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| Quarter | Site | Title | Description | |
|---|---|---|---|---|
| 05000219/FIN-2005006-03 | 2005Q2 | Oyster Creek | Failure to Perform an Adequate 10 CFR 50.59 Analysis (ESW Overboard) | The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.59 Changes, Tests, and Experiments, requirements for the failure to perform an adequate safety evaluation of a change to the facility. Specifically, the safety evaluation did not evaluate the potential for a new type of malfunction of an installed liner associated with the 30-inch overboard discharge line on the emergency service water (ESW) system. This finding was addressed using traditional enforcement since it potentially impacts or impedes the regulatory process in that a required 10 CFR 50.59 evaluation was not adequate. This is contrary to the regulatory process that allows licensees to make changes without a license amendment provided that licensees comply with 10 CFR 50.59 process. The finding is more than minor because there was a reasonable likelihood that the change could have required Commission review and approval prior to implementation. However, the finding has been evaluated as very low safety significance (Green) because the liner was subsequently determined to have not have introduced a new malfunction that would impact on the ESW system. |
| 05000220/FIN-2008004-01 | 2008Q3 | Nine Mile Point | Incorrect Risk Assessment for RCIC Unavailability | An NRC-identified non-cited violation (NCV) of 10 CFR 50.65(a)(4) was identified for inaccurate risk assessments completed for August 5 and 6, 2008. Specifically, the unavailable reactor core isolation cooling (RCIC) system was not properly incorporated into the risk assessment. The cause was determined to be that an error had been made while entering a change to the risk monitor computer software, which resulted in RCIC incorrectly being assigned a zero risk importance. As corrective actions, the modeling of RCIC was corrected and a verification of all mapping codes used in the risk monitor was performed. The finding was greater than minor because the risk assessment for RCIC system maintenance was inadequate due to inaccurate information that was provided to the risk assessment tool. As a result, the overall elevated plant risk, when correctly assessed, put the plant into a higher licensee-established risk category. The finding was evaluated in accordance with IMC 0609, Appendix K, and determined to be of very low safety significance because the incremental core damage probability deficit (ICDPD) was less than 1E-6. The finding had a cross-cutting aspect in the area of human performance because NMPNS did not appropriately plan work activities by incorporating valid risk insights (H.3.a per IMC 0305). (Section 1R13) |
| 05000220/FIN-2008004-02 | 2008Q3 | Nine Mile Point | Inadequate Maintenance Practices Result in a Plant Transient | A self-revealing finding was identified on July 14, 2008, when inadequate maintenance practices, during replacement and troubleshooting of a Unit 2 radioactive waste sump pump, caused an electrical transient that resulted in the loss of numerous plant components and required a power reduction. The inadequate maintenance practices included failure to perform post-maintenance testing and continuation of troubleshooting despite having obtained results that were not consistent with the troubleshooting plan. This issue was entered into NMPNSs corrective action program for evaluation. The finding was greater than minor because it affected the human performance attribute of the Initiating Events cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The finding was evaluated in accordance with IMC 0609 and determined to be of very low safety significance because the finding did not contribute to both the likelihood of a rector trip and the likelihood that mitigation equipment or functions would not be available, and did not screen as potentially risk significant due to external events. The finding had a cross-cutting aspect in the area of human performance because NMPNS did not appropriately plan the pump troubleshooting activity by incorporating abort criteria (H.3.a per IMC 0305). (Section 4OA3) |
| 05000220/FIN-2008005-02 | 2008Q4 | Nine Mile Point | Qualification of HPCS Power Cables for SubmergencE | The inspectors inspected NMPs evaluation of the GL and corrective actions taken to resolve the potential adverse condition documented in CR 2007-1977. Specifically, on April 1, 2007, a few months after receiving GL 2007-01, NMPNS identified a condition where water was leaking into both the control and reactor buildings indicating that the HPCS power cables were submerged in water. Then on May 7, 2007, NMPNS provided the requested information in GL 2007-01 to the NRC. GL 2007-01 informed licensees of an increase in inaccessible or underground cable failure in the industry due to moisture-induced degradation. The GL discussed that periodic draining may decrease the rate of cable insulation degradation, but would not prevent cable failures. In addition, GL 2007-01 discussed that some licensees have detected cable degradation prior to failures through techniques for measuring and trending the condition of cable insulation. Although NMPNS inspected and pumped down manholes every six months and tested the insulation resistance to ground (megger) of some inaccessible/underground power cables as part of the associated HPCS motors routine maintenance, the inspectors noted that NMPNS did not evaluate the potential impact of moisture-induced failure on the HPCS power cables. In addition, the inspectors were informed that NMPNS did not consider the GL recommendations because they believed the HPCS power cables were qualified for submergence and have had no failures of underground cables at the site. The NRC reviewed NMPNSs HPCS power cable documentation to determine the HPCS power cables qualification for submerged conditions. The NRC identified that the HPCS power cables are very similar, if not identical to other power cables recently reviewed. Based on the information provided by NMPNS, it was not clear that the HPCS power cables are qualified to be submerged for the life of the plant. As a result, the submergence qualification of the HPCS power cables was a potential performance deficiency, in that 10 CFR 50, Appendix B, Criterion III requires that measures shall be established to ensure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. This issue is an unresolved item pending NMPNS providing documentation that the HPCS cables were purchased, tested and evaluated to be qualified for submergence for the life of the plant and NRC review of these documents. (URI 05000410/2008005-02, Qualification of HPCS Power Cables for Submergence |
| 05000220/FIN-2009004-01 | 2009Q3 | Nine Mile Point | Unqualified HPCS Pump Power Cables Used in Submerged Conditions | An NRC-identified non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified, in that Nine Mile Point Nuclear Station (NMPNS) failed to maintain the Unit 2 high pressure core spray (HPCS) pump power cables in an environment for which they were designed. Although NMPNS had indications that these cables were periodically submerged in water, they could not demonstrate that the cables were designed for submerged conditions. As immediate corrective action, NMPNS dewatered and inspected the HPCS cable run, and changed the frequency of dewatering to monthly. Based on the inspection results, along with the cable design specifications and most recent test results, NMPNS concluded that the HPCS pump power cables would remain operable while they conduct a design change evaluation to examine methods to reduce cable exposure to submerged conditions. The issue was entered into the corrective action program (CAP) as condition report (CR) 2009-2901.The finding was more than minor because, if left uncorrected, it had the potential to lead to a more significant safety concern. The finding affected the equipment performance attribute of the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was of very low safety significance because it was a qualification deficiency that did not result in loss of operability. The finding had a cross-cutting aspect in the area of problem identification and resolution, operating experience, because NMPNS did not use operating experience, such as Generic Letter (GL) 2007-01, Inaccessible or Underground Power Cable Failures That Disable Accident Mitigation Systems or Cause Plant Transients, to evaluate possible adverse effects of periodic submergence of the HPCS pump power cables (P.2.a per IMC 0305). |
| 05000220/FIN-2010003-01 | 2010Q2 | Nine Mile Point | Excessive Reactor Pressure Vessel Drain Down due to Inadequate Procedure | A self-revealing finding of very low safety significance associated with a non-cited violation (NCV) of Technical Specification (TS) 5.4, \"Procedures,\" was identified when Nine Mile Point Nuclear Station (NMPNS) Unit 2 operators used an inadequate procedure for reactor cavity drain down, which resulted in water being drained from the reactor pressure vessel (RPV) to a level that was significantly lower than had been planned. As a result, the steam dryer was partially uncovered, which produced elevated radiation levels on the refueling floor. As immediate corrective action, the control room operators took actions to raise water level back to the RPV flange. The event was entered into the corrective action program as condition report (CR) 2010-4408. The finding was more than minor because it was associated with the procedure quality attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was evaluated in accordance with Inspection Manual Chapter (IMC) 0609, Appendix G, \"Shutdown Operations Significance Determination Process.\" The change in core damage frequency (!\'.CDF) was determined to be of very low safety significance because of the multiple methods to inject water into the vessel and the time available to align these systems. The finding had a cross-cutting aspect in the area of human performance, resources, because NMPNS did not ensure that the RPV drain down procedure was adequate to assure nuclear safety (H.2.c per IMC 0310). |
| 05000220/FIN-2010005-01 | 2010Q4 | Nine Mile Point | Reactor SCRAM Due to Inadequate POST-MAINTENANCE Testing | A self-revealing finding of very low safety significance associated with a non-cited violation (NCV) of 10 CFR Part 50, Appendix 8, Criterion V, Instructions, Procedures, and Drawings, was identified when previously unidentified inadequate electrical connections for two solenoid operated valves (SOVs) in the control air system for the Nine Mile Point Nuclear Station (NMPNS) Unit 1 outboard main steam isolation valves (MSIVs) led to an inadvertent closure of the outboard MSIVs and resultant reactor scram. The SOV electrical connections had not been identified as defective after installation due to inadequate post-maintenance testing. As immediate corrective action, the plant was taken to cold shutdown and an investigation into the cause of the event was commenced. The issue was entered into the corrective action program (CAP) as condition report (CR) 2010-11008. The finding was more than minor because it was associated with the procedure quality attribute of the Initiating Events cornerstone and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Additionally, the finding was similar to example 4.b in Appendix E of Inspection Manual Chapter (IMC) 0612, in that it resulted in a reactor scram. The finding was determined to be of very low safety significance in accordance with IMC 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations, based on a Phase 3 analysis. The Region I senior reactor analyst (SRA) evaluated the safety significance of the finding using the Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) and Nine Mile Point Unit One Standardized Plant Analysis Risk (SPAR) models. The finding did not have a cross-cutting aspect because the performance deficiency did not occur within the past three years and therefore was not reflective of present performance. |
| 05000220/FIN-2011004-01 | 2011Q3 | Nine Mile Point | Inadequate Actions to Prevent Vibration Induced Failure on a Socket Weld for a Vent Line on the lA\' FWP Minimum Flow Line | A Green self revealing finding was identified for inadequate implementation of corrective actions regarding vibration induced failures of socket welds. This finding resulted in an August 11, 2011, Nine Mile Unit 2 scram due to a failed socket weld on the vent line for the \'A\' feedwater pump (FWP) minimum flow line. NMPNS did not properly consider the impact of high vibration levels on a vent line attached to the \'A\' FWP mini-flow recirculation line. NMPNS corrective actions included upgrading the socket weld to the requirements outlined in industry operating experience (OE). The inspectors determined that the finding was of very low safety significance (Green) through performance of a Phase 1 SOP in accordance with IMC 0609.04, Table 4a, Characterization Worksheet for Initiating Events, Mitigating Systems (MS) and Barrier Integrity Cornerstones. Specifically, 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. This finding has a cross-cutting aspect in the area of problem identification and resolution in that NMPNS did not implement and institutionalize OE through changes to station processes, procedures, equipment and training programs. Specifically in 1998 and again in 2010, NMPNS did not institutionalize external and internal OE to reduce the probability of a socket weld failure. |
| 05000220/FIN-2012004-01 | 2012Q3 | Nine Mile Point | Inadequate Implementation of Operational Decision Making Issues Monitoring Plan for EPR Results in Reactor Scram | A self-revealing Green finding (FIN) was identified for NMPNS failure to adequately implement the monitoring activities specified in the operation decision making issues (ODMI) plan for the Unit 1 electronic pressure regulator (EPR) in accordance with procedure CNGOP- 1.01-1001, Operational Decision Making . As a result, when the EPR system began to degrade on June 21, 2012, this condition was not identified by station personnel and corrective action (CA) was not implemented. The EPR subsequently malfunctioned while in service, causing a July 17, 2012, reactor scram. NMPNS removed the EPR from service and entered the issue into its corrective action program as CR-2012-006792. This finding is more than minor because it adversely affected 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. The inspectors evaluated the finding using Attachment 0609.04, Initial Characterization of Findings, in Inspection Manual Chapter (IMC) 0609, Significance Determination Process. The finding was determined to be of low safety significance (Green) because while it caused a reactor scram, it did not result in the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The finding has a cross-cutting aspect in the area of human performance, work practices, because NMPNS did not ensure proper supervisory and management oversight of the ODMI implementation plan. |
| 05000220/FIN-2012004-02 | 2012Q3 | Nine Mile Point | \\\"Inadequate Evaluation and Implementation of Design Modifications to the Turbine Gland Seal Supply System | A self-revealing Green finding (FIN) was identified for NMPNS failure to properly evaluate and implement plant design changes on the Unit 2 turbine gland seal steam supply system. Specifically, incorrect implementation of ECP-11-000977, Turbine Gland Seal and Exhaust System Instrument Changes, in May 2012 caused a reactor scram on July 13, 2012, following a return to full power operations from refueling outage N2RF13. NMPNS immediate corrective actions (CAs) included implementing ECP-12-000629 to revise the initiation setpoints for the emergency seal steam (ESS) system to accommodate higher gland seal operating pressures and properly gagging 2TME-RV135. NMPNS entered this issue into its corrective action program as CR 2012-006615. This finding is more than minor because it is similar to examples 5.a, 5.b and 5.c of IMC 0612 Appendix E, Examples of Minor Issues. In each example, plant modifications were installed and the system was returned to service without identifying and correcting a problem with the design change. This finding also adversely affects the design control attribute of the Initiating Events Cornerstone and affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors evaluated the finding using Attachment 0609.04, Initial Characterization of Findings, in Inspection Manual Chapter (IMC) 0609, Significance Determination Process. The finding was determined to be of low safety significance (Green) because while it did cause a reactor scram, it did not result in the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The finding has a cross-cutting aspect in the area of human performance in that NMPNS did not ensure that personnel and procedures were available and adequate to assure nuclear safety. Specifically, the procedures that were necessary to implement ECP-11-000977, by gagging relief valve 2TME-RV135, were not adequate to ensure proper installation of the gagging device. |
| 05000220/FIN-2012004-03 | 2012Q3 | Nine Mile Point | Inadequate Installation Instructions for Control Rod Blade Storage Rack | A self-revealing Green finding (FIN) was identified for NMPNS failure to provide adequate instructions for the installation of a control rod blade storage rack in the Unit 1 spent fuel pool. Specifically, certain critical steps were missing from the installation instructions and as a result, the rack was not properly installed, causing it to shift. The rack could have dropped, potentially resulting in damage to the spent fuel bundles stored beneath the rack. NMPNS immediate CAs were to halt further control rod blade moves and install temporary slings to hold up the rack. The rack was then re-leveled and the jacking pad was welded to the spent fuel pool curb. NMPNS entered this issue into its corrective action program as CR 2012-006547. This finding is more than minor because it would have the potential to lead to a more significant safety concern; e.g. spent fuel bundle damage and a radiological release. The inspectors evaluated the finding using Attachment 0609.04 of Inspection Manual Chapter (IMC) 0609, Significance Determination Process, Exhibit 3, Barrier Integrity Screening Questions, pertaining to spent fuel pools and determined this finding to be of very low safety significance (Green), because the finding did not adversely affect decay heat removal capabilities or pool water inventory, and did not result from fuel handling errors, dropped fuel assembly, dropped storage cask, or crane operations over the spent fuel pool that caused mechanical damage to fuel clad and a detectible release of radionuclides. The finding has a cross-cutting aspect in the area of work practices because NMPNS did not ensure supervisory and management oversight of work activities, including contractors, such that nuclear safety is supported. Specifically, NMPNS supervision did not ensure that critical assumptions contained in the control rod storage rack design analysis concerning the configuration of the Unit 1 spent fuel pool curb were translated into the installation instructions, and differences between Units 1 and 2 curbs noted during the installation were captured or evaluated by engineering, work control, or the CA process. |
| 05000220/FIN-2012004-04 | 2012Q3 | Nine Mile Point | Failure to Maintain Radiation Exposure ALARA During Refueling Activities | A self-revealing Green finding (FIN) was identified due to NMPNS having unplanned, unintended occupational collective dose resulting from deficiencies in As Low As Reasonably Achievable (ALARA) planning and work control while performing refueling activities at Unit 2. Specifically, inadequate work planning and control of refueling activities resulted in unplanned, unintended collective exposure that was greater than 50 percent above the intended collective exposure, and greater than five person-rem due to conditions that were reasonably within NMPNSs ability to foresee and correct. These factors resulted in the collective exposure for refueling activities increasing from the original estimate of 31 person-rem to an actual dose of 56 person-rem. NMPNS entered this issue into its corrective action program as CR 2012-005939. This finding is more than minor because it was associated with the program and process attribute of the Occupational Radiation Safety cornerstone, and affected the cornerstone objective to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material during routine reactor operation. This performance deficiency is similar to example 6.i of IMC 0612, Appendix E Examples of Minor Issues in that the actual collective dose exceeded 5 person-rem and exceeded the planned, intended dose by more than 50 percent. The inspectors evaluated the finding using Attachment 0609.04, Initial Characterization of Findings, of Inspection Manual Chapter (IMC) 0609, Significance Determination Process. The finding was determined to be of very low safety significance (Green) because NMPNS\\\'s current three year rolling average collective dose (143 person-rem/reactor year for 2009 to 2011) is less than the criterion of 240 person-rem per boiling water reactor unit. The finding has a cross-cutting aspect in the area of human performance, work control, in that the job site conditions which impacted human performance were not adequately incorporated into the outage plan. Specifically, the ALARA planning and work controls for refueling activities did not avert a significant unplanned and unintended collective occupational dose. |
| 05000220/FIN-2012005-03 | 2012Q4 | Nine Mile Point | Assessment of Containment Leakage Due to Containment Isolation Valve Failure | On December 3, 2012, at 11:31 a.m., Unit 1 established primary containment integrity and commenced a reactor startup from an unplanned outage. The following day at 2:40 a.m., NMPNS commenced injecting nitrogen into the primary containment as part of a planned activity to reduce primary containment oxygen concentration to less than four percent as required by TS 3.3.1, Oxygen Concentration . This activity was completed 10:55 a.m., on December 4. Once an appropriate nitrogen concentration has been achieved in containment, additional makeup is generally not required. However, from December 6 - 8, on three occasions, operators added additional nitrogen to the containment to maintain pressure within procedural limits. This issue was documented in CR 2012-011157, Adverse Trend in Unit 1 Nitrogen Usage. This issue was initially classified as a priority 2 work item and NMPNS commenced initial troubleshooting activities, which included examining systems/components that were possible sources of nitrogen leakage. However, a definitive source for the leakage was not identified. On December 12, following a fourth addition of nitrogen that occurred on December 11, NMPNS increased the importance of the issue to Priority 1, formed an incident response team and staffed the outage coordination center. As part of the investigation process, NMPNS cycled several containment isolation valves in the nitrogen purge and vent system, and attempted to quantify the amount of seat leakage through the valves by opening test fittings located between isolation valves. In parallel with the troubleshooting efforts, NMPNS and vendor personnel began to develop analytical tools that could be used to quantify the amount of containment leakage. On December 13, at 6:47 p.m., after observing a decrease in containment pressure following a fifth nitrogen addition, and receiving preliminary data that a containment isolation valve local leak rate test (LLRT) between reactor containment inert gas purge and fill drywell cooling system isolation valves IV-201-31 and IV-201-32 may fail, NMPNS commenced a plant shutdown because primary integrity as required in TS 3.3.3 could not be assured. The plant reached cold shutdown on December 13 at 11:33 p.m. Subsequent NMPNS testing of containment isolation valves revealed that three valves in the reactor containment inert gas purge and fill drywell cooling system IV-201-10, IV- 201-31 and IV-201-32 had unacceptable seat leak rates. These conditions were documented in several CRs including 2012-011210 and 2012-011288. When the valves were disassembled and examined, NMPNS identified that iron oxide buildup on the valve resilient seats had prevented the valves from closing tightly and adversely impacted seat leakage performance. The reactor containment inert gas purge and fill drywell cooling system is a carbon steel system and the internal piping surface adjacent to the valves had visible signs of iron oxide degradation (rust). NMPNS corrective action included removing the loose surface rust, installing new seats on the valves, and successfully performing as-left LLRTs on the subject valves. Additional corrective actions are outlined in CR 2012-011157. This issue will be tracked as a URI pending NMPNS quantification of the drywell leakage that existed from December 3 - 13, 2012 and NRC review of the NMPNS evaluation to determine whether the issue is more than minor and whether a violation exists. NMPNS intends to complete the evaluation by January 31, 2013. |
| 05000220/FIN-2013003-01 | 2013Q2 | Nine Mile Point | Failure to Follow Containment Isolation System Surveillance Procedure Resulting in Isolation of the Reactor Coolant Isolation Cooling System | A self-revealing NCV of TS 5.4.1, Procedures, was identified at Unit 2 when a CENG instrumentation and control (I&C) technician did not properly implement procedure N2-ISP-LDS-Q010, Reactor Building General Area Temperature Instrument Channel Functional Test, Revision 00102. As a result, a residual heat removal (RHR)/reactor core isolation cooling (RCIC) isolation bypass switch was inadvertently left in the NORMAL position during surveillance testing resulting in an unplanned RCIC isolation. CENG entered this issue into their CAP as CR-2013-002461. Other corrective actions included performing a human performance stand down that reinforced use of human performance tools and the need to identify and mark critical steps during pre-job briefs, retraining the I&C technicians involved in the event on proper use of human performance error prevention techniques, and improving bypass switch verification steps for procedure N2-ISP-LDS-Q010 and other similar lead detection system surveillances procedures. This finding is more than minor because it is associated with the human performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the inadvertent isolation rendered the RCIC system inoperable and unable to perform its function for approximately 6 hours. Additionally, this finding is similar to example 4.b of IMC 0612, Appendix E, Examples of Minor issues, and is more than minor due to the procedural error leading to a plant transient, i.e. an unplanned RCIC isolation. This finding was evaluated 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, issued June 19, 2012. Unit 2 is a boiling-water reactor (BWR)-5, and as a result, RCIC is treated as having a separate high-pressure injection safety function. A detailed analysis was conducted using SAPHIRE version 8.0.8.0 and Unit 2 SPAR model 8.17. Using an exposure period of 6 hours and conservatively assuming no recovery of the failed equipment, this finding had a change in core damage frequency of low E-8. The dominant accident sequence was a grid-related loss of offsite power with a failure of Division III power and the failure to recover offsite power and the emergency diesel generators (EDGs) in 30 minutes. Since the change in core damage frequency was less than 1E-7, contributions from large early release and external event did not need to be considered. Therefore, this finding was of very low safety significance (Green). This finding has a cross-cutting aspect in the area of Human Performance, Work Practices, because the I&C technicians did not effectively employ self-checking and place-keeping when implementing the test procedure which directly contributed to the resulting procedural error. |
| 05000220/FIN-2013003-02 | 2013Q2 | Nine Mile Point | Inadequate Procedural Implementation for Battery Cell Replacement | The inspectors identified an NCV at Unit 2 of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because CENG did not assure that the replacement of cells in battery 2C were prescribed and performed by appropriate procedures which resulted in degraded accuracy of test results and potential degradation of safety-related battery cells. In response to this issue, CENG generated CR-2013-005235 and initiated actions to evaluate replacing the new cells. This finding is more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. 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, issued June 19, 2012, the inspectors determined this 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 a 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. This finding has a cross-cutting aspect in the area of Human Performance, Decision-Making component, because CENG did not use conservative assumptions in decision making. Specifically, CENG did not monitor the cells in storage, question the adequacy of the discharged cells, charge the cells prior to installation, or fully evaluate the implications of the test and recharge results. |
| 05000220/FIN-2013003-03 | 2013Q2 | Nine Mile Point | Inadequate Design Control for Battery Sizing Calculation | The inspectors identified an NCV at Unit 2 of 10 CFR Part 50, Appendix B, Criterion III, Design Control, because CENG did not verify the adequacy of the design with respect to battery 2C. Specifically, by failing to size the battery to the most limiting time period, the sizing calculation significantly overstated the available design margin. CENGs corrective actions included generating condition report CR-2013-005117 and evaluating the condition for operability. This finding is more than minor because it was associated with the design control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. 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, issued June 19, 2012, the inspectors determined this 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 a 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 inspectors did not assign a cross-cutting aspect because the finding was not indicative of current performance. |
| 05000220/FIN-2013003-04 | 2013Q2 | Nine Mile Point | Improper Bus Restoration Results in a Loss of Shutdown Cooling | A self-revealing apparent violation of Technical Specification (TS) 6.4.1, Procedures, was identified at Unit 1 because CENG failed to properly recover from a loss of a vital direct current (DC) bus in accordance with station off-normal procedures resulting in an unplanned loss of all shutdown cooling (SDC) when time to boil was less than 2 hours. Specifically, during the restoration from the loss of battery bus 12, operators failed to identify a SDC trip signal before attempting restoration of the DC bus, which ultimately lead to a SDC pump trip (i.e. loss of decay heat removal from the reactor). Corrective actions included conducting a prompt human performance event review, entering the issue into their corrective action program (CAP), and conducting a root cause analysis. Planned corrective actions include a review of all emergency, off-normal, and normal system operating procedures. The inspectors determined that CENGs failure to properly restore battery bus 12 in accordance with N1-SOP-47A.1, Loss of DC, Revision 00101, and N1-OP-47A, 125 VDC Power System, Revision 02500, was a performance deficiency that was reasonably within CENGs ability to foresee and correct and should have been prevented. The performance deficiency was determined to be more than minor because the inspectors determined it affected the configuration control aspect of the Initiating Events cornerstone and adversely affected the associated cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The significance of the finding is designated as To Be Determined (TBD) until a Phase 3 analysis can be completed by the NRCs Senior Reactor Analysts. The inspectors determined this finding has a cross-cutting aspect in the area of Human Performance, Resources, because CENG did not ensure that personnel, equipment, procedures, and other resources were available and adequate to assure nuclear safety - complete, accurate and up-to-date design documentation, procedures, work packages, and correct labeling of components. Specifically, CENG procedures N1-SOP-47A.1 and N1-OP-47A did not contain adequate guidance to ensure recovery from a loss of a DC bus would not result in an unexpected plant transient. |
| 05000220/FIN-2013003-05 | 2013Q2 | Nine Mile Point | Containment Leakage Exceeds Technical Specification 3.3.3 Limits | A self-revealing NCV of TS 3.3.3, Leakage Rate, was identified for CENGs failure from December 3 to December 13, 2012, to maintain containment leakage less than 1.5 percent by weight of the containment air per day and less than 0.6 percent by weight of the containment air per day for all penetrations and all primary containment isolation valves subject to 10 CFR Part 50, Appendix J, Types B and C tests, when pressurized to 35 pound per square inch gauge when reactor coolant system (RCS) temperature is above 215F and primary containment integrity is required. CENG entered this issue into their CAP as CR-2012-011247. Corrective actions included cleaning iron oxide from the primary containment vent and purge valve and replacing the resilient seals. This finding is more than minor because it is associated with the structure, system, component (SSC), and barrier performance 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. Specifically, containment leakage exceeded the leakage limits outlined in the Unit 1 TS 3.3.3 from December 3 to December 13, 2012. This finding was evaluated in accordance with IMC 0609.04, Initial Characterization of Findings, and Table 6.2, Phase 2 Risk Significance-Type B Findings at Full Power, of IMC 0609, Appendix H, Containment Integrity Significance Determination Process, issued May 6, 2004. The inspectors determined this finding was of very low safety significance (Green) because the leakage was less than 100 percent of containment volume per day for the duration of the leak. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, CAP, because CENG failed to take appropriate corrective action to address safety issues and adverse trends in a timely manner commensurate with their safety significance. Specifically, following identification of the adverse trend regarding the frequency of nitrogen addition to the drywell, CENG did not assess in a timely manner the significance of the leakage and the impact on primary plant containment. |
| 05000247/FIN-2006005-03 | 2006Q4 | Indian Point | Assess Reliability / Unavailability of the Gas Turbine System and Impacts on Functionality | The inspectors identified the URI during a routine Maintenance Rule inspection on the gas turbine system. Gas turbines 1 and 3 (GT-1 and GT-3) are credited in Entergy's analysis to cope with station blackout and Appendix R fire scenarios to ensure safe shutdown of the reactor. The system is classified as risksignificant in accordance with Entergys Maintenance Rule program. This system has been in a category (a)(1) monitoring status since the inception of the Maintenance Rule in 1996. An (a)(1) action plan was established to improve overall system performance. However, Entergy may not have provided justifiable (a)(1) goals for maintenance preventable functional failures (MPFFs) and Entergy may not have appropriately classified repeat maintenance preventable functional failures (RMPFFs). Specific to reliability, the goal was set as less than or equal to five MPFFs and no RMPFFs in a 24 month rolling cycle. The number of allowable MPFFs was calculated under the assumption that there would be, on average, 82 start demands during the 24 month cycle. The inspectors reviewed the operating history over the last three years and determined that the number of start demands averaged 38 during the 24 month cycle. The inspectors need more information to evaluate Entergy's goals for MPFFs to determine their adequacy. Additional information is required to evaluate Entergy's implementation of the Maintenance Rule as it pertains to the gas turbine system. Actual unavailability and reliability information is needed to evaluate the gas turbine system performance and to assess whether performance of the system is bounded by the Station Blackout / Appendix R commitments, and assumptions in the design basis. This issue will be treated as a URI pending additional licensee input and inspector evaluation of gas turbine system performance. |
| 05000247/FIN-2007007-03 | 2007Q1 | Indian Point | Use of Motor Control Center Methodology for Periodic Verification of the Design Basis Capability of Safety- Related MOV | The team identified an unresolved item (URI) concerning the adequacy of the motor control center (MCC) testing methodology used for periodic verification of the design bases capability of safety-related MOVs. Entergy implemented MCC testing in 2004 as a method of implementing periodic verification in addition to the previously NRCreviewed method of taking stem thrust and torque measurements at the valve. The MCC method uses motor current, voltage, and winding resistance measured at the MCC to calculate motor torque of the valves motor operator. The calculated motor torque is then compared to motor torque target and limit values based on 1) packing loads, 2) thrust required to close the valve, 3) stall motor torque, and 4) valve or actuator structural limits. Entergy Report IP-RPT-04-00890, Technical Basis for Using MCC Technology for Periodic Verification Testing at IP 2 and IP 3, states that this methodology would be used initially on MOVs with generally low safety significance and high operating margin, but also states that the report applies to all safety related MOVs at IP 2 and IP 3. Since 2004, Entergy has used the MCC methodology for periodic verification on nine safety-related MOVs: three high risk, three medium risk, and three low risk MOVs, where risk significance is defined as the combined effects of MOV risk of failure and safety significance. Based on the available information, the team was unable to verify that the MCC method had been appropriately validated. Specifically, there did not appear to be a justified correlation between the MCC methodology calculated motor torque and actual stem thrust and torque. It was also unclear whether the MCC methodology had adequate allowances to compensate for its uncertainties in establishing MOV design basis capability (such as uncertainties related to stem friction coefficient, load sensitive behavior, and actuator efficiency) since stem thrust and stem torque are not directly measured. MCC testing was performed in 2004 as a periodic verification test on MOV 747, the No. 21 RHR heat exchanger discharge valve, a high risk valve. The team identified that this test was invalid because this was not performed in accordance with IP-RPT-04-00890. Specifically, MOV 747 was tested using the Motor Torque Method of MCC testing which, according to IP-RPT-04-00890, is only valid for motors whose torque is between 2 and 60 foot-pounds. The motor on MOV 747 is an 80 foot-pound. motor and use of the Correlated Thrust/Torque Method was required. As a result, Entergy exceeded the sixyear periodic verification test interval for MOV 747 because the last at the valve valid performance verification test was performed in May 2000. Entergy has provided justification for the reasonable continued operability of the valve until its scheduled testing in 2008 based on successful in-service tests, stem lubrication and actuator preventive maintenance and inspection performed in 2006. The team reviewed Entergys basis for the operability of MOV 747 and determined that there was reasonable assurance of continued operability of the MOV. Entergy committed to follow the Joint Owners Group program for periodic verification of MOVs in their response to NRC Generic Letter 96-05. This periodic verification program established valve margin by measuring stem thrust and torque at the valve. Entergys response did not indicate that the MCC method would be used for MOV periodic verification. The MCC test method for periodic verification of MOVs is a departure from the NRCreviewed method, which is based on direct measurement of stem thrust and torque. The acceptability of the use of the MCC methodology for periodic verification of MOVs will be an unresolved item pending further NRC review. Included with this review will be a determination of whether the MOV performance testing conducted on MOV 747 constitutes a violation of NRC requirements. |
| 05000271/FIN-2009004-01 | 2009Q3 | Vermont Yankee | Failure to initiate corrective action condition reports for all deficient items identified during cooling tower inspections | The inspectors identified a Green NCVof 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and DraWings, in that Entergy did not initiate corrective action condition reports (CRs)\'for all deficient items identified during Cooling Tower (CT) inspections. Entergy entered this issue into their corrective action program (CAP) and performed an operability assessment which determined that the safety related function of the CTs was always available. The inspectors determined that the finding was more than minor because if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, deficiencies might not be tracked to resolution, management attention or other independent reviews would not be appropriately applied, and the need for operability determinations may be missed. 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. This finding had a cross-cutting aspect in the Work Practices component of the Human Performance cross-cutting area because Entergy did not follow procedures and initiate CRs to identify cooling tower deficiencies as required by operating procedure (OP) 52114. IH.4(b) |
| 05000272/FIN-2012002-01 | 2012Q1 | Salem | Licensee-Identified Violation | TS 3.3.1.1, Reactor Trip System Instrumentation, requires one operable intermediate range instrument in Modes 1, 2, and 3. When that requirement is not met, TS 3.0.3 requires that action be initiated within one hour to place the unit in hot standby within the next six hours. Contrary to the above requirements, on November 24, 2011, after technicians input incorrect trip setpoints that rendered both TS 3.3.1.1 required intermediate range NISs inoperable, PSEG did not initiate action within one hour to place the unit in hot standby within the next six hours. Upon discovery of the inoperable instruments on December 11, 2011, PSEG took immediate corrective action to input the correct intermediate range setpoints and restore operability and entered this issue into the CAP as notification 20538425. The inspectors determined that the finding was of very low safety significance (Green) in accordance with NRC IMC 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, Mitigating Systems Cornerstone, because the finding did not represent a loss of safety system function, an actual loss of safety function of a single train for greater than its TS allowed outage time, an actual loss of safety function of one or more non-technical specification trains of equipment designated a risk significant per 10 CFR 50.65 for greater than 24 hours, and did not screen as risk significant due to seismic, flooding or severe weather initiating event. |
| 05000277/FIN-2007002-04 | 2007Q1 | Peach Bottom | Incorrect size breaker resulted in a fire in the 4T4 480 volt load center | At approximately 9:16 a.m. on February 27, 2007, a fire was suspected to have started based on the receipt of numerous secondary plant alarms in the main control room (MCR) and the report of smoke near the 4T4\' 480 Volt load center. The inspectors responded to the MCR following a site announcement for the fire brigade to respond to a suspected fire in the Unit 3 turbine building. The inspectors monitored the operators response to the event and the status of plant equipment. The observations were primarily focused on the nuclear safety aspects of the plants and operators responses. The inspectors also monitored the response of PBAPSs emergency response organization to the declaration of an UE. Subsequent to the fire, the inspectors discussed the fire with operations, engineering and PBAPS management personnel to gain an understanding of the event and to assess their followup actions. The inspectors reviewed operator logs and operators actions taken in accordance with licensee procedures. Based on the operators narrative logs, the fire brigade was dispatched to the Unit 3 turbine building at approximately 9:20 a.m. Fire personnel investigated and notified the MCR that an actual fire existed at 9:38 a.m. An Unusual Event for a fire not extinguished within 15 minutes (emergency action level (EAL) HU6) was declared at 9:41 a.m. All state and local government notifications were completed by 9:59 a.m. and the NRC Headquarters Operations Officer was notified of the event at 10:36 a.m. The fire was considered to be extinguished at approximately 10:32 a.m. At 11:37 a.m., the Unusual Event was terminated. Prior to the report of the potential fire, Unit 3 was operating at full power. As a result of fire and the associated response actions, numerous non-safety-related loads powered by the 4T4\' 480 Volt load center were de-energized. Equipment that was de-energized included: the B isophase bus cooler fan, the B drywell chiller, the B recirculation pump speed controller, the leading edge flow meters and the B reactor feed pump. Plant operators took the required TS actions and responded to the equipment losses by performing controlled reactor power reductions and stabilized the plant at approximately 50 percent of rated power. The inspectors verified that the required reports were made during the event and that no further reports are planned. The inspectors also verified that this issue (IR 569889) was placed into the CAP. Preliminarily, PBAPS has determined that the fire resulted from an apparent mismatch between the ratings of one breaker and its cubicle in the 4T4\' 480 volt load center. A root cause investigation was ongoing at the end of the inspection period and will be reviewed by the inspectors during a future inspection period. At the close of this inspection period, the inspectors were reviewing the event and awaiting the results of the root cause evaluation to understand the potential performance deficiencies. This issue is unresolved pending review of PBAPSs causal evaluation and corrective actions by the inspectors to characterize the issue. URI 05000277/2007002-04, Incorrect Size Breaker Resulted in a Fire in the 4T4\' 480 Volt Load Center. |
| 05000277/FIN-2007002-05 | 2007Q1 | Peach Bottom | Missed procedure step resulted in unplanned overloading of the E-3 EDG | The inspectors reviewed selected applicable plant records, correction action documents and approved procedures while evaluating the performance of operations personnel in response to non-routine evolutions. The inspectors assessed personnel performance to determine what occurred and how the operators responded, and to determine if plant personnels response was in accordance with plant procedures and training. The following non-routine evolution was reviewed: During the conduct of surveillance testing of the E-3 EDG on March 15, 2007, a licensed operator missed the performance of a required step in a supporting system operating procedure. The omission of the procedure step placed the E-3 EDG in the isochronous mode while synchronized with offsite power through a 4 kilovolt (kV) vital bus. This condition resulted in unexpectedly loading the E-3 EDG beyond its 30-minute load rating. The ST and supporting procedures directed the synchronization of the E-3 EDG to a selected 4 kV bus to pick up the bus loads. The procedure subsequently directed opening the offsite power feeder breaker to the 4 kV vital bus (the missed step) before placing the EDG in the isochronous mode. PBAPS placed this issue in the CAP by initiating IR 604364. Prompt corrective actions included the selected implementation of additional peer checking of procedure performance place-keeping. The E-3 EDG was inspected for potential damage and tested before being returned to an operable condition in accordance with TS on March 17, 2007. The causal evaluation of this event was ongoing at the end of the inspection period. At the close of this inspection period, the inspectors were reviewing the event and awaiting the results of the causal evaluation to understand the potential performance deficiencies. This issue is unresolved pending review of PBAPSs causal evaluation and corrective actions by the inspectors to characterize the issue. URI 05000277/2007002-05, Missed Procedure Step Resulted in Unplanned Overloading of the E-3 EDG. |
| 05000278/FIN-2018001-01 | 2018Q1 | Peach Bottom | Untimely Corrective Actions to Address Primary Containment Isolation Valve Condition Adverse to Quality | A Green self-revealing non-cited violation(NCV)of 10 Code of Federal Regulations(CFR)50, Appendix B, Criterion XVI, Corrective Action, was identified because Exelon did not implement prompt corrective actions to address a condition adverse to quality (CAQ) on primary containment isolation valve (PCIV) SV-3-7D-3671G.Specifically, drywellair sampling valve SV-3-7D-3671G failed to perform its PCIV function on February 1, 2018, by failing to stroke closed during its surveillance test as a result of untimely corrective actions.Exelon isolated the associated piping in accordance with technical specifications(TSs) |
| 05000289/FIN-2014002-01 | 2014Q1 | Three Mile Island | Failure to Perform a 10 CFR 50.59 Evaluation for the BWST Seismic Qualifications | The inspectors identified a Severity Level IV (SL-IV), Non-Cited Violation of 10 CFR 50.59, Changes, Tests, and Experiments, and an associated finding of very low safety significance (Green) for Exelons failure to perform a 50.59 evaluation review to determine whether a license amendment was required to align the borated water storage tank (BWST) to non-seismic piping. Specifically, Exelon staffs 50.59 screening accepted the alignment of the seismically qualified BWST to a non-seismically qualified clean-up system. The inspectors determined the alignment would involve a change to the BWST that adversely affects its Updated Final Safety Analysis Report chapter 5.1.1, Classes of Structures and Systems for Seismic Design, described design function of being seismically qualified. Additionally, the inspectors determined that following the 50.59 review Exelon placed the line-up in service. The inspectors determined these two actions were performance deficiencies that were reasonably within Exelons ability to foresee and prevent. Furthermore, the 50.59 screening credited unapproved operator manual actions to ensure functionality of the BWST. Exelon documented this as issue report 1631468 and implemented interim corrective actions to isolate the BWST from the clean-up system until a permanent resolution is determined and implemented. The inspectors determined the 50.59 violation regarding the failure to perform an evaluation was more than minor because the inspectors could not reasonably determine that the alignment would not have ultimately required NRC prior approval, because the BWST alignment was not in accordance with the current licensing basis and the evaluation credited the use of unapproved operator manual actions. The inspectors also determined that the performance deficiency of accepting and aligning the adverse clean-up line-up, challenging the BWST seismic qualification, was more than minor because it adversely affected the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated the finding using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At-Power, and determined that this finding required a detailed risk evaluation. The detailed evaluation was performed which determined that the performance deficiency was a finding of very low safety significance (Green). Additionally, In accordance with Section 6.1.d.2 of the NRC Enforcement Policy, the 50.59 violation is categorized as a Severity Level IV. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Operating Experience, in that the station did not effectively evaluate and internalize relevant external operating experience (Information Notice (IN) 2012-01) regarding connections between safety-related seismic and non-seismic qualified piping and components. |
| 05000289/FIN-2015004-01 | 2015Q4 | Three Mile Island | Failure to Trend Vibration Data for Safety Related River Water Pump | The inspectors identified a finding of very low safety significance involving an NCV of 10 Code of Federal Regulations (CFR) 50, Appendix B Criterion XVI, Corrective Action Program, because Exelon did not identify and correct a condition adverse to quality on the B nuclear river water pump (NR-P-1B). Specifically, Exelon did not properly evaluate an adverse vibration trend on NR-P-1B, which resulted in exceeding its in-service test (IST) required action level and declared inoperable on October 10, 2015. Exelon entered the condition into their corrective action program (CAP) as issue report 2568763 and emergently replaced the pump, engaged the vendor for short and long term design and material changes to correct the vibration, and created process and peer check corrective actions to ensure all vibration data is reviewed timely and trends are addressed commensurate with their safety significance. The performance deficiency is more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the elevated vibrations reduced the reliability and capability of NR-P-1B to perform its safety function. The inspectors evaluated the finding using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 2, and determined this finding to be of very low safety significance (Green) because the degraded condition was not a design deficiency that affected system operability; did not represent an actual loss of function of a system; did not represent an actual loss of function of a single train or two separate trains for greater than its technical specification (TS) allowed outage time and did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety significant. The finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, because the station did not thoroughly evaluate the elevated vibration data such that the issue was addressed before NR-P-1B became inoperable (P.2). |
| 05000293/FIN-2013003-01 | 2012Q2 | Pilgrim | Failure to Follow Procedures Results in Loss of Shutdown Cooling | A self-revealing NCV of Technical Specification (TS) 5.4.1, Procedures, was identified for operators not implementing procedures to supply safety-related alternate electrical power to shutdown cooling valves during shutdown cooling operation. Specifically, because operators did not perform all applicable steps in a procedure, a loss of shutdown cooling resulted when operators were shifting power supplies for the B train shutdown cooling suction and discharge valves on May 2, 2013. Corrective actions included restoring shutdown cooling following a prompt investigation of the event. Entergy has captured this event in their corrective action program (CAP) as CR-PNP-2013-3457. The performance deficiency is more than minor because it affects the objective of the Initiating Events cornerstone to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The unavailability of shutdown cooling for five hours challenged the safety function of decay heat removal (DHR) supplied by the residual heat removal (RHR) system. A review of IMC 0612, Appendix E, Examples of Minor Issues, found no more than minor examples that applied. The inspectors evaluated the finding using IMC 0609, Attachment 4, Phase 1 Initial Screening and Characterization of Findings. The inspectors determined that the finding required further review using IMC 0609, Appendix G, Shutdown Operations Significance Determination Process, because the issue affected the safety of the reactor during a refueling outage. The inspectors determined that this finding was of very low safety significance (Green), using IMC 0609, Appendix G, Checklist 7, BWR Refueling Operation with Reactor Coolant System (RCS) Level >23. This determination did not require a further phase 2 or phase 3 analysis in that it did not increase the likelihood of a loss of RCS inventory; did not result in the loss of RCS level instrumentation; did not degrade Entergys ability to terminate a leak path or add RCS inventory; and did not degrade Entergys ability to recover DHR once it was lost. In addition, a loss of thermal margin did not occur since the change in RCS temperature resulted in less than 20 percent of the temperature margin to boil. The inspectors determined that this finding had a cross-cutting aspect in the Human Performance area, Work Practices component, because personnel did not follow procedures. |
| 05000334/FIN-2010004-01 | 2010Q3 | Beaver Valley | Inadequate Maintenance Procedure Results in Auto-Disassembly of EDG Intake Damper | A Green, self-revealing finding (FIN) was identified in that an inadequate procedure resulted in a failure to adequately retain a 1-1 Emergency Diesel Generator (EDG) room damper after louver adjustment. Specifically, the adjustment of the 1-1 EDG upper damper (1VS-D-22-2A) in April 2010 led to retention hardware not being sufficiently secure to prevent damper failure and resulted in the linkage failing to open the upper dampers. This was selfrevealing during a crew investigation for a 1-1 EDG alarm on September 5,2010. This issue was entered into the licensee\'s corrective action program under CR 10-82257. Traditional enforcement does not apply because the issue did not have an actual safety consequence or the potential for impacting NRC\'s regulatory function, and was not the result of any willful violation of NRC requirements. The finding is more than minor because it is associated with the procedure quality attribute of the Mitigating Systems cornerstone and affects the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. In accordance with IMC 0609.04 (Table 4a), Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to be of very low safety significance (Green). The cause of this finding relates to the cross-cutting aspect of Human Performance, Resources, in that FENOC did not provide complete procedures to conduct the damper adjustment and retention. (H.2.(c)) |
| 05000334/FIN-2014003-02 | 2014Q2 | Beaver Valley | Removal of Missile Barrier Renders Containment Inoperable | The inspectors identified a Green non-cited violation of TS limiting condition for operation (LCO) 3.6.1, Containment. Specifically, the inspectors determined that FENOC removed the missile barriers for the unit 1 and unit 2 containment equipment hatches while in a mode when containment was required to be operable. As a result FENOC did not have adequate tornado protection for containment and then did not take the actions directed by the LCO action statement when the LCO was not met. FENOC entered the issue into their corrective action program, CR 2014-11878, and placed the procedures to remove the missile barriers on administrative hold. The performance deficiency is more than minor because it adversely affected the configuration control attribute of the Barrier Integrity Cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. In accordance with IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 3, Barrier Integrity Screening Questions, this finding screens to Green, very low safety significance. This finding has a cross-cutting aspect in the area of conservative bias where individuals use decision making-practices that emphasize prudent choices over those that are simply allowable and that a proposed action is determined to be safe in order to proceed, rather than unsafe in order to stop. Specifically, FENOC did not adequately consider the containment operability implications of removing the missile barriers for the unit 1 and unit 2 containment equipment hatches while in a mode where containment is required to be operable. (H14) |
| 05000334/FIN-2014003-03 | 2014Q2 | Beaver Valley | Licensee-Identified Violation | Technical Specification 5.7.2, High Radiation Area, requires, in part, that locked doors be provided for each high radiation area in which the intensity of radiation exceeds 1000 millirem per hour. Contrary to the above, on April 26, 2014, for approximately 2.5 hours, the door to the Regenerative Heat Exchanger room was not locked. FENOCs immediate corrective action included placing chains and padlocks on this door and all similar style entrances to locked high radiation areas, entering this issue into their corrective action program (CR-2014-07646), and performing a root cause evaluation. The finding is of very low safety significance, Green, because it did not involve ALARA, there was no overexposure, there was no substantial potential for an overexposure, and the ability to assess dose was not compromised. |
| 05000334/FIN-2014005-01 | 2014Q4 | Beaver Valley | Failure to Adequately Implement Risk Management Actions | The inspectors identified an NCV of 10 CFR 50.65(a)(4), Requirements for monitoring the effectiveness of maintenance at nuclear power plants, for FENOCs failure to implement adequate risk management actions (RMAs) associated with maintenance on the alternate intake structure A bay. Specifically, FENOC did not establish a contingency plan for the maintenance activity as required by FENOCs risk management procedure. FENOC entered the issue into their corrective action program as CR 2015-00267. The performance deficiency is more than minor because it is 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. Specifically, FENOCs failure to implement a contingency plan resulted in an increase in the duration of an elevated risk condition and unavailability of equipment relied upon to mitigate the consequences of a loss of the main intake structure. The finding was determined to be of very low safety significance (Green) because the incremental core damage probability (ICDP) for the event was less than 1.0 E-6. The inspectors determined that this finding had a cross-cutting aspect in the Human Performance, Work Management, because the FENOC work process failed to adequately manage the risk commensurate to the work (H.5). |
| 05000334/FIN-2014005-02 | 2014Q4 | Beaver Valley | Failure to Properly Ship Category 2 Radioactive Material | The inspectors identified an NCV of 10 CFR 71.5, Transportation of licensed material, and 49 CFR 172, Subpart I, Safety and Security Plans. Specifically, FENOC personnel shipped a category 2 radioactive material of concern (RAM-QC) on public highways to a waste processor without adhering to a transportation security plan. FENOCs corrective actions included revising procedure NOP-OP-5201, Shipment of Radioactive Material Waste, to reflect the appropriate Department of Transportation requirements for shipment of Category 2 radioactive material. FENOC entered the issue into their corrective action program as CR 2014-17260. The issue is more than minor because it is associated with the Program and Process attribute of the Public Radiation Safety cornerstone and adversely affected its objective to ensure the safe transport of radioactive material on public highways in accordance with regulations. The finding was determined to be of very low safety significance (Green) because FENOC had an issue involving transportation of radioactive material, but it did not involve: (1) a radiation limit that was exceeded; (2) a breach of package during transport; (3) a certificate of compliance issue; (4) a low level burial ground nonconformance; or (5) a failure to make notifications or provide emergency information. The inspectors determined that the finding did not have a cross-cutting aspect because the issue was not reflective of current plant performance. Specifically, FENOC implemented changes to the radioactive waste shipment procedure that addressed applicable requirements and implemented a formal process for reviewing pending regulatory changes for impacts to FENOC operations and support activities. |
| 05000334/FIN-2015002-01 | 2015Q2 | Beaver Valley | Failure to Utilize Respiratory Protection as Specified by the Radiation Work Permi | The inspectors identified a self-revealing NCV of Technical Specification 5.4.1, Procedures, for FENOCs failure to utilize respiratory protection, as required by the applicable radiation work permit (RWP), for entry into the 722-foot elevation of the solid radioactive waste building on March 12, 2014. This resulted in the unplanned internal exposure of one worker. Immediate corrective actions included reestablishing RWP controls of the area and entering this issue into their corrective action program as condition report 2015-06636. The inspectors determined that the performance deficiency is more than minor because it affected the Program and Process attribute of the Occupational Radiation Safety cornerstone objective to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. The inspectors evaluated the finding using NRC Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, and determined the finding to be of very low safety significance (Green) because it was not related to as low as (is) reasonably achievable (ALARA), did not result in an overexposure or a substantial potential for overexposure, and did not compromise the licensee's ability to assess dose. The finding has a cross-cutting aspect of Human Performance, Conservative Bias, in that individuals did not use decision making-practices that emphasized prudent choices over those that are simply allowable. Specifically, a radiation protection technician did not use conservative decision making practices and make prudent choices when entering an area with unknown radiological conditions. Examples of non-conservative decision making included: failure to wear respiratory protection when entering into unknown radiological conditions, the failure to complete and evaluate an air sample prior to entry, and not taking into account the adverse radiological conditions of the adjoining area above (735 foot elevation). (H.14) |
| 05000334/FIN-2015003-01 | 2015Q3 | Beaver Valley | Failure to Correct a Low Oil Level in the Condensate Pump Motor | A self-revealing finding was identified for FENOCs failure to correct a low oil level in the lower motor bearing of the Unit 1 A condensate pump in accordance with NOP-LP- 2001, Corrective Action Program. Specifically, FENOC incorrectly cancelled the work order to add oil to the A condensate pump motor and installed a placard on the oil level sight glass with incorrect minimum and maximum oil levels. This led to the motor bearing failure, which caused the pump to trip on overcurrent, and required the operators to insert a manual reactor trip. FENOC entered the issue into their correct action program, condition report (CR) 2015-05256. The performance deficiency was more-than-minor because it was associated with the human performance attribute of the Initiating Events cornerstone, and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, NOP-LP-2001, section 4.2.3, states that condition report/correct action owners should ensure that actions are developed to resolve the primary cause identified in the condition report. Instead of correcting the low oil level in the motor, FENOC cancelled the work order to add oil. This subsequently caused the operators to trip the plant when the condensate pump motor bearing overheated and the motor tripped on overcurrent. The inspectors determined that this finding was of very low safety significance (Green) because it did not cause a reactor trip and the loss of mitigation equipment. This finding has a crosscutting aspect in the area of Human Performance, Consistent Process, because FENOC did not seek input from the appropriate work group (engineering) prior to cancelling the work order to add oil to the condensate pump motor (H.13) |
| 05000334/FIN-2015004-01 | 2015Q4 | Beaver Valley | Inadequate Maintenance Rule Monitoring of the Auxiliary Feedwater System | The inspectors identified an NCV of Title 10 of the Code of Federal Regulations (CFR) 50.65, Requirements for monitoring the effectiveness of maintenance at nuclear power plants, for FENOCs failure to monitor the performance of the Unit 1 auxiliary feedwater (AFW) system against licensee-established goals. Specifically, FENOC did not identify and properly account for a maintenance preventable functional failure (MPFF) of the turbine driven auxiliary feedwater (TDAFW) pump, which demonstrated that performance of the Unit 1 AFW system was not being effectively controlled through appropriate preventive maintenance. FENOCs immediate corrective actions included entering this issue into their corrective action program, re-evaluating and classifying the TDAFW pump failure as a MPFF, performing a 10 CFR 50.65 (a)(1) evaluation of the Unit 1 AFW system, and placing the system in (a)(1) status. The performance deficiency was determined to be more-than-minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Additionally, example 7.d from IMC 0612 Appendix E details that a performance deficiency is more than minor if equipment performance problems were such that effective control of performance through appropriate preventive maintenance under (a)(2) could not be demonstrated. This finding was determined to be of very low safety significance (Green) since it was not a deficiency affecting the design or qualification of a mitigating structure, system, or component (SSC), it did not represent the loss of a system and/or function, it did not represent an actual loss of function of at least a single train or two separate safety systems out-of-service for greater than its technical specifications allowed outage time, and it did not represent an actual loss of a non-technical specification equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. This finding has a cross-cutting aspect in Human Performance, Avoid Complacency, because FENOC failed to consider the extent of condition and their causes following the failure of the Unit 1 TDAFW pump on January 6, 2014 (H.12). |
| 05000334/FIN-2016001-01 | 2016Q1 | Beaver Valley | Failure to Properly Evaluate Control Room Envelope Test Results | The inspectors identified an NCV of Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion XI, Test Control, for FENOCs failure to properly evaluate the test results of the Control Room Envelope (CRE) unfiltered air in-leakage test performed in December 2015. Specifically, the test results exceeded the acceptance criteria specified in the test procedure and required further engineering evaluation to determine if the control room emergency ventilation system (CREVS) could meet its specified safety function. The inspectors identified that the engineering evaluation of the test results did not account for all of the in-leakage and resulted in a reasonable doubt of operability of CREVS. FENOCs immediate corrective action was to re-evaluate the December 2015 calculation and verify that CREVS remained operable with the increased in-leakage. FENOC entered the issue into their corrective action program, condition report (CR) 2016-03836. The performance deficiency is more-than-minor because it is associated with the human performance attribute of the Barrier Integrity cornerstone, and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect from radionuclide releases caused by accidents or events. Specifically, FENOCs evaluation did not account for in-leakage from the non-tested portions of the control room radiological barrier, and therefore, did not provide reasonable assurance that the control room dose would not exceed five rem during an uncontrolled release of radioactivity. Additionally, this issue is similar to example 3j and 3k of IMC 0612 Appendix E, Examples of Minor Issues, in that FENOCs December 2015 engineering evaluation failed to adequately account for CRE in-leakage and resulted in a reasonable doubt of the operability of CREVS. The inspectors determined that this finding was of very low safety significance (Green) because it only represented a degradation of the radiological barrier function provided for the control room. This finding has a cross-cutting aspect in the area of Human Performance, Conservative Bias, because FENOC did not take a conservative approach to decision making, particularly when the in-leakage information was incomplete (H.14). |
| 05000334/FIN-2016002-01 | 2016Q2 | Beaver Valley | Procedure Change Results in Failure to Maintain the Design Basis for the Service Water System | The inspectors identified an NCV of Title 10 of the Code of Federal Regulations (CFR) 50, Appendix B, Criterion III, Design Control, for FENOCs failure to assure that the regulatory requirements and design basis for the Unit 2 service water system were correctly translated into procedures. Specifically, FENOC implemented a procedure revision in 2002 that inappropriately removed the step to declare the Unit 2 service water system inoperable while the non-seismic standby service water system is aligned to it. FENOCs immediate corrective actions included issuing instructions that prohibit planned testing of or swapping to the standby service water system and revising procedure 2OST-30.1A. FENOC entered the issue into their CAP as condition report (CR) 2016-01710. The performance deficiency is more-than-minor because it is associated with the Design Control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, FENOCs revision to 2OST-30.1A in 2002 resulted in reduced reliability of the service water system while connected to the standby service water system for over ten hours on February 1, 2016, and nine hours on April 3, 2014. This finding was of very low safety significance (Green) because it did not represent a loss of system and/or function, an actual loss of function of a single train for greater than its technical specification allowed outage time, an actual loss of function of one non-technical specification trains designated as high safety significant, and did not involve a loss or degradation of equipment designed to mitigate a seismic, flooding, or severe weather initiating event. This finding does not have a cross-cutting aspect because it is not representative of current performance. The inadequate review of revision 17 to 2OST-30.1A was an isolated instance that occurred over 14 years ago. Furthermore, the most recent NRC inspection of Changes, Tests, or Experiments and Permanent Plant Modifications, performed in 2013, and the Component Design Basis Inspection, performed in 2014 did not document any findings related to procedure changes. (Section 1R15) |
| 05000334/FIN-2016002-03 | 2016Q2 | Beaver Valley | Failure to Appropriately Utilize Multiple and Diverse Indications Results in Plant Transient | A self-revealing finding of NOP-OP-1002, Conduct of Operations, was identified for FENOCs failure to adequately implement operator fundamentals. Specifically, operators did not appropriately utilize multiple and diverse indications when making the decision to isolate electro-hydraulic control (EHC) to a Unit 1 main turbine governor valve. This resulted in an unanticipated reactor power reduction of 2.7 percent. FENOCs immediate corrective actions included re-opening the governor valve, verifying proper system response, and entering this issue into their corrective action program (CAP) as CR 2015-08263. The performance deficiency is more-than-minor because if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern. Additionally, example 4.b from IMC 0612 Appendix E details that a performance deficiency is more-than minor if it causes a reactor trip or other transient. This finding was determined to be of very low safety significance (Green) since it did not cause both a reactor trip and the loss of mitigation equipment relied upon to transition the plant to a stable shutdown condition. This finding has a cross-cutting aspect in Human Performance, Challenge the Unknown, because individuals did not consult the system expert when confronted with an unexpected condition (H.11). |
| 05000334/FIN-2016003-01 | 2016Q3 | Beaver Valley | Failure to Identify Conditions Adverse to Quality Leads to Inoperable Emergency Bus Degraded Voltage Relays | The inspectors identified an NCV of Title 10 Code of Federal Regulations (CFR) 50, Appendix B, Criterion XVI, Corrective Action, for FENOCs failure to assure that a condition adverse to quality was promptly identified and corrected. Specifically, FENOC failed to promptly identify and correct a negative trend in setpoint drift and as found dropout voltage values in the AB 27N model 411T6375HF 4160 volts alternating current (VAC) and 480 VAC emergency bus degraded voltage relays. FENOCs immediate corrective actions included recalibrating or replacing the relays and entering the issue into their corrective action program (CAP) as condition report (CR) 2016-12018. The performance deficiency is more than minor because it is associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, FENOCs failure to promptly identify and address a negative trend in dropout voltage setpoint drift and as found values resulted in the reduced reliability of safety related bus degraded voltage relays (seven surveillance failures and inoperable degraded bus relays between 2011 and 2016). Inoperable emergency bus degraded voltage relays could lead to damage of safetyrelated equipment during a loss of offsite power. This finding is of very low safety significance (Green) because it does not represent a loss of system and/or function, an actual loss of function of a single train for greater than its technical specification allowed outage time, an actual loss of function of one non-technical specification trains designated as high safety significant, and did not involve a loss or degradation of equipment designed to mitigate a seismic, flooding, or severe weather initiating event. The finding has a crosscutting aspect in the area of Problem Identification and Resolution, Trending, because FENOC did not periodically analyze the results of the degraded voltage relay surveillances to provide early indication of a declining trend (P.4). |
| 05000334/FIN-2016004-01 | 2016Q4 | Beaver Valley | Failure to Follow Procedure Results in an Inoperable A River Water Train | A self-revealing NCV of Title 10 of the Code of Federal Regulations (CFR) 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified for FENOCs failure to assure that activities affecting quality were accomplished in accordance with procedures. Specifically, FENOC failed to follow NOP-OP-1001, Clearance/Tagging Program, and clearance 1W11-30-MNM-002 when removing the clearance for the A bay of the main intake structure. This resulted in disabling the automatic start capability of the standby C river water pump and made the A river water train inoperable and unavailable. FENOCs immediate corrective action was to rack the breaker for the A river water pump to the disconnect position, which cleared the annunciator and restored operability to the A train of river water. FENOC entered this issue into their corrective action program (CAP) as condition report (CR) 2016-14253. The performance deficiency is more-than-minor because it is associated with the Human Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, FENOC incorrectly racked the A river water pump breaker onto the 1AE 4160 volts alternating current (VAC) safety bus while the C river water pump was already racked onto the bus. This caused the A train of river water to be inoperable and unavailable because the automatic start capability of the C pump was disabled. The inspectors determined that this finding was of very low safety significance (Green) because it did not represent a loss of system and/or function, an actual loss of function of a single train for greater than its technical specification allowed outage time, or an actual loss of function of one non-technical specification train designated as high safety significance. This finding has a cross-cutting aspect in Human Performance, Avoid Complacency, because the operators did not plan for the possibility of mistakes and did not implement appropriate error-reduction tools (H.12). |
| 05000334/FIN-2016004-02 | 2016Q4 | Beaver Valley | Licensee-Identified Violation | The following licensee-identified violation of NRC requirements was determined to be of very low safety significance and meets the NRC Enforcement Policy criteria for being dispositioned as a NCV. Radioactive material shipment B-4655, was made from Beaver Valley on May 5, 2016, to ResinSolutions in Erwin, TN. During a self-assessment performed by the FENOC staff on November 3, 2016, it was identified that the scaling factors used to determine the hard-to-detect nuclides listed on the manifest (NRC Form 540) for shipment B-4655 were incorrect. The scaling factors used to manifest the shipment were not for the waste stream shipped. Recalculation of the isotopic values using the correct waste stream scaling factors resulted in different numeric values for multiple radionuclides in the shipment, but did not cause a change in the proper shipping name, packaging, or labeling. 10 CFR 71.5 requires, in part, that radioactive materials be transported with an accurate shipment manifest. Contrary to the above, on May 5, 2016, FENOC transported radioactive materials with a shipment manifest that incorrectly stated that the radiological activity of the package was higher than the actual activity. FENOC documented this issue in CR 2016-13071, and provided a corrected shipment manifest to the recipient of the material. In accordance with IMC 0609, Appendix D, "Public Radiation Safety Significance Determination Process," the finding was determined to be of very low safety significance (Green) because FENOC had an issue involving transportation of radioactive material, but it did not involve a radiation limit that was exceeded, a breach of package during transport, a certificate of compliance issue, a low level burial ground nonconformance, or a failure to make notifications or provide emergency information. |
| 05000334/FIN-2017001-01 | 2017Q1 | Beaver Valley | Failure to Follow the ASME OM Code for a Failed Relief Valve Set Pressure Test | Severity Level IV. The inspectors identified a Severity Level IV NCV of Title 10 of the Code of Federal Regulations (CFR) 50.55a(z), Alternatives to codes and standards requirements, for FENOCs failure to obtain prior authorization for implementing an alternative to the American Society of Mechanical Engineers Code for Operation and Maintenance of Nuclear Power Plants (ASME OM Code). Specifically, until prompted by the inspectors, FENOC did not submit to the NRC and receive an alternative to the ASME OM Code requirement to not test the residual heat removal (RHR) relief valve, RV-1RH-721, during a recent refueling outage for Unit 1 when the charging system letdown relief valve, RV-1CH-203, failed to lift within three percent of set-pressure. FENOCs immediate corrective actions included performing a prompt operability determination, submitting a relief request, and entering the issue into the corrective action program (CAP) as condition report (CR) 2017-03937. The inspectors determined that this violation impacted the ability of the NRC to perform its regulatory oversight function, and was therefore subject to traditional enforcement. Section 2.2.1.c of the Enforcement Policy states that failure to receive prior NRC approval for changes in licensed activities when required is an example of impacting the ability of the NRC to perform its regulatory oversight function. After considering the factors in Section 2.2.1.c of the Enforcement Policy, the inspectors determined that the performance deficiency was a Severity Level IV violation because the change implemented by FENOC would likely be approved by the NRC. Because this violation involves the traditional enforcement process and does not have an associated finding that is more than minor, the inspectors did not assign a cross-cutting aspect to this violation in accordance with IMC 0612, Appendix B. |
| 05000334/FIN-2017001-02 | 2017Q1 | Beaver Valley | Operability Determinations and Functionality Assessments | 10 CFR 50, Appendix B, Criterion III, Design Control, requires, in part, that measures shall be established to assure that the applicable regulatory requirements and the design basis for structures, systems, and components are correctly translated into specifications, drawing, procedures, and instructions. Contrary to the above, FENOC failed to correctly translate the design basis for protection against tornado-generated missiles into their specifications and procedures. Specifically, FENOC did not adequately protect Unit 1 and Unit 2s main steam safety and atmospheric dump valve exhausts from tornado-generated missiles. Additionally, FENOC did not adequately protect Unit 2s component cooling pumps and spent fuel from tornado-generated missiles by failing to include in their procedures actions for closing the tornado doors in the event of a tornado. The inspectors evaluated FENOCs immediate compensatory measures, which included verifying that procedures are in place and training is current for performing actions in response to a tornado. Because this violation was identified during the discretion period covered by Enforcement Guidance Memorandum 15-002, Revision 1, Enforcement Discretion for Tornado Missile Protection non-compliance (ML16355A286) and because FENOC has implemented compensatory measures, the NRC is exercising enforcement discretion, is not issuing enforcement action, and is allowing continued reactor operation. |
| 05000334/FIN-2017002-01 | 2017Q2 | Beaver Valley | Licensee-Identified Violation | The following violation of very low safety significance (Green) was identified by FENOC and is a violation of NRC requirements which meets the criteria of the NRC Enforcement Policy for being dispositioned as a NCV . TS 3.7.8, "Service Water System", requires two service water trains to be operable. There is no associated action provided for both trains inoperable. LCO 3.0.3 states, in part, that when an LCO is not met and an associated action is not provided, the unit shall be placed in a MODE or other specified condition in which the LCO is not applicable. Act ion shall be initiated within one hour to place the unit, as applicable, in M ODE 3 within 7 hours. Contrary to the above, on August 20, 2015 and August 31, 2015 , FENOC had both trains of service water inoperable for greater than 7 hours while performing the service water full flow test and did not place Unit 2 in Mode 3. FE NOC entered this issue into the CAP as CR 2017- 04023. The inspectors evaluated this finding using IMC 0609.04, Initial Characterization of Findings . Because the finding represented a loss of function of a system, a detailed risk evaluation was performed. A Region I senior reactor analyst used the BVPS Unit 2 Standardized Plant Analysis Risk Model version 8.5 to perform the evaluation. A seismic initiating event frequency was obtained from the Risk Assessment of Operational Events Handbook Volume 2, External Events. A surrogate loss -of-offsite - power event was used applying the seismic initiating event frequency for BVPS with a train of service water being failed with no recovery assumed. The finding was determined to be of very low safety significance (Green) because the limited exposure time in this configuration resulted in a change in core damage frequency in the 1E -10/yr range. The dominant core damage sequence was a seismic event with failure of the EDG . |
| 05000334/FIN-2017003-01 | 2017Q3 | Beaver Valley | Operability Determinations and Functionality Assessments | Inspection Scope The inspectors reviewed operability determinations for the following degraded or non- conforming conditions based on the risk significance of the associated components and systems: Unit 1 Anchor Darling double disk gate valves evaluation resulting from NRC Information Notice 2017- 03 on July 13, 2017 Unit 1 fire protection system functionality during a fire water header break on July 20, 2017 Impact on Unit 1 SSST 1A from nearby fire water header break on July 20, 2017 Unit 1 EDG exhaust piping not protected from tornado- generated missiles on July 25, 2017 Unit 1 degraded main steam valve room high energy line break door on July 26, 2017 Unit 2 inoperable DRPI impact on verifying operability of control rod F10 on August 25, 2017 Unit 1 EDG 1 -2 building exhaust damper missing louver on September 22, 2017 The inspectors evaluated the technical adequacy of the operability determinations to assess whether TS operability was properly justified and the subject SSC remained available such that no unrecognized increase in risk occurred. The inspectors compared the operability and design criteria in the appropriate sections of the TS s and UFSAR to FENOCs evaluations to determine whether the SSCs were operable. The inspectors confirmed, where appropriate, compliance with bounding limitations associated with the evaluations. Where compensatory measures were required to maintain operability , the inspectors determined whether the measures in place would function as intended and were properly controlled by FENOC. 11 b. Findings 10 CFR 50, Appendix B, Criterion III, Design Control, requires, in part, that measures shall be established to assure that the applicable regulatory requirements and the design basis for SSCs are correctly translated into specifications, drawing, procedures, and instructions. Contrary to the above, FENOC failed to correctly translate the design basis for protection against tornado generated missiles into their specifications and procedures. Specifically, FENOC did not adequately protect Unit 1 EDG s exhausts from tornado generated missiles. FENOC documented the condition adverse to quality in their CAP under condition report 2017 -07550 and took immediate compensatory actions. The inspectors evaluated FENOCs immediate compensatory measures, which included verifying that procedures are in place and training is current for performing actions in response to a tornado. Because this violation was identified during the discretion period covered by Enforcement Guidance Memorandum 15- 002, Revision 1, Enforcement Discretion for Tornado Missile Protection Non- compliance (ML16355A286) and because FENOC has implemented compensatory measures, the NRC is exercising enforcement discretion and is not issuing enforcement action and is allowing continued reactor operation |
| 05000334/FIN-2017004-01 | 2017Q4 | Beaver Valley | Inadequate Control of Entry into High Radiation Areas | A self-revealing, very low safety significance NCV of Technical Specification (TS) 5.7.1 for failure to control a high radiation area (HRA) was identified. On November 8, 2017, during independent spent fuel storage installation (ISFSI) dry cask loading campaign activities, the failure of multiple barriers resulted in a worker gaining access to an HRA while signed onto an incorrect radiation work permit (RWP) and a subsequent dose rate alarm. Specifically, a worker signed on to an incorrect RWP during a break, and did not recognize that the surveyed work area dose rates were higher than the RWP setpoints. Additionally, radiation protection personnel controlling access to the HRA failed to ensure that the worker was on the correct RWP per plant procedure requirements for a subsequent entry into anHRA. This resulted in the worker entering an HRA under the incorrect RWP and receiving a dose rate alarm of 1,070 millirem per hour. Upon receiving a dose rate alarm, the worker backed away from the area and reported the issue to radiation protection personnel. FENOCs immediate corrective actions included putting the work in a safe condition, performing follow-up surveys, and verifying remaining personnel trip tickets to ensure all individuals were on the correct RWP. FENOC entered the issue into their corrective action program (CAP) as condition report (CR) 2017-11206.The failure to control access to an HRA is a performance deficiency that was within FENOCs ability to foresee and correct and should have been prevented. The performance deficiency is more than minor because it is associated with the Program and Process attribute (Procedures) of the Occupational Radiation Safety cornerstone and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation from radioactive material during routine reactor operation. Specifically, the failure of multiple barriers resulted in a worker gaining access to an HRA while signed on to an incorrect RWP and receiving a dose rate alarm. IMC 0612, Appendix E, Section 6, Health Physics, General Screening Criteria, states that a performance deficiency involving more than one barrier or the loss of a significant barrier would be classified as a more-than-minor performance deficiency. Using IMC 0609,Appendix C, Occupational Radiation Safety Significance Determination Process, the finding was determined to be of very low significance (Green) because: (1) it was not an as low as reasonably achievable (ALARA) finding, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. The finding was a human performance cross-cutting aspect associated with avoiding complacency because FENOC failed to ensure individuals recognize and plan for the possibility of mistakes and ensure individuals implement the appropriate error reduction tools, even when expecting a successful outcome (H.12) |
| 05000334/FIN-2018003-01 | 2018Q3 | Beaver Valley | Inadequate Verification of Full Low Head Safety Injection Suction Piping | A self-revealed Green non-cited violation (NCV) of technical specification(TS)5.4.1, Procedures, was identified when FENOC failed to adequately implement procedure 1OM-52.4.R.2.A, Station Startup Mode 6 to Mode 1 Administrative and Local Actions, to verify that the low head safety injection (LHSI) suction pipes were full of water. Specifically, the non-destructive examination (NDE) inspector incorrectly determined that the suction pipes were full, which led to inoperability of one or more trains of LHSI for in excess of four hours on May 22, 2018,when the suction lines were found to be voided. |
| 05000336/FIN-2007007-01 | 2007Q4 | Millstone | Control Room Fire Evacuation Procedure | Pending further inspector review of control room fire scenario developments, the team identified a potential vulnerability in Dominions safe shutdown method and operating procedures during a control room fire scenario that would require evacuation. Specifically, the reactor coolant makeup and reactor coolant pump seal injection functions of the charging pumps were potentially affected. The team reviewed 25212-BTP-9.5-1, MP3 Branch Technical Position 9.5-1 Compliance Report, Rev. 003, and noted that the charging pump volume control tank (VCT) outlet level control valves were subject to spurious operation during a control room fire. The team questioned engineers regarding the validity of this conclusion and if Dominion considered the impact of the VCT outlet isolating the suction path to an operating charging pump during the time necessary to evacuate the control room and to establish control at the auxiliary shutdown panel. The engineers confirmed that a VCT outlet level control valve spurious closure was possible for a control room fire and that an operating charging pump would fail in a relatively short time without a suction path. Dominion also provided the team condition reports, CR-04-08399 and CR-04-08450, that were initiated on September 16, 2004, for the same potential issue the inspectors raised: catastrophic failure of an operating charging pump due to loss of a suction path from a spurious closure of the VCT outlet level control valves. Only two charging pumps, A and B, were analyzed for safe shutdown, and only the A charging pump could be operated remotely for a fire requiring control room evacuation. The third charging pump, C, was available but was normally secured without a 4kV breaker installed. The team noted that the C charging pump was also not analyzed for safe shutdown. Dominions safe shutdown analysis for a Unit 3 control room fire only credited the A charging pump for boration and reactor coolant pump (RCP) seal injection functions. The team further identified that Kirk interlock keys were not available outside of the main control to rack in a normally disconnected 4kV breaker to the C charging pump. Regarding the potential issue, Dominion completed several immediate corrective actions including: equipment configuration changes (i.e., 4kV breaker alignments and kirk interlock key locations were implemented) to ensure the C charging pump would remain available after such a scenario, operators were notified of the condition through a night order entry, fire protection compensatory measures were initiated to minimize the potential for a fire in the areas of concern, and an extent of condition review was performed to identify additional plant areas where this potential vulnerability may exist. Dominion entered the issue into the corrective action program as CR-07-10124, CR-07- 10158, CR-07-10363, and CR-07-10614. The team determined that the potential vulnerabilities to the reactor coolant makeup and RCP seal injection functions during a control room fire that required evacuation will be treated as an unresolved item (URI), pending further inspector review of credible fire scenarios and their potential impact on the VCT outlet valve control cabling. An unresolved item is an issue requiring further information to determine if it is acceptable, if it is a finding, or if it constitutes a violation of NRC requirements. In this case, additional NRC inspection will be required to assess credible control room fire scenarios that lead to a loss of reactor coolant makeup and conditions that may require control room evacuation |
| 05000336/FIN-2008002-01 | 2008Q1 | Millstone | Failure to Evaluate a Unit 2 Charging System Non-conforming Condition against the Current Licensing Bases (Section 1R15) | Green. The inspectors identified a finding for Dominions failure to evaluate a nonconforming plant condition against the current licensing basis (CLB) as required by Dominion procedure OP-AA-102-1101, Revision 0, Development of Technical Basis to Support Operability Determinations. Specifically, Dominion, in multiple instances, failed to evaluate the impact that a potential common mode charging system failure would have on the Updated Final Safety Analysis Report Chapter 14.6.1, Inadvertent Opening of Power Operated Relief Valves (PORVs), event, the analysis of record for which credited both charging and safety injection availability. Corrective actions for this issue included the initiation of an operations standing order and crew briefings to ensure all crews understood the CLB related to Unit 2 charging and the need to implement the compensatory action for this chapter 14.6.1 event, and a subsequent operability determination (OD) revision to ensure charging was properly evaluated and documented within the OD. This finding is more than minor because, if left uncorrected, the issue would become a more significant safety concern. Specifically, degraded and non-conforming plant conditions must be evaluated against their credited functions in the CLB to ensure the adverse condition is properly evaluated for operability. This finding was determined to be of very low safety significance (Green) because it did not result in a loss of charging system operability or functionality. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program component, because Dominion did not thoroughly evaluate a Unit 2 charging system non-conforming condition against the CLB (P.1(c)). (Section 1R15 |