements. Entergy later modified the applicable surveillance procedures and successfully response time tested all RPS scram contactors. In Task Interface Agreement (TIA) 2008-004, the NRC staff disagreed with Entergy on its implementation of TS 4.0.3 and considered Entergy to have been in violation of TS 3.1, Reactor Protection System, as a result. Discretion is warranted because: (1) licensee current basis documents do not specifically clarify the distinction between a missed surveillance and one that has never been performed, (2) the licensee subsequently completed the surveillance testing satisfactorily, and (3) the issue was of very low safety significance, since when the correct testing was accomplished, it was completed satisfactorily indicating that the timing of the reactor scram function was not negatively impacted. Accordingly, the NRC staff is exercising enforcement discretion for the TS 3.1 violation in accordance with Section VII.B.6 of the NRC Enforcement Policy and no violation will be issued. |
| 05000293/FIN-2008007-01 | Inadequate Corrective Actions for \"B\" Battery Charger Circuit Breaker | 2008Q2 | The team identified a finding of very low safety significance involving a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, in that, Entergy did not promptly correct deficiencies with the B 125 Vdc battery charger supply circuit breaker. Specifically, Entergy did not properly evaluate and take adequate corrective actions for a condition adverse to quality associated with elevated temperatures on the circuit breaker terminals; and subsequently, the circuit breaker failed when recharging the B 125 Vdc battery. Entergy entered the issue into their corrective action system, completed an operability assessment, and reviewed installed circuit breakers to ensure a similar condition did not exist. The finding is more than minor because the degraded condition represented reasonable doubt on the operability of the B 125 Vdc charger and its associated breaker. The finding is 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. The team determined the finding was of very low safety significance (Green) because it did not result in the loss of system safety function. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program Component, because Entergy did not adequately evaluate the condition adverse to quality, which they originally identified in January 2006 (IMC 0305, Aspect P.1(c)). (1R21.2.1.1 |
| 05000293/FIN-2008007-02 | Inadequate Corrective Actions in Response to an Intake De-watering Event. | 2008Q2 | The team identified a finding of very low safety significance involving a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, in that, Entergy did not properly evaluate and take adequate corrective actions for a condition adverse to quality associated with an intake de-watering event on September 14, 2007. Specifically, a fish intrusion event resulted in a significant lowering of intake level and challenged the continued availability of the A loop of salt service water (SSW). Entergys issue prioritization, operability review, and subsequent evaluation did not adequately assess and correct the plant response relative to the safety-related SSW design and licensing bases. Entergy entered the issue into their corrective action system, implemented short-term corrective actions, and completed an operability assessment for the affected equipment. This finding is more than minor because it is associated with the external factors attribute (loss of heat sink) for the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The team determined the finding was of very low safety significance (Green) because it did not result in the loss of system safety function. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program Component, because Entergy did not adequately evaluate a condition adverse to quality, including properly classifying, prioritizing, and evaluating for operability (IMC 0305, Aspect P.1(c)). (1R21.2.1.2 |
| 05000293/FIN-2008007-03 | Inadequate Design Control for Switchyard Voltage Criteria | 2008Q2 | The team identified a finding of very low safety significance involving a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, in that, Entergy did not perform a calculation to demonstrate that the switchyard voltage used in procedures was adequate. Such a calculation was necessary to ensure that a spurious loss of the preferred offsite power source during transient conditions would not occur. Entergy entered this issue into their corrective action system, and demonstrated there was sufficient margin to assure operability of the preferred offsite power source. This finding is more than minor because the deficiency represented reasonable doubt on the operability of the preferred offsite power system. The finding is associated with the design control attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team determined the finding was of very low safety significance (Green) because it was a design deficiency confirmed not to result in a loss of the preferred offsite power source. (1R21.2.1.9 |
| 05000293/FIN-2008007-04 | Non-Conservative Calculation for SSW Pump Minimum Flowrate | 2008Q2 | Green. The team identified a finding of very low safety significance involving a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, in that, Entergy did not properly translate design basis parameters into specifications and procedures for the salt service water (SSW) system. Specifically, the system hydraulic analysis did not establish a system leakage limit, and the plant operating procedures allowed alignments that could have led to a condition where system leakage could have been in excess of the available margin. Entergy entered this issue into their corrective action system and instituted immediate corrective actions. This finding is more than minor because the deficiency represented reasonable doubt on the operability of the SSW system. The finding is associated with the design control attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team determined the finding was of very low safety significance (Green) because it was a design deficiency confirmed not to result in a loss of the SSW system. (1R21.2.1.11 |
| 05000293/FIN-2009002-01 | Failure to Establish and Maintain Adequate Design Measures to Monitor Critical Parameters of the EDG Air Start System | 2009Q1 | The inspectors identified a non-cited violation (NCV) of very low safety significance of 10 CFR Part 50, Appendix B, Criterion III, Design Control, because Entergy personnel did not establish and maintain measures to monitor critical design parameters to assure that equipment and processes essential to the safety-related function of the emergency diesel generator (EDG) air start system were adequate. Specifically, Entergy did not establish adequate measures to assure that an adequate supply of air was available to the air receivers for a minimum of two cold engine starts without recharging. This resulted in the A EDG being inoperable on March 8, 2009.Entergy entered this issue into their corrective action program (CAP) for resolution asCR-PNP-2009-00807. The immediate corrective actions included establishing compensatory requirements to increase the monitoring frequency for the air start system critical parameters. This finding is more than minor because it is associated with the design control attribute of the Mitigating Systems cornerstone and affects the cornerstone objective to ensure the reliability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the finding affected the reliability of the EDG to ensure a minimum of two cold engine starts without recharging to help mitigate the consequences of design basis events. The inspectors determined that the finding is of very low safety significance (Green) because it is not a design or qualification deficiency, did not represent a loss of safety function, and did not screen as potentially risk significant due to external events. There is no cross-cutting aspect identified for this finding because the inspectors determined that the performance deficiency is not reflective of current plant performance. The monitoring frequencies of the EDG air start system critical parameters were established for an extended period and prior to this problem there had not been recent issues with monitoring EDG air start capability |
| 05000293/FIN-2009002-02 | Failure to Implement Scaffolding Procedure Requirement | 2009Q1 | The inspectors identified an NCV of very low safety significance of Technical Specification 5.4.1 Procedures, because Entergy personnel did not adequately implement procedure requirements in accordance with EN-MA-133, Control of Scaffolding. Specifically, personnel did not erect scaffold in accordance with procedure EN-MA-133 and maintain the minimum distance erection requirements for safety-related equipment or alternatively perform engineering evaluations that concluded the equipment will not be impacted by the scaffolds. Entergy entered this issue into their CAP for resolution as CR-PNP-2009-00064, implemented prompt actions to correct the scaffolds, and performed engineering evaluations to assess the affect of the scaffolds on the safety-related equipment. The finding is more than minor because it is associated with the external factors attribute of the Mitigating Systems cornerstone and affects the cornerstone objective to ensure the reliability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Additionally, the finding is similar to example 4.a in Appendix E of IMC 0612 in that personnel did not routinely perform engineering evaluations for scaffolds constructed less than the minimum allowed distance to safety-related equipment. The inspectors determined that the finding is of very low safety significance (Green) because the scaffold issues identified were not a design or qualification deficiency, did not represent a loss of safety function, and did not screen as potentially risk significant due to external events. This finding has a cross-cutting aspect in the area of Human Performance because Entergys supervisory and management staff did not provide adequate oversight of workers or communicate expectations to workers to ensure scaffold erection requirements were fully understood (H.4.c of IMC 305) |
| 05000293/FIN-2009003-01 | Failure to Follow Procedure Resulting in Damage to Refueling Mast | 2009Q2 | A self-revealing non-cited violation (NCV) of Technical Specification 5.4.1 Procedures, was identified, because Entergys refueling bridge operators did not continuously monitor a Double Blade Guide (DBG) that was moved into the core to ensure the DBG did not encounter any obstructions, interferences, or other abnormal indications required by Pilgrim Procedure 4.3, Revision 113, Fuel Handling. Specifically, the failure to follow the procedure resulted in damaging the refueling mast when the mast was moved and still latched to the DBG. Entergy entered this issue into their corrective action program as CR-PNP-2009-2083. Corrective actions included replacing a section of the refueling mast, replacing the grapple camera, conducting additional training with the refueling crews including a table top dry run, performing a Human Performance Error Review and requiring Operations Senior Management to provide oversight during one hour of each three hour shift when the refueling crew was on the bridge moving fuel. The inspectors determined that the finding was more than minor because the finding was associated with the human performance attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstones objective to provide reasonable assurance that physical design barriers (i.e. fuel cladding) protect the public from radionuclide releases caused by accidents or events. The risk significance of the performance deficiency was determined to be of very low safety significance (Green) using IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings. Specifically, since the finding did not affect spent fuel pool cooling or inventory and since no fuel or control rod was damaged when the mast was bent, the finding was determined to be of very low safety significance. The finding has a cross cutting aspect in Human Error Prevention Techniques in the Work Practices component of the Human Performance area. Specifically, Entergy did not employ effective self and peer checking techniques such that refueling activities were performed safely. |
| 05000293/FIN-2009004-01 | Failure to Include Security Diesel Generator into the Maintenance Rule Scoping Document | 2009Q3 | The inspectors identified a non-cited violation (NCV) of very low safety significance (Green) of 10 CFR Part 50.65 paragraph (b), Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, because Entergy did not include all aspects of the emergency lighting system into the Pilgrim Maintenance Rule scoping document. Specifically, Entergy did not include the security diesel generator in the scoping document, which provides backup power to emergency yard lighting, and is required to meet Appendix R emergency lighting requirements. Entergy has entered the issue into their corrective action program (CAP) to add the security diesel generator and normal power supplies for yard emergency lighting into the Maintenance Rule scoping document. The finding is more than minor because it is associated with the Procedure Quality attribute of the Mitigating Systems cornerstone, in that, the issue affected emergency lighting reliability in support of the accomplishment of EOPs. The finding was determined to be of very low safety significance (Green) because the finding did not involve a design or qualification deficiency resulting in loss of operability or functionality, did not result in a loss of system safety function, and did not screen as potentially risk significant due to external initiating events. The finding does not have a cross-cutting aspect since the failure to scope this equipment into the Maintenance Rule was not recognized during the initial Maintenance Rule scoping activities and as a result, is not indicative of current performance. |
| 05000293/FIN-2009004-02 | Inadequate Procedures Result in the Failure to Evaluate Operability of the B RBCCW/SSW Heat Exchanger | 2009Q3 | The inspectors identified a NCV of very low safety significance (Green) of 10 CFR Part50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because Entergy procedures directed operators to take corrective actions for degra.ded conditions prior to assessing operability of the affected system. Specifically, operators conducted corrective actions (backwashing) of the B RBCCWISSW heat exchanger (HX) prior to assessing operability when the HX failed to meet the procedural differential pressure (dP) acceptance criteria. Entergy entered this issue into their CAP (CR-PNP-2009-03596) and performed a past operability evaluation which showed that the HX would have been able to meet its intended function during accident conditions. The finding is more than minor because if left uncorrected, it has the potential to lead to a more significant safety concern. The finding was determined to be of very low safety significance (Green) because the finding did not involve a design or qualification deficiency resulting in loss of operability or functionality, did not result in a loss of system safety function, and did not screen as potentially risk significant due to external initiating events. The inspectors determined that this finding had a cross-cutting aspect in the Resources component of the Human Performance cross-cutting area because Entergy did not provide procedures adequate to ensure nuclear safety. H.2(c |
| 05000293/FIN-2009004-03 | Failure of the A Standby Liquid Control Train | 2009Q3 | An unresolved item (URI) was opened related to the failure of the Standby Liquid Control (SBLC) A pump relief valve during system testing. The performance deficiency cannot be determined until the definitive causers) of the issue are known. On July 10, 2009, during the quarterly surveillance on the A SBLC train, the pump relief valve, PSV-1105A, lifted and failed to reseat, which diverted flow such that the system could not meet its TS acceptance criteria. The train was declared inoperable, the relief valve was replaced, and the system was restored to service on July 12, 2009. This issue has been entered into Pilgrim\'s CAP (CR-PNP-2009-03088) and an apparent cause evaluation (ACE) was conducted. However, Entergy has determined that the ACE may not definitively address the causers) of the SBLC train failure. The inspectors require Entergy\'s final ACE in order to evaluate whether or not a performance deficiency exists.URI 05000293/2009004-03, Failure of the A Standby Liquid Control Train |
| 05000293/FIN-2009004-04 | Human Error Resulting in Unplanned High Pressure Coolant Injection (HPCI) Isolation | 2009Q3 | A self-revealing, NCV of very low safety significance (Green) of Technical Specification (TS) 5.4.1, Procedures, was identified for a procedure error which resulted in the inadvertent isolation of the High Pressure Coolant Injection (HPCI) system. Specifically, Entergy procedure 8.M.2-2.6.3, HPCI High Steam Line Temperature, which describes the conduct of continuity checks of temperature switches, was not adequately implemented and caused the HPCI system to isolate. Corrective actions have included revising the procedure to include a step requiring concurrent verification for resetting the temperature switch, and a wait time of five minutes before Entergy proceeds to test the next switch. This finding is more than minor because it is associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability of systems that respond to initiating events to prevent undesirable consequences. The finding was determined to require a Phase II analysis because the finding resulted in an actual loss of system safety function. Using the Pilgrim pre-solved initiating event sequences and an exposure time of less than three days with the HPCI system unavailable, the Phase II estimation determined this finding was of very low safety significance (Green). The inspectors determined that this finding had a cross-cutting aspect in the Work Practices component of the Human Performance cross-cutting area because Entergy did not conduct effective self or peer checks to ensure that continuity checks were adequately performed. H.4(a |
| 05000293/FIN-2009005-01 | Incomplete licensed operator medical examinations | 2009Q4 | A Severity level lV violation (VIO) of 10 CFR 50.9, Completeness and Accuracy of Information, was identified due to the submittal of inaccurate medical information for licensed operators. The submittals to the NRC were inaccurate because they certified that the operators had been medically examined and had met all medical qualifications, when in fact, olfactory testing to detect odor of products of combustion had not been performed. The facility has completed corrective actions to develop and administer an appropriate test. All licensed operators passed this new test, and no new license conditions were required. The licensee\'s medical physician failed to adequately test all licensed operators (both initial and renewal licensees) in accordance with 10 CFR 55.21 and 55.33 with respect to ANSI/ANS-3.4 1983. The licensee submitted medical information for its licensed operators and applicants that was incomplete and incorrect in its assessment of the medical condition and general health of its licensed operators and initial applicants. The licensee\'s failure to provide complete and accurate information to the NRC, which could have resulted in an incorrect licensing action, is a performance deficiency because the licensee is expected to comply with 10 CFR 50.9, and because it was within the licensee\'s ability to foresee and prevent. Because violations of 10 CFR 50.9 are considered to be violations that potentially impede or impact the regulatory process, they are dispositioned using the Traditional Enforcement process. The applicability of crosscutting aspects related to the performance deficiency of this finding will be determined after NRC review of Entergy\'s response to the Notice of Violation. |
| 05000293/FIN-2009005-02 | Procedure change to allow disabling HPCI during transients | 2009Q4 | The inspectors identified an Unresolved Item (URI) involving10 CFR 50.59, Changes, Tests and Experiments, due to a failure to properly implement a procedure change which may have resulted in a more than minimal increase in the likelihood of occurrence of a malfunction of a structure, system or component (SSC)important to safety.... Pilgrim incorrectly allowed an existing 50.59 evaluation to support a High Pressure Coolant Injection (HPCI) procedure change that allowed actions that were beyond the scope of the existing 50.59 evaluation. In February 2000, Pilgrim revised procedure 2.2.21.5, HPCI Injection and Pressure Control, to provide procedural direction when shutting down the HPCI system. One of the changes that were made in this revision was the addition of a new section, Section 8, Preventing HPCl lnjection, to the procedure. These changes were evaluated using NOP83E5, 10 CFR 50.59 Process, to evaluate whether this change was allowed per 10 CFR 50.59 regulations. Pilgrim concluded that the change to the facility could be made without a safety evaluation or license amendment. During the performance of a simulator exam scenario, the examiners observed the HPCI system being defeated as drywell pressure approached the automatic initiation setpoint for the HPCI system (2.2 pounds per square inch gage (psig. The system was defeated by a Reactor Operator placing the HPCI oil pump in Pull-to-Lock (PTL). This action prevents HPCI from starting in response to Emergency Safeguards Feature (ESF) .automatic initiation signals. The order to defeat HPCI was made before the automatic initiation setpoint was reached, which is also the required entry into the Emergency Operating Procedures (EOPs). The examiners requested the procedural guidance that directed this action, since the Emergency Operating Procedures (EOPs) had not yet been entered when HPCI was defeated. The examiners reviewed the revised procedure as well as the procedure change paperwork and 50.59 preliminary evaluation checklist developed to support the revised procedure. The examiners concluded that the 50.59 preliminary evaluation checklist developed to support the revision to procedure 2.2.21.5 was incorrect and did not support the procedure revision. The basis for the procedure change was to provide enhanced instructions for the operation of the HPCI System under various emergency operating modes. Several of these operations, such as pressure control and placing HPCI to \'inhibit\' are required to be performed during the execution of various EOPs. These evolutions are analyzed in Emergency Procedure Guidelines (EPGs) Rev. 4 and have been approved by the NRC per SER 1.88.196.Since the procedure change can be used outside the EOPs, the EPGs and SER 1.88.196 do not fully support the conditions under which the HPCI system may be secured. Therefore, the 50.59 preliminary evaluation checklist was incorrect, and a separate safety evaluation was required to allow the HPCI system to be secured .outside the EOPs. This issue remains unresolved until the facility completes their development of a new safety evaluation to determine whether prior NRC\'s approval would have been required before implementing the described procedure change. URI05000293/2009005-02, Procedure change to allow disabling HPCI during transients. |
| 05000293/FIN-2009005-03 | Inadequate Surveillance Procedure Resulting in Failed Standby Liquid Control Train | 2009Q4 | A self-revealing, non-cited violation (NCV) of very low safety significance (Green) of Technical Specification (TS) 5.4.1, Procedures, was identified for inadequate procedural guidance which resulted in repeated lifting of the A Standby Liquid Control (SBlC) system relief valve and the subsequent failure of the A SBlC system. Specifically, the SBlC system test procedure did not provide precautions or identify methods to avoid exceeding the pressure set point of the system relief valve during testing. The issue was entered into the corrective action program and the surveillance procedure was revised to add cautions against exceeding 1300 psig and to reduce the test pressure window upper limit. In addition, if 1350 psig is exceeded, a condition report must be written to evaluate the impact on the system. Corrective actions are also planned to increase the relief valve design set point and to replace the test throttle valve with one more suited to adjusting system pressure. The performance deficiency was that Entergy did not specify adequate test controls to ensure that SBLC system relief valve set points were not challenged during test performance. This led to repeated relief valve lifts which over time contributed to the degradation of the relief valve that rendered the A SBLC train inoperable. The inspectors determined that the finding was more than minor because the finding was associated with the Procedure Quality attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone\'s objective to ensure the availability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, one train of SBLC was unavailable for several days. Using Inspection Manual Chapter 0609, Significance Determination Process, Attachment0609.04, Phase 1-lnitial Screening and Characterization of Findings, the inspectors determined that the finding is of very low safety significance because it is not a design or qualification deficiency, did not represent a loss of system safety function, did not represent an actual loss of a single train for greater than its TS allowed outage time and was not made risk significant because of external events. This finding has a crosscutting aspect in the Human Performance cross-cutting area, Resources component, because Entergy did not provide complete procedures. Specifically, the procedure did not include precautions and/or techniques to avoid exceeding the relief valve set point during testing. H.2(c |
| 05000293/FIN-2009007-01 | Failure to establish adequate procedures to prevent adverse impact due to spurious valve closure casued by fire damage | 2009Q4 | The inspectors identified that Entergy did not ensure that plant procedures were adequate to prevent a spurious actuation from affecting the ability to provide a source of makeup water to the reactor vessel within 20 minutes following the evacuation of the control room during a fire as specified in procedure 2.4.143, Shutdown From Outside the Control Room, Revision 40. The finding was determined to be of very low safety significance (Green) and a NCV of the Pilgrim Nuclear Power Station Technical Specification 5.4.1. d, Procedures. Entergy entered the issue into the corrective action program and planned to implement changes to the procedure to resolve the issue. Entergy also reviewed completed reactor core isolation cooling(RCIC) and high pressure coolant injection (HPCI) system startup job performance measures(JPMs) and performed procedure walkthroughs to assess the time needed to attempt a RCIC start and then transfer to, and start HPCI to confirm these actions could be taken in within the time necessary to prevent lowering vessel level to that of the top of active fuel. The inspectors determined that this finding was more than minor because it was associated with the procedure quality attribute of the mitigating system cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences (Le. core damage). Specifically, during a fire in the control room or cable spreading room there are four valves in each of the RCIC and HPCI systems that could spuriously close due to fire damage to cabling. Procedure 2.4.143 does not ensure that the associated motor control center circuit breakers are opened (to prevent spurious closure) and that the valves are in the correct position prior to starting one of the systems to provide make-up to the reactor vessel. Unidentified spurious closure of the valves during or after startup of the systems could disable the system and delay establishment of reactor vessel makeup. The inspectors assessed this finding in accordance with NRC IMC 0609, Appendix F, Fire Protection Significance Determination Process. This finding affected safe shutdown capabilities and screened to very low safety significance (Green) in Phase 1 of the SDP because it was assigned a low degradation rating. A low degradation rating was assigned because it was determined to be a minor procedure issue that could be compensated for by operator experience and familiarity. No cross-cutting aspect was assigned because the inspectors concluded this issue was not indicative of current licensee performance. |
| 05000293/FIN-2009007-02 | Licensee-Identified Violation | 2009Q4 | License Condition 3.F for Pilgrim Nuclear Power Station (PNPS) states in part that, Entergy shall implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report for the facility and as approved in the Safety Evaluation Report (SER) dated December 21, 1978. The SER, states in part, an analysis of safety related shutdown systems will be conducted to demonstrate that fire-related damage in any fire area will not inhibit the capability to safely shutdown. Contrary to this requirement, Entergy calculation PS32, Appendix R Safe Shutdown Analysis Report, incorrectly assumed that the reactor head vent valves (SV220-46 and SV220-47) would not be vulnerable to spurious operation (opening) for a fire in the control room or cable spreading room and would remain closed while the reactor was at power. During review of a recent modification, the licensee determined that the head vent valves could spuriously open due to a control room or cable spreading room fire during a potential control room evacuation scenario. The issue was entered into Entergy\'s corrective action program (CR 2009-01376) and corrected by implementation of a modification (EC 14506) to the head vent valve circuits. PS32 was subsequently updated to reflect this change. The inspectors determined that the finding was of very low safety significance(Green) because appropriate plant procedures were in place to operate systems required for post-fire safe shutdown during a control room evacuation scenario and those systems would be have been able to provide adequate makeup to the reactor vessel to compensate for inventory loss through the head vent piping. |
| 05000293/FIN-2009403-01 | Security | 2009Q4 | |
| 05000293/FIN-2009403-02 | Security | 2009Q4 | |
| 05000293/FIN-2010002-01 | Failure to Implement Operability Determination Process and Temporary Modification Process for Compensatory Measures Required to Maintain Operability of Secondary Containment | 2010Q1 | The NRC identified a Green non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for Entergy\'s failure to accomplish procedures prescribed for activities affecting quality. Specifically, Entergy did not implement their operability determination process or their temporary modification process for compensatory measures needed to maintain the secondary containment operable. Entergy\'s corrective actions included designating the compensatory measures as necessary to maintain operability for both torus troughs and implementation of temporary modifications for the equipment installed in the plant to support these compensatory measures. The inspectors determined that the finding was more than minor because the finding was associated with the Human Performance attribute of the Barrier Integrity cornerstone, and adversely affected the cornerstone\'s objective to provide reasonable assurance that physical design barriers (containment) protect the public from radionuclide releases caused by accidents or events. Specifically, operations and engineering personnel did not adequately implement operability determination and temporary modification procedures when degraded and/or non-conforming conditions associated with the secondary containment torus troughs were identified. The inspectors determined the significance of the finding using IMC 0609.04, Phase 1 -Initial Screening and Characterization of Findings. The finding was determined to be of very low safety significance (Green) because the finding only represented an impact to the radiological barrier function provided by secondary containment and the standby gas treatment system. This finding had a cross-cutting aspect in the area of Human Performance, Work Practices component, because Entergy personnel did not follow procedures. Specifically, Entergy did not implement their operability determination or temporary modification procedures for compensatory measures needed to maintain the secondary containment operable. H.4(b |
| 05000293/FIN-2010002-02 | Inadequate Corrective Actions to Promptly Correct Leaking Snubber Valves on the A Emergency Diesel Generator | 2010Q1 | The NRC identified a Green non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for Entergy\'s failure to promptly correct a condition adverse to quality. Specifically, Entergy did not correct defective material in their A Emergency Diesel Generators (EDG) in a prompt manner which led to emergent maintenance and additional unplanned unavailability of the A EDG while they replaced cracked snubber valves. Entergy\'s corrective actions include entering this issue into the corrective action program and replacing the seven remaining snubber valves on their AEDG with those of a material properly hardened and not susceptible to the same mode of cracking. The inspectors determined that the finding was more than minor because the finding was associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone, and adversely affected the cornerstone\'s objective to ensure the availability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the A EDG was unavailable during snubber valve replacements. The inspectors determined the significance of the finding using IMC 0609.04, Phase 1 -Initial Screening and Characterization of Findings. The finding was determined to be of very low safety significance (Green) because the finding did not result in a loss of system safety function of a single train for greater than its Technical Specifications outage time, and did not screen as potentially risk significant due to external initiating events. This finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program component, because Entergy did not take corrective actions in a timely manner. Specifically, Pilgrim did not replace the A EDG snubber valves in a prompt manner after repeated fuel leaks from cracked snubber valves over the previous two years. |
| 05000293/FIN-2010002-03 | Licensee-Identified Violation | 2010Q1 | Technical Specification (TS) 5.4.1 requires written procedures shall be established, implemented, and maintained covering procedures specified in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Contrary to this, on March 25, 2010, a Demister Door on the B train of the Standby Gas Treatment (SBGT), required to be closed following the surveillance activity in procedure 8.M.3-18, was found to be left open by an Entergy security officer conducting normal rounds. SBGT was declared inoperable and then was restored to service. This event is documented in Entergy\'s corrective action program as CR-PNP-2010-1079. The finding is of very low safety significance because the finding only represents a degradation of the radiological barrier function provided for the SBGT system |
| 05000293/FIN-2010002-04 | Licensee-Identified Violation | 2010Q1 | Technical Specification (TS) 3.7.C.1 requires secondary containment to be operable in the Run, Startup and Hot Shutdown Modes, during movement of recently irradiated fuel assemblies in the Secondary Containment and during operations with a potential for draining the reactor vessel. Contrary to the above, on December 22, 2009, the A torus trough, which is required to be maintained at a water level above Reactor Building Close Cooling Water drain pipe openings, was found dry. Secondary Containment was declared inoperable, the torus trough water level was restored and TS 3.7.C.1 was exited. This event is documented in Entergy\'s Corrective Action Program as CR-PNP-2009-5295 and CR-PNP-2009-5309. The finding was determined to be of very low safety Significance (Green) because the finding only represented an impact to the radiological barrier function provided by secondary containment and the standby gas treatment system. |
| 05000293/FIN-2010003-01 | Submerged Medium Voltage Cables | 2010Q2 | The inspectors identified a Green finding (FIN) for improper maintenance of underground non-safety related medium voltage electric cables. The inspectors identified that Entergy allowed non-safety related medium voltage cables to remain submerged in water for extended periods of time. Entergy entered this issue into their Corrective Action Program (CAP), and specified corrective actions to identify all underground medium voltage cables included under the Cable Reliability Program, and to identify which manholes should have dewatering capability. The inspectors determined that the finding was more than minor because it was associated with the Design Control 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, continued submergence of the non-safety related power cables (from the start-up transformer to electrical buses A2 and A4) could lead to cable failure and cause an event that would affect plant stability. The inspectors performed a Phase 1 Significance Determination Process screening of the finding in accordance with NRC Inspection Manual Chapter 0609, Attachment 4, \"Phase 1 - Initial Screening and Characterization of Findings,\" and determined that the finding was of very low safety significance because the condition did not contribute to both the likelihood of a reactor trip and the unavailability of mitigating systems equipment. The inspectors determined that this finding had a cross-cutting aspect in the Problem Identification and Resolution cross-cutting area, Corrective Action Program component, because Entergy personnel did not thoroughly evaluate the problem when submerged cabling was initially identified (P.1(c)). |
| 05000293/FIN-2010003-02 | Licensee-Identified Violation | 2010Q2 | Technical Specification (TS) 3.5.B.3, Reactor Building Closed Cooling Water (RBCCW) System, requires that two RBCCW subsystems shall be operable whenever irradiated fuel is in the reactor vessel, reactor coolant temperature is >212 F, and prior to startup from a cold shutdown. With one RBCCW subsystem inoperable, the required action is to restore the SUbsystem to operable within 72 hours or be in Cold Shutdown within an additional 24 hours. Contrary to the above, the \"A\" train of RBCCW was inoperable for an indeterminate amount of time that likely exceeded the 72 hours of TS allowed outage time. Upon discovery of the broken bolt on the seismic support, the \"A\" train of RBCCW was declared inoperable. An immediate corrective action was completed to install a new bolt on the seismic support and TS 3.5.B.3 was exited. The event is documented in Entergy\'s Corrective Action Program as CR-PNP-2010-0130. The finding was evaluated using IMC 0609, Significance Determination Process, and was determined to be of very low safety significance (Green) because the finding would not have resulted in the total loss of a safety function during a seismic event. |
| 05000293/FIN-2010004-01 | Failure to Manage a Yellow Risk Condition for an Unplanned Half Scram | 2010Q3 | The inspectors identified a Green non-cited violation (NCV) of 10 CFR 50.65 paragraph (a)(4) for Entergy\'s failure to manage a Yellow risk condition for an unplanned half-scram. Specifically, Entergy performed an incorrect risk assessment and thereby did not recognize an increase in risk to a Yellow condition had occurred, and as a result Entergy did not specify any risk management actions. Entergy entered this issue into their corrective action program, specified corrective actions to upgrade this risk to Yellow, and implemented appropriate risk management actions This finding was determined to be more than minor because Entergy did not consider the increase in Initiating Event likelihood where the outcome of the overall elevated plant risk put the plant into a higher risk management category, and thereby required additional risk management actions. In addition, the finding affected the Human Performance attribute of the Initiating Events cornerstone\'s objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions. The inspectors performed an evaluation in accordance with IMC 0609, Significance Determination Process, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, because the finding related to Entergy\'s assessment and management of risk. The finding was determined to be of very low safety significance (Green) because the Incremental Core Damage Probability Deficit for the medium trip risk for the duration of the activity was less than 1.0 E-6 per year (approximately 1.0 E-9 per year). The inspectors determined that this finding had a cross-cutting aspect in the Human Performance cross-cutting area, Decision Making component, because when faced with an unexpected plant condition, Entergy did not correctly implement its systematic process to make a risk-significant decision. |
| 05000293/FIN-2010005-01 | Failure to Manage a Yellow Risk Condition During HPCI Testing from the Alternate Shutdown Panel | 2010Q4 | The inspectors identified a Green non-cited violation (NCV) of 10 CFR 50.65 paragraph (a)(4) for Entergy\'s failure to correctly assess and manage a Yellow risk condition for planned testing of the High Pressure Coolant Injection (HPCI) system from the Alternate Shutdown Panel (ASP). Specifically, Entergy considered HPCI available by crediting multiple manual actions to restore the automatic function. However, these actions were not few or simple and would not have restored the HPCI automatic function in a timeframe consistent with guidance discussed in NUMARC 93-01, Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants. In addition, HPCl\'s automatic function would not have been restored in a timeframe consistent with Pilgrim\'s Updated Final Safety Analysis Report (UFSAR), Section 6.4.1, which specifies 90 seconds for HPCI to reach its required design flow rate. Corrective actions included issuing a standing order to alert Operators of the specific requirements to maintain a system available during maintenance and testing. Corrective actions planned include revising Entergy\'s Risk Assessment Procedure to verify systems credited as available have clear and simple direction to restore automatic functional status during maintenance and testing. This finding was determined to be more than minor because Entergy\'s elevated plant risk would put the plant into a higher risk category and require additional risk management actions, namely protecting the Reactor Core Isolation Cooling system. In addition, the finding affected the Human Performance attribute of the Mitigating System\'s cornerstone objective to ensure the availability of systems to respond to initiating events and prevent undesirable consequences. The inspectors performed an evaluation in accordance with IMC 0609, Significance Determination Process, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, because the finding related to Entergy\'s assessment and management of risk. The finding was determined to be of very low safety significance (Green) because the Incremental Core Damage Probability Deficit for the unavailability of HPCI for the duration of the activity was less than 1.0E-6 per year (approximately 2.6E-9 per year). The inspectors determined that this finding had a cross-cutting aspect in the Human Performance cross-cutting area, Work Control component, because Entergy did not correctly plan and coordinate work activities by incorporating appropriate risk insights (H.3(a)). (Section 1R13) |
| 05000293/FIN-2010005-02 | Failure to Perform Required Quality Control Inspections | 2010Q4 | The inspectors identified an NCV of 10 CFR 50, Appendix B, Criterion X, Inspection, for the failure to ensure that Quality Control verification inspections were consistently included and correctly specified in quality-affecting procedures and work instructions for construction-like work activities as required by the Quality Assurance Program. Entergy initiated prompt fleet-wide corrective actions to ensure proper work order evaluation and proper inclusion of Quality Control verification inspections. This issue was entered into the corrective action program as condition reports (CR) CR-HQN 2009-01184 and CR-HQN-2010-0013. The failure to ensure that adequate Quality Control verification inspections were included in quality-affecting procedures and work instructions as required by the Quality Assurance Program was a performance deficiency. This issue was more than minor because, if left uncorrected, it could lead to a more significant safety concern; in that, the failure to check quality attributes could involve an actual impact to plant equipment. This issue affected the Design Control attribute of the Mitigating Systems cornerstone because missed or improper quality control inspections during plant modifications could impact the availability, reliability, and capability of systems needed to respond to initiating events. This performance deficiency was determined to be of very low safety significance (Green), since it was confirmed to involve a qualification deficiency that did not result in a loss of operability or functionality. The inspectors determined that this issue had a cross-cutting aspect in the Human Performance cross-cutting area, Decision-Making component, because the licensee did not have an effective systematic process for obtaining interdisciplinary reviews of proposed work instructions to determine whether Quality Control verification inspections were appropriate (H.1(a)). |
| 05000293/FIN-2010005-03 | Failure to Implement the Experience and Qualification Requirements of the Quality Assurance Program | 2010Q4 | The inspectors identified an NCV of 10 CFR 50, Appendix B, Criterion II, Quality Assurance Program, for the failure to implement the experience and qualification requirements of the Quality Assurance Program. As a result, the licensee failed to ensure that an individual assigned to the position of Quality Assurance Manager met the qualification and experience requirements of ANSIIANS 3.1-1978 as required by the Quality Assurance Program. Specifically, the individual assigned to be the responsible person for the licensee\'s overall implementation of the Quality Assurance Program did not have at least one year of nuclear plant experience in the overall implementation of the Quality Assurance Program within the quality assurance organization prior to assuming those responsibilities. This issue was entered into the corrective action program as CRHQN- 2010-00386. The failure to ensure that an individual assigned to the position of Quality Assurance Manager met the qualification and experience requirements of ANSI/ANS 3.1-1978 as required by the Quality Assurance Program was a performance deficiency. This issue was more than minor because, if left uncorrected, it could create a more significant safety concern. The failure to have a fully qualified individual prOViding overall oversight to the Quality Assurance Program had the potential to affect all cornerstones, but the inspectors determined that this finding will be tracked under the Mitigating Systems cornerstone as the area most likely to be impacted. The issue was not suitable for quantitative assessment using existing NRC Significance Determination Process (SOP) guidance, so it was determined to be of very low safety significance (Green) using NRC Inspection Manual Chapter (IMC) 0609, Appendix M, Significance Determination Process Using Qualitative Criteria. The inspectors determined that there was no cross-cutting aspect associated with this finding because this issue was not indicative of current performance as it occurred more than three years ago. |
| 05000293/FIN-2010005-04 | Licensee-Identified Violation | 2010Q4 | Procedure, EN-QV-111, \'Training and Certification of InspectionNerification and Examination Personnel, Section 4.0 (4)(i), requires that the Entergy corporate ANSI Level III 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 was 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 the licensee\'s CAP as CR-HQN-2009-00111. |
| 05000293/FIN-2010008-01 | Failure to Maintain a Reliable Method for Transporting the SFP External Make-up Strategy Portable Fire Pump and Support Equipment. | 2010Q3 | This finding, affecting the Barrier Integrity Cornerstone, is related to mitigative measures developed to cope with losses of large areas of the plant; in response to Section B.5.b. of the February 25, 2002, Interim Compensatory Measures (ICM) Order (EA-02-026) and related NRC guidance. This finding has been designated as \\\"Official Use Only - Security-Related Information;\\\" therefore, the details of this finding are being withheld from public disclosure. This finding has a cross-cutting aspect in the area of Human Performance (Resources). (H.2(d)). See inspection report for more details. |
| 05000293/FIN-2010402-01 | Security | 2010Q2 | |
| 05000293/FIN-2010402-02 | Security | 2010Q2 | |
| 05000293/FIN-2011002-01 | Application of TS 3.3.8.1 When Control Rod Position Indication is Lost | 2011Q1 | An unresolved item (URl) was identified because additional information regarding the operability of control rods after control rod position indication was lost at PNPS is required to determine whether a performance deficiency exists, Control rod position indication was restored shortly after it was lost. The inspectors will review additional information when it is submitted by Entergy to determine if TS 3.3.8.1, Control Rod Operability, should have been entered when control rod position indication was lost. On January 20,2011, at 5:19 p.m., PNPS lost control rod position indication for all control rods. Instrumentation and control (l&C) technicians began troubleshooting the Rod Position Indication System (RPIS) and identified that the power supply feeding RPIS was inoperable. Operators determined that TS surveillance 4.3.8.1.5, Control Rod Operability, had been completed successfully just prior to losing RPIS and therefore they concluded that they had 24 hours from the surveillance completion before they would consider the surveillance not met. The inspector\\\'s review of Pilgrim\\\'s TS Bases identified the following statement: The OPERABILITY of an individual control rod is based on a combination of factors, primarily the scram insertion times, the associated control rod scram accumulator status, the control rod coupling integrity, and the ability to determine control rod position. When control rods are determined to be inoperable, TS 3.3.8.1, Control Rod Operability, requires the control rod to be fully inserted into the core within 3 hours. In addition, the associated Control Rod Drive for each control rod is required to be disarmed within 4 hours. l&C technicians repaired the power supply. RPIS was restored at 9:53 p.m., and control room personnel observed that there had been no change in control rod position. Condition Report (CR) CR-PNP-2O11-0272 was written to address the power supply failure and CR-PNP-2011-0511 was subsequently written to address Entergy\\\'s interpretation and administration of TS 3.3.8.1. URI 05000293/2011002-01, Application of TS 3.3.B.1 when Control Rod Position Indication is Lost. |
| 05000293/FIN-2011002-02 | Inadequate Corrective Actions for RCIC Torus Suction Valve | 2011Q1 | A self-revealing Green NCV of 10 CFR 50, Appendix B, Criterion XVl, Corrective Action, was identified for Entergy\'s failure to correct a condition adverse to quality. Entergy did not correct a Reactor Core lsolation Cooling (RCIC) torus suction valve which had failed to close during testing on October 4,2010. On January 5,2011, the same valve again failed to close during testing. Pilgrim\'s corrective actions included cleaning and replacing circuit breaker contacts and revising maintenance procedures to perform periodic resistance checks on motor control center circuit breaker cubicle secondary disconnects. Entergy has entered this issue into the corrective action program (CR-PNP-201 0-3486 and CR-PNP-2011-0046). The inspectors determined that the finding was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone, and adversely affected the cornerstone\'s objective to ensure the reliability and availability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the RCIC torus suction valve failure to close affected the reliability of the RCIC system, and the RCIC system was made unavailable during system troubleshooting and repairs in January 2011. The inspectors determined the significance of the finding using IMC 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The finding was determined to be of very low safety significance (Green) because the finding did not involve a design or qualification deficiency resulting in a loss of operability or functionality, did not result in a loss of system safety function of a single train for greater than its Technical Specification outage time, and did not screen as potentially risk significant due to external initiating events. The capability of RCIC to perform its function was not lost since the torus suction valve would have been able to be cycled open in the event RCIC needed to be aligned to the torus. This finding had a cross-cutting aspect in the Problem ldentification and Resolution cross-cutting area, Corrective Action Program component, because Entergy did not thoroughly evaluate the problem with the RCIC torus suction valve such that the resolution in October 2010 addressed the causes and corrected the problem. |
| 05000293/FIN-2011002-03 | Need For Clarification on Condensate Storage Tank Suction Piping ASME Classification | 2011Q1 | An unresolved item (URI) was identified because additional information is needed to determine if a performance deficiency exists regarding a discrepancy between various piping and instrumentation drawings (P&IDs) for the condensate and demineralized water storage/transfer systems, and the associated in-service inspection (ISI) drawings for the High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) piping and their related ASME Code safety classification. During this review the inspector noted a lack of clear description in assignment of piping safety classification shown on plant drawings for the aforementioned systems. The inspector observed that symbol convention used on the drawings was not reconciled with the requirements presented in the piping specification (M300, Rev 107, Section 3.0, ltem 1.0, Classification of Piping Components), and the P&lD legend. Entergy initiated CR PNP-2011-0127, which identified the need for a definition of the safety classification of the HPCI and RCIC suction piping from the condensate storage tanks to the first isolation valve within the auxiliary building. Appropriate code safety classification is necessary to ensure accurate in-service inspection requirements. Entergy\\\'s actions with regard to the resolution of the HPCI and RCIC piping safety classification (and boundary class changes) as shown on the referenced drawings remain to be reviewed and assessed to ascertain conformance with the applicable ASME Code (Section Xl) and NRC regulatory requirements. While the discrepancy may be related to a drawing error, it is possible that the subject piping is ASME Class 2 (or other), and should have been subject to ASME Code Class 2 (or other) design and testing requirements. URI 05000293/2011002-03, Need For Clarification on Condensate Storage Tank Suction Piping ASME Classification. |
| 05000293/FIN-2011003-01 | Transient Combustible Loading in SLC Room in Excess of the Fire Hazards Analysis Limit | 2011Q3 | The inspectors identified a Green NCV of License Condition 3.F of the Pilgri facitity Operating License (DPR-35) for the failure to evaluate transient combustible fir loading in the Standby Liquid Control (SLC) room. Specifically, Entergy did not evaluat the acceptability of transient combustibles that had been moved into the SLC room whic were in excess of the allowed combustible loading discussed in the Fire Hazards Analysis Entergy immediately walked down the area, established compensatory measures, an completed a transient combustibles evaluation. Entergy has since removed the transien combustibles from the area The inspectors determined that the failure to evaluate the transient combustibles was mor than minor based on a similar example described in Inspection Manual Chapter 0612 \\\"Power Reactor Inspection Reports,\\\" Appendix E, \\\"Examples of Minor lssues,\\\" Section 4k Specifically, the fire loading exceeded the Fire Hazard Analysis assumption and was no evaluated for acceptability. The finding is also associated with the Protection Agains External Events attribute of the Mitigating Systems cornerstone and could have adversel affected the cornerstones objective to ensure the availability of systems that respond t events to prevent undesirable consequences (i.e,, core damage). Specifically, a fire in th SLC room could affect the availability of the SLC system to respond to an event. IMC 0609 \\\"significance Determination Process,\\\" Appendix F, \\\"Fire Protection Significanc Det,ermination Process,\\\" was used to evaluate the significance of the finding. The safet significance of the finding was determined to be very low because the degradation facto was low; that is, the transient combustible evaluation process subsequently identified nearl the same level of fire protection effectiveness and reliability for the SLC room as it woul have if the degradation had not been present This finding had a cross-cutting aspect in the Human Performance cross-cutting area Work Control component; in that, Entergy did not coordinate work activities to ensure th interdepartmental coordination necessary to assure plant and human performance Specifically, the refueling organization did not notify fire protection engineering to ensure a evaluation of the\\\'transient combustible loading was completed for the SLC room (H.3(b)) (Section 1R05) |
| 05000293/FIN-2011003-02 | Submerged Medium Voltage Cables | 2011Q3 | The inspectors identified a Green finding (FlN) for the improper maintenance o underground non-safety related medium voltage electric cables. The inspectors identifie that Entergy allowed non-safety related medium voltage cables to remain submerged i water for extended periods of time. Entergy entered this issue into their corrective actio program, specified corrective actions to increase the dewatering frequency of the affecte manhole, and then installed an automatic dewatering pump The inspectors determined that the finding was more than minor because it was associate with the Design Control attribute of the Initiating Events cornerstone and affected th cornerstone objective of limiting the likelihood of those events that upset plant stability an challenge critical safety functions during shutdown as well as power operations Specifically, continued submergence of the non-safety related power cables (from the startup transformer to electrical buses A2 and 44) could lead to cable failure and cause a event that would affect plant stability. The inspectors performed a Phase 1 Significanc Determination Process screening of the finding in accordance with NRC Inspection Manua Chapter 0609, Attachment 4, \\\"Phase 1 - Initial Screening and Characterization of Findings,\\\ and determined that the finding was of very low safety significance because the conditio did not contribute to both the likelihood of a reactor trip and the unavailability of mitigatin systems equipment The inspectors determined this finding had a cross-cutting aspect in the Proble ldentification and Resolution cross-cutting area, Corrective Action Program component because Entergy personnel did not implement corrective actions in a timely manner t ensure that underground cables were not submerged, commensurate with the safet significance and complexity of the degraded condition (P.1(d)). (Section 1R06) |
| 05000293/FIN-2011003-03 | lnadequate Risk Assessment for Planned Maintenance and Testing on RCIC, SLC and ATS Systems | 2011Q3 | The inspectors identified a Green NCV of 10 CFR 50.65 paragraph (aX4) fo Entergy\'s failure to conduct an adequate risk assessment for planned Analog Trip Syste (ATS |
| 05000293/FIN-2011003-04 | Failure to Enter Technical Specifications for CHREAFS | 2011Q3 | The inspectors identified a Green NCV of Technical Specification (TS) 3\'7.8.2\'f StanOOV Gas Treatment System and Control Room High Efficiency Air Filtration Syste (CRHEAFS),\" for Entergy\'s failure to enter and perform the actions prescribed in TS afte the Control Room Envelope (CRE) was breached during work on a vital area door into th CRE. Entergy has since repaired the vital area door and restored the CRE. This finding was more than minor because it was associated with the Human Performanc attribute of the Barrier Integrity cornerstone (maintain the radiological barrier function of th control room) and adversely affected the cornerstone objective to provide reasonabl assurance that physical design barriers protect the public from radionuclide release caused by accidents or events. Specifically, the inoperable CRE could affect the operator\' ability to occupy the control room under adverse radiological, chemical, or smoke condition while responding to an event. IMC 0609, \"Significance Determination Process,\" Attachmen 0609.04, \"Phase 1- Initial Screening and Characterization of Findings,\" was used t evaluate the impact of the finding on loss of operability or functionality of the CRE an CHREAFS, and it was determined that further evaluation was required since the finding ha the potential to impact the control room envelope due to the effects of smoke and toxic gas As a result of this screening, a Phase 3 evaluation was conducted by a Senior Reacto Analyst (SRA). The SRA conducted a qualitative evaluation and determined the risk impac on control room habitability, due to this finding, from smoke and toxic gas to be lo (Green). Specifically, the Pilgrim Station Individual Plant Examination for External Event (IPEEE), sections 5.3.3 and 5.3.4, identified that the overall risk from on-site and off-sit chemical release was low The inspectors determined that this issue had a cross-cutting aspect in the Work Contro component of the Human Performance cross-cutting area. Specifically, Entergy did no plan and coordinate work activities affecting the CRE such that interdepartmenta coordination assured plant and human performance. In this case, Operations was no made aware that Maintenance would be working on the control room vital door (H.3(b)) (Section 1R15) |
| 05000293/FIN-2011003-05 | Failure to Enter Technical Specifications after Loss of Control Rod Indication | 2011Q3 | The inspectors identified a Green NCV of Technical Specification (TS) 3.3.B. Control Rod Operability,\\\" for Entergy\\\'s failure to enter and perform the actions prescribe in Technical Specifications after losing control rod position indication. Entergy has sinc restored control rod position indication by repairing a failed power supply, Condition repor CR-PNP-2011-0272 was written to address the power supply failure and condition repor CR-PNP-201 1-051 1 was subsequently written to address Entergy\\\'s administration of TSs The inspectors determined that the issue was more than minor because the finding wa associated with the Equipment Performance attribute of the Mitigating Systems cornerston and adversely affected the cornerstone\\\'s objective to ensure the reliability of systems tha respond to events to prevent undesirable consequences (i.e., core damage). Specifically the locations of the control rods were indeterminate which could substantially impac operator\\\'s abilities to implement Emergency Operating Procedures, IMC 0609 \\\"Signi1cance Determination Process,\\\" Attachment 0609.04, \\\"Phase 1-lnitial Screening an Characterization of Findings,\\\" was used to evaluate the significance of the finding Attachment 0609.04, Table 4a, was used to evaluate the impact of the finding on loss o operability or functionality. The inspectors determined that the function of the control rod to add negative reactivity to the core during an event was not affected (SCRAM time an control rod worth were not affected). In addition, alternate means were available t operators to determine control rod position and once the power supply was restored, th control rods were determined to have remained in their original positions. Also, since th finding is not potentially risk significant due to a seismic, flooding or severe weathe initiating event, the finding was determined to be of very low safety significance (Green) The inspectors determined that this issue had a cross-cutting aspect in the Decision Makin component of the Human Performance cross-cutting area. Specifically, Entergy did not us conservative assumptions in decision making and adopt a requirement to demonstrate tha the proposed action is safe in order to proceed rather than a requirement to demonstrat that it is unsafe in order to disprove the action. ln this case, Entergy did not take th conservative approach to enter Technical Specifications when faced with a degrade condition affecting control rod operability (H.1(b)). (Section 1R15) |
| 05000293/FIN-2011004-01 | Failure to Verify the Adequacy of the Design for the \'C\' Salt Service Water Pump | 2011Q3 | The inspectors identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, because Entergy\'s design control measures did not ensure two-over-one seismic protection of the \'C\' Salt Service Water (SSW) Pump. Specifically, Entergy did not ensure that a Class I to Class II interface would not result in a failure of a Class I component (\'C\' SSW Pump). Corrective actions included installing a temporary modification (Le., water shield), to protect the pump motor from potential spray effects of a Class II piping failure and performing an extent of condition review. The inspectors performed a review of Inspection Manual Chapter (IMC) 0612, Appendix E, Examples of Minor Issues, and did not find a similar more than minor example. The finding was determined to be more than minor because it was associated with the Protection Against External Events (Le., seismic) attribute of the Mitigating Systems Cornerstone, and adversely affected the cornerstone\'s objective to ensure the reliability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the \'C\' SSW pump motor was vulnerable to water spray from a failed Class II pipe during a seismic event which could have rendered the pump inoperable. The inspectors used IMC 0609.04, Phase 1 - Initial Screening and Characterization of Findings, and determined that further evaluation was required since the finding was potentially risk significant due to a seismic initiating event. As a result of this screening, a Phase 3 evaluation was conducted by a regional Senior Reactor Analyst (SRA). The condition was assessed as Green, with a change in core damage frequency (CDF) calculated to be 1.29E-8. Since the finding was assessed to have a CDF of less than 1E-7, large early release frequency was not required to be assessed. The finding does not have a cross-cutting aspect since the failure to verify the adequacy of design with respect to ensuring two-over-one seismic protection for the \'C\' SSW pump is not indicative of current licensee performance. In addition, current Entergy design procedures require rigorous Class II-over-I criteria for all new modifications. |