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| Q | Significance | CCA | Identified by | Title | Description | |
|---|---|---|---|---|---|---|
| 05000219/FIN-2018410-03 | 2018Q2 | Green | H.9 | NRC identified | Security | |
| 05000219/FIN-2018410-02 | 2018Q2 | Green | H.12 | NRC identified | Security | |
| 05000219/FIN-2018410-01 | 2018Q2 | Green | H.11 | NRC identified | Security | |
| 05000219/FIN-2018410-04 | 2018Q2 | Green | Licensee-identified | Licensee-Identified Violation | ||
| 05000219/FIN-2018001-02 | 2018Q1 | Severity level Enforcement Discretion | NRC identified | Enforcement Action (EA)-18-007: No. 2 Emergency Diesel Generator Ring Lug Failure | On October 9, 2017, during a routine surveillance load test, the No. 2 emergency diesel generator failed approximately 5 minutes into the run due to a broken ring lug on a current transformer. Laboratory analysis of the broken ring lug determined that the ring lug failed due to fatigue cracking that was initiated due to stresses caused by bending and twisting of the electrical lug. Exelon last conducted a load surveillance on the No. 2 emergency diesel generator on September 25, 2017. Corrective Actions: Corrective actions included replacement on the broken ring lug on the No. 2 emergency diesel generator, extent of condition inspections on the No. 1 and No. 2 emergency diesel generators for additional bent or twisted ring lug connectors, and revision to the electrical ring lug installation and emergency diesel generator procedures to include inspection for bent or twisted ring lugs. Corrective Action Reference(s): Issue report 4060815 Enforcement:Violation: Oyster Creek Technical Specification 3.7.C.2.b states, in part, that if one diesel generator becomes inoperable during power operation, the reactor may remain in operation for a period not to exceed 7 days. Contrary to the above, on October 9, 2017, it was recognized that one diesel generator was inoperable for greater than the technical specification allowed outage time of 7 days, and Oyster Creek continued power operation. Specifically, on October 9, 2017, No. 2 emergency diesel generator failed to run during a routine surveillance test due to a broken ring lug on a current transformer, which resulted in a total inoperability time of 6.5 months.Severity/Significance: For violations warranting enforcement discretion, Inspection Manual Chapter 0612 does not require a detailed risk evaluation, however, safety significance characterization is appropriate. A Region I Senior Reactor Analyst (SRA) performed a best estimate analysis of the safety significance using the Oyster Creek Standardized Plant Analysis Risk (SPAR) model, Version 8.50 and Systems Analysis Programs for Hands-On Integrated Reliability Evaluations (SAPHIRE). The evaluation estimated the total (internal and external events risk) increase in core damage frequency (CDF) to be in the mid to high E-6/yr range, or a low to moderate safety significance. The SRA evaluated the internal events risk contribution due to the inoperability of the No. 2 emergency diesel generator for an approximate 6.5 month exposure time. The exposure time relative to when the No. 2 emergency diesel generator was no longer capable of meeting its 24 hour mission time is uncertain due to the effect of vibration induced fatigue, and therefore the method prescribed within the RASP handbook guidance was used. 9 The analyst used the guidance in Section 2.5 of the Handbook, Revision 2.0, to estimate the exposure time of 6.5 months based on the cumulative 24 hour summation of the No. 2 emergency diesel generator surveillance test proven run time. This approach is appropriate for periodically operated components that degrade during operation (i.e. vibration induced fatigue only occurs while the emergency diesel generator is in-service/operating). Given this approach, the dominant internal events, loss of offsite power were evaluated for the estimated internal risk increase. This contribution was estimated at 2E-6/yr increase in CDF. The dominant sequences involved loss of offsite power events with a concurrent failure of the No. 1 emergency diesel generator, failure of the combustion turbines, and failure to recover offsite power or recover an emergency diesel generator prior to core damage.The SRA performed various modeling changes after a review of revised calculations for DC battery life:Analysis noted that Oyster Creek Generating Station recirculation pump seals are similar in design to those tested in reports generated for Nine Mile Point Unit 1 with the use of CAN2A seals. Therefore, the failure probability of the seals in the station blackout sequence wasadjusted from 0.1 to 5E-2 similar to Nine Mile Point Unit 1 SPAR model 8.50.The failure to load shed action (DCP-XHE-XM-LSHED) in the model was calculated using the SPAR-H method and revised to 1.2E-2 versus being assumed to always fail (TRUE).Failure probabilities for 1, 2, or 3 stuck open electromatic relief valves were revised to be consistent with the previous model version 8.22 because of the isolation condenser design at Oyster Creek Generating Station which limits cycling and significantly reduces the probability of a failed open electromatic relief valve due to isolation condensers controlling pressure.The depressurization function using electromatic relief valves, if required, was calculated through SPAR-H to be 1E-2 for sequences where total seal failure is assumed (DEPSEALFAIL) (conservatively assumed limited time available).The diesel driven firewater pumps are both available and were set to calculated fault tree failure probabilities instead of always failed in the previous model. These are 2,000 gallons per minute pumps with a large supply of water and relatively simple operator actions to inject to the reactor pressure vessel. The firewater was assumed to fail at 0.1 when a total recirculation seal failure occurs due to assumed time constraints.The offsite power and the emergency diesel generator required recovery time events were increased to 24 hours for events where DC load shedding was successful, without seal failures and isolation condenser success along with diesel driven firewater success.The SRA noted the No. 2 emergency diesel generator was recoverable. In fact, the diagnosis of the failed condition was performed in a nominal 8-10 hours from the failure. Therefore, a probability of failure to recover event for the conditional case was developed. The SRA used SPAR-H as simple guidance, which conservatively supported a reasonable assumption of a 0.10 conditional probability of failure to recover the emergency diesel generator within 24 hours. The base case utilized a calculation within SPAR of 0.33 failure to recover probability for 24 hour sequences. To estimate the external risk contribution, the SRA identified that the most significant external risk contribution was from fire events. Seismic, external flooding, and high wind events were not significant contributors for the issue. From discussions with Oyster Creek Fire probabilistic risk analysts and a review of this failure condition, the increase in CDF due to the failed No. 2 emergency diesel generator for the assumed 6.5 month exposure time was estimated at 4.5E-6/yr ((8.5E-5/yr-4.5E-5/yr) x (6.5/12 months) x 0.2).The DC safety-related battery life would be at least a nominal 14 hours and longer if DC bus stripping occurred, this allows for extended isolation condenser or electromatic relief valve function, with injection from diesel driven firewater. Given the time considerations and characteristics of the failure, an assumed recovery at a failure probability of 0.2 (slightly higher than internal due to less time) was applied for the No. 2 emergency diesel generator, which was a best estimate determined through SPAR-H insights. The dominant fire sequence was a fire affecting the A and B 4kV switchgear rooms, where combustion turbine support would be lost, with failure of the No. 1 emergency diesel generator breaker to close, and failure of locally operating the isolation condenser due to eventual loss of power. The SRA noted that FLEX credit was not quantified and would result in a lower risk estimation likely in the low E-6/yr range. Combining internal and external risk contributions, the total increase in CDF was 6.5E-6/yr, or low to moderate safety significance. The SRA determined that Exelon uses a Large Early Release Frequency (LERF) factor value of 8E-2. This value takes into consideration operator action for those relevant high pressure vessel breach scenarios (fuel-coolant interaction, liner-melt-through, and direct containment heating). This also credits procedure strategies where other mitigating actions are taken such as flooding the drywell. The SRA review of the dominant sequences and time to core damage affirmed that LERF did not increase the risk over that determined from the increase in CDF.Basis for Discretion: The inspectors determined that the ring lug failure was not within Exelons ability to foresee and prevent. As a result, no performance deficiency was identified. The inspectors assessment considered:1. Exelons review of emergency diesel maintenance performed in 2015 checked allconnections of the current transformer for tightness. The inspectors did not identify any gaps or deficiencies in the 2015 inspections. Inspectors also reviewed completed biennial inspections of the connection dating back to 1991 and did not identify any gaps.2. At the time of the failure, the current transformer connections did not have a time directed replacement frequency recommended by the Emergency Diesel Generator Owners Group. The inspectors did not identify any additional vendor or industry recommendations specific to the failed component or considerations specific to the failed component that existed prior to the failure.3. Industry operating experience information available to Exelon did not identify the potential for the fatigue cracking of the bent wire ring lug that was experienced.4. The bent ring lug failure was not the result of a failure on the part of Exelon staff; no standards existed on bending of the lug during installation and is considered skill of the craft.The NRC determined that it was not reasonable for Exelon to have been able to foresee and prevent this violation of NRC requirements, and as such, no performance deficiency existed. Therefore, the NRC has decided to exercise enforcement discretion in accordance with Sections 2.2.4 and 3.10 of the NRC Enforcement Policy and refrain from issuing enforcement action for the violation of technical specifications (EA-18-007). Further, because Exelons actions did not contribute to this violation, it will not be considered in the assessment process or the NRC Action Matrix. Exelons equipment corrective action program evaluation report (ECAPE) determined that the ring lug failed on the No. 2 emergency diesel generator as a result of fatigue cracking, which was initiated due to excessive stress caused by bending and twisting of the ring lug beyond limits specified in industry guidelines. The inspectors noted that the ECAPE did not provide supporting information regarding how the ring lug was bent and twisted beyond industry guidelines. Specifically, industry guidance states that ring lugs can be bent up to 90 degrees. The broken ring lug found in the No. 2 emergency diesel generator was bent at approximately 45-55 degrees per the ECAPE, which was within industry guidelines. Additionally, the ECAPE did not include specific guidance on twisting allowances for ring lugs. Exelon documented the inspectors observation in Issue Report 4089829. As a result of the inspectors observation, Exelon revised the ECAPE to say the ring lug failed on the No. 2 emergency diesel generator as a result of fatigue cracking, which was initiated due to excessive stress caused by bending and twisting of the ring lug. | |
| 05000219/FIN-2018001-01 | 2018Q1 | Severity level IV | NRC identified | Untimely Licensee Event Report for Reportable Conditions Associated with the No. 2 Emergency Diesel Generator | The inspectors identified a non-cited, Severity IV violation of 10 CFR 50.73(a)(1) for a failure to submit a licensee event report (LER) within 60 days after the discovery of an event requiring a report. Specifically, on October 9, 2017, Exelon determined that the No. 2 emergency diesel generator was inoperable for longer than the allowed outage time, which is reportable as a condition prohibited by technical specifications. Exelon did not submit an LER for this event until January 3, 2018 | |
| 05000219/FIN-2017405-01 | 2017Q4 | Licensee-identified | Licensee-Identified Violation | |||
| 05000219/FIN-2017003-01 | 2017Q3 | Green | H.12 | Self-revealing | Inadequate Augmented Offgas System Procedure Resulted in a Manual Scram | A self -revealing NCV of Technical Specification 6.8.1, Procedures and Programs, was identified because Exelon did not adequately establish and maintain the augmented offgas (AOG) system operation procedure as required by NRC Regulatory Guide 1.33, Quality Assurance Requirements (Operation), Appendix A, Section 7, Procedures for Control of Radioactivity. Specifically, Exelon procedure 350.1, Augmented Offgas System Operation, did not include adequate guidance for placing the AOG system into a recycle or shutdown configuration following a system trip. Without this guidance, Operations personnel failed to ensure the correct configuration of the AOG system following a partial trip of the system which resulted in degraded main condenser vacuum and a subsequent manual reactor scram on July 3, 2017. This issue was entered into the corrective action program as issue report 4028402. The corrective actions included placing the AOG system in the correct configuration and revising the AOG system operation procedure to provide guidance for verifying proper alignment of the AOG system when the system is in recycle or shutdown. The inspectors determined the performance deficiency was more than minor because it was associated with the Initiating Events cornerstone attribute of Procedure Quality and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the failure to establish an adequate procedure for verifying proper alignment of the AOG system following a full or partial trip of the system resulted in the AOG inlet valve being left in the open position, which allowed demineralized water to be siphoned from the flame arrestor tank and slowly fill the offgas hold- up pipe. This caused a degradation of main condenser vacuum and resulted in operators inserting a manual reactor scram on July 3, 2017. The inspectors evaluated the finding using IMC 0609, Attachment 4, Initial Screening and Characterization of Findings, and IMC 0609, Appendix A, Exhibit 1, Initiating Event Screening Questions. The inspectors determined the finding was a transient initiator that did not contribute to both the likelihood of a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of a trip to a stable shutdown condition, and therefore was of very low safety significance (Green). The finding had a cross- cutting aspect in the area of Human Performance, Avoid Complacency , because Exelon failed to recognize and plan for the possibility of mistakes or latent errors and implement appropriate error reduction tools by verifying the AOG system was properly aligned following a system trip ; instead , Operations personnel relied upon using a procedure that did not contain adequate guidance to place the AOG system in the correct configuration following a system trip (H. 12) |
| 05000219/FIN-2017404-01 | 2017Q3 | Green | P.3 | NRC identified | Security | |
| 05000219/FIN-2017002-01 | 2017Q2 | Green | P.2 | NRC identified | Inadequate Assessment of Degraded Fuel Oil Filter Impact to Emergency Diesel Generator Operability | The inspectors identified a finding associated with Exelon procedure OP-AA-108-115, Operability Determinations, because Exelon did not adequately assess the No. 2 emergency diesel generator operability with a degraded fuel oil filter. Specifically, Exelon did not adequately assess the capability of the emergency diesel generator to perform its function during its credited duration time of 72 hours. Exelon entered this issue into the corrective action program for resolution as issue report (IR) 3999576 and IR 3990799 and subsequently replaced the fuel oil filter. The finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone objective to ensure the reliability and capability of systems that respond to initiating events to prevent undesirable consequences. This issue was also similar to Example 3j of IMC 0612, Appendix E, Examples of Minor Issues, because the condition resulted in reasonable doubt of the operability of the No. 2 emergency diesel generator and additional analysis was necessary to verify operability. The inspectors evaluated the finding using Exhibit 2, Mitigating System Screening Questions, in Appendix A to IMC 0609, Significance Determination Process. The inspectors determined that this finding was a deficiency affecting the design or qualification of a mitigating structure, system, or component (SSC), where the SSC maintained its operability or functionality. Therefore, inspectors determined the finding to be of very low safety significance (Green). The finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, because Exelon did not thoroughly evaluate the issue associated with the degraded fuel oil filter and its impact to the No. 2 emergency diesel generator operability (P.2). |
| 05000219/FIN-2016004-01 | 2016Q4 | White | H.8 | NRC identified | E EMRV Failureto Stroke Due to Incorrect Reassembly | The NRC identified a preliminary White finding and associated apparent violation of Technical Specification 6.8.1, Procedures and Programs, and Technical Specification 3.4.B, Automatic Depressurization System, because Exelon failed to implement a procedure related to the maintenance of safety related equipment. Specifically, Exelon personnel did not follow electromatic relief valve (EMRV) reassembly instructions that required personnel to reinstall previously removed lock washers from the E EMRV cut-out switch lever. The incorrect reassembly caused excessive friction between the solenoid frame and the cut-out switch lever, which led to the E EMRVs failure to perform its safety function. This resulted in one inoperable EMRV for greater than the Technical Specification allowed outage time. The issue was entered into the corrective action program as issue report 2722109, and Exelons immediate corrective actions include installing new cut-out switch lever plates with increased clearances, replacing star lock washers with split ring lock washers for additional clearance, and verifying the five EMRV solenoid actuators being installed into the drywell following the most recent refueling outage were correctly assembled. The finding is more than minor because it adversely affects the human performance quality attribute of the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the missing lock washers due to the incorrect EMRV lever plate reassembly caused excessive friction between the solenoid frame and the cut-out switch lever, causing the cut-out switch lever to become bound in the energized position. This led to the E EMRVs failure to perform its safety function. The inspectors screened this issue for safety significance in accordance with Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, and determined a detailed risk evaluation was required because the E EMRV had potentially failed or was unreliable for greater than the Technical Specification allowed outage time. A detailed risk evaluation concluded that the increase in core damage frequency (CDF) related to the failure of the E EMRV is 5.4E-6/year; therefore, this finding was preliminary determined to have a low to moderate safety significance (White). Due to the nature of the failure, no recovery credit was assigned. The dominant core damage sequences involve loss of main feedwater events with operator errors resulting in failure to make-up to the 4 isolation condensers or otherwise maintain reactor vessel level and the loss of reactor pressure vessel depressurization capability (due to common cause failure of the remaining four EMRVs). The finding has a cross-cutting aspect in the area of Human Performance, Procedure Adherence, because Exelon personnel did not follow station processes. Specifically, Exelon did not follow written instructions when reassembling the E EMRV. The missing lock washers resulted in excessive friction between the solenoid frame and cut-out switch lever, causing the cut-out switch lever to become bound in the energized position, which led to the E EMRVs failure to perform its safety function. (H.8) |
| 05000219/FIN-2016002-01 | 2016Q2 | Green | Self-revealing | Inadequate Maintenance Procedure associated with Reactor Recirculation Pump Seal | A self-revealing NCV of Technical Specification 6.8.1, Procedures and Programs, was identified because Exelon did not adequately establish and maintain the reactor recirculation pump (RRP) reassembly maintenance procedures as required by NRC Regulatory Guide 1.33, Appendix A, Section 9, Procedures for Performing Maintenance. Specifically, the RRP reassembly procedure, 2400-SMM-3226.03, Reactor Recirculation Pump Mechanical Seal Rebuild Using CAN-2A Parts, did not provide critical dimensional checks for the locking plate and seal adjusting cap. This led to the incorrect reassembly of the D RRP. Exelon entered this issue into their corrective action program as issue report 2663436. The corrective actions included repairing the D RRP and revising RRP maintenance procedures to include critical dimensional information. This finding is more than minor because it is associated with the procedure quality attribute of the Initiating Events cornerstone and affected the objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown and power operation. Specifically, the incorrect reassembly of the D RRP created a leakage path, which led to an unexpected increase in reactor coolant system (RCS) unidentified leakage. As a result, the operators inserted a manual scram on April 30, 2016. The inspectors evaluated the finding using IMC 0609, Attachment 4, Initial Screening and Characterization of Findings, and IMC 0609, Appendix A, Exhibit 1, Initiating Event Screening Questions. The inspectors determined that this finding is a transient initiator that did not contribute to both the likelihood of a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition, and therefore was of very low safety significance (Green). The inspectors determined that there was no cross-cutting aspect associated with this finding since it was not representative of current Exelon performance. Specifically, in accordance with IMC 0612, the causal factors associated with this finding occurred outside the nominal three-year period of consideration and were not considered representative of present performance. | |
| 05000219/FIN-2016001-01 | 2016Q1 | Green | P.1 | NRC identified | Failure to Identify a Slower than Normal Scram Time of a Control Rod Drive | The inspectors identified an NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, because Exelon did not promptly identify and correct a condition adverse to quality. Specifically, Exelon did not identify that the scram time test result for control rod drive 18-47 was beyond the analyzed scram time, which resulted in a degraded control rod drive. Exelon entered this issue into their corrective action program. Immediate corrective actions included fully inserting the control rod drive and developing a casual analysis to determine the degraded condition. The performance deficiency is more than minor because it is associated with the configuration control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the performance deficiency affected the reliability of control rod drive 18-47 to perform its safety function due to a slower than normal scram time. The inspectors evaluated the finding using IMC 0609, Attachment 4, Initial Screening and Characterization of Findings. The inspectors determined that this finding is a deficiency that affected the design or qualification of a mitigating structure, system, or component (SSC), when the SSC maintained its operability or functionality. Therefore, the inspectors determined the finding to be of very low safety significance (Green). The finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Identification, because Exelon did not identify issues completely, accurately, and in a timely manner in accordance with the program. Specifically, Exelon did not identify that the actual scram time of control rod drive 18-47 was beyond the analyzed scram time, resulting in a degraded control rod drive. |
| 05000219/FIN-2016001-04 | 2016Q1 | Green | Licensee-identified | Licensee-Identified Violation | From 2010 to 2014, Oyster Creek made a total of four shipments of radioactive material which contained category 2 quantities of radioactive material. Oyster Creek did not implement a transportation security plan for any of these shipments, which is contrary to the requirements of 49 CFR 172, Subpart I, Safety and Security Plans. This performance deficiency adversely affected the Public Radiation Safety cornerstone attribute of Program and Process based on inadequate procedures associated with the transportation of radioactive materials. The finding was determined to be of very low safety significance (Green) because the transportation of radioactive material issue 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. This issue was documented in the Exelons corrective action program as IR 2484646. Corrective actions included contracting with a vendor to receive regular, prompt notifications of potentially applicable rule changes in the Federal Register. | |
| 05000219/FIN-2016001-02 | 2016Q1 | Green | H.8 | Self-revealing | Failure to Use Respiratory Protection as Required in RWP/ALARA Plan for Drywell Head Reassembly | A self-revealing NCV of Technical Specification 6.8.1, Procedures and Programs was identified for Exelons failure to use respiratory protection, as required in the radiation work permit (RWP)/as low as reasonably achievable (ALARA) plan 14-406 for drywell head reassembly work on October 2, 2014. The radiation protection (RP) supervisor overseeing this work removed the respiratory protection requirement for this work contrary to the RWP/ALARA requirement and without engineering approval. As a result, two workers received an unplanned intake of radioactive material that resulted in unintended internal dose. Upon identification of the intake, Exelon stopped work on this task and subsequently reinstituted the respiratory protection requirements to complete the remaining work and entered this event into their corrective action program as issue report 2390111. This finding is more than minor because it is associated with the Occupational Radiation Safety cornerstone to ensure adequate protection of the worker from radiation exposure. Specifically, without the use of respiratory protection two workers received unintended internal dose. The inspectors evaluated the finding using inspection manual chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process. The inspectors determined that this finding is of very low safety significance (Green), because it did not result in an overexposure as defined by 10 CFR 20.1201, there was no substantial potential for an overexposure, and the ability to assess dose was not compromised. This finding has a cross-cutting aspect in Human Performance, Procedural Adherence, because Exelon did not follow procedures and work instructions. Specifically, RP supervision instructed the workers that respiratory protection was not required contrary to the applicable RWP/ALARA plan. |
| 05000219/FIN-2016001-03 | 2016Q1 | White | NRC identified | Inadequate Instructions for the Flexible Coupling Hose Preventative Maintenance Resulting in an Inoperable Emergency Diesel Generator | The inspectors identified a preliminary White finding and associated apparent violation of Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because Exelon did not appropriately prescribe instructions or procedures for maintenance on the emergency diesel generator (EDG) No. 1 cooling water system to ensure the EDG cooling flexible coupling hose was maintained to support the EDG safety function. Specifically, Exelon did not have appropriate work instructions to replace the EDG cooling flexible coupling hoses every 12 years as specified by Exelons procedure and vendor information. As a result, the flexible coupling hose was in service for approximately 22 years and subjected to thermal degradation and aging that eventually led to the failure of EDG No. 1 during operation on January 4, 2016. As a consequence of this inappropriate work instruction issue, Exelon violated Technical Specification 3.7.C because EDG No. 1 was determined to be inoperable for greater than the technical specification allowed outage time of seven days. Exelons immediate corrective actions included entering the issue into their corrective action program (issue reports 2607247 and 2610027), replacing of the EDG No. 1 and No. 2 flexible coupling hoses, and initiating a failure analysis to determine the causes of the failed flexible coupling hose. This finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the ruptured flexible coupling hose caused the failure of EDG No. 1 to perform its safety function. 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, this finding required a detailed risk evaluation (DRE) because EDG No. 1 was inoperable for greater than the technical specification allowed outage time. The DRE estimated the increase in core damage frequency was 7E-6, or White (low to moderate safety significance) for this finding. This finding does not have an associated cross-cutting aspect because the performance deficiency occurred in 2005 and is not reflective of present performance. | |
| 05000219/FIN-2015004-01 | 2015Q4 | Green | P.2 | NRC identified | Preconditioning of the Standby Liquid Control Relief Valves | The inspectors identified an NCV of 10 Code of Federal Regulations (CFR) 50, Appendix B, Criterion XI, Test Control, because Exelon conducted unacceptable preconditioning of the standby liquid control (SLC) relief valves prior to American Society of Mechanical Engineers (ASME) code testing. Specifically, Exelon performed a SLC system functional test prior to performing the SLC relief valve as-found testing. Exelons immediate corrective actions included completing the as-found test prior to the functional test. Exelon entered this issue into their corrective action program (CAP) as issue report 2566036 to track the resolution of the issue. The performance deficiency is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and affects the cornerstone objective of ensuring availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Additionally, if left uncorrected, the performance deficiency could have the potential to lead to a more significant safety concern. Specifically, completion of the functional test prior to the replacement of the SLC relief valves masks the actual as-found condition by solidifying the valve internals. As a result, the as-found condition of the SLC relief valves have not been conducted and in the worst case scenario, could open below the design setpoint, which would divert flow back to the liquid poison tank instead of into the vessel to shut down the reactor during an anticipated transient without scram (ATWS) condition. The inspectors evaluated the finding using IMC 0609, Attachment 4, Initial Screening and Characterization of Findings, and determined the finding was of very low safety significance (Green) because the structure, system or component (SSC) maintained its operability. The finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation because Exelon did not thoroughly evaluate issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, Exelon did not evaluate the effect of performing the SLC system functional test prior to conducting the ASME code as-found test on the SLC relief valves. |
| 05000219/FIN-2015004-02 | 2015Q4 | Green | Self-revealing | Inadequate Problem Identification and Resolution Leading to Degradation of EPR Causing a Reactor Scram | A self-revealing finding was identified because Exelon did not adequately identify and correct conditions, per LS-AA-120, Issue Identification and Screening Process, that led to degradation of the electric pressure regulator (EPR) wiring, which resulted in an uncontrolled rise in reactor pressure and subsequent reactor scram on average power range monitor (APRM) Hi-Hi Flux. Specifically, Exelon failed to generate issue reports to document degraded EPR wiring that was previously identified, and therefore did not take corrective action prior to a reactor scram. Planned corrective actions include reinforcing with station personnel that an issue report is required when issues are identified. This finding is more than minor because it is associated with the equipment performance attribute of the Initiating Events cornerstone and adversely impacted its objective to limit the likelihood of events that upset plant stability and challenge critical safety functions. In accordance with IMC 0609, Attachment 4 and Exhibit 1 of Appendix A, the inspectors determined that this finding is of very low safety significance (Green) because the finding did not cause both a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The inspectors determined there is no cross-cutting aspect associated with this finding since it is not representative of current Exelon performance. Specifically, in accordance IMC 0612, the causal factors associated with this finding occurred outside the nominal three-year period of consideration and considered not representative of present performance. | |
| 05000219/FIN-2015003-01 | 2015Q3 | Green | NRC identified | Non-Conservative Temperature Input in the Electromatic Relief Valve Voltage Drop Calculation | The inspectors identified an NCV of 10 Code of Federal Regulations (CFR) 50, Appendix B, Criterion III, Design Control, in that Exelons measures for verifying the adequacy of design of the electromatic relief valve (EMRV) voltage drop calculation were inadequate. Specifically, non-conservative temperature inputs were used for the safety related EMRV direct current voltage drop calculation, which reduced the margin of available voltage to the EMRV solenoids. Exelon entered this issue into the corrective action program for resolution as issue report 2522756, and corrective actions included revising the calculation to include the correct temperature values and conduct an extent of condition of other voltage drop calculations that could have similar temperature values. The performance deficiency is more than minor because it is associated with the design control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, lower voltage to the EMRV solenoid at higher temperatures could affect the reliability and capability of the EMRV to perform its design function. In addition, the performance deficiency is determined to be more than minor because it is similar to example 3.j of NRC IMC 0612, Appendix E, Example of Minor Issues, in that as a result of the calculation errors and the magnitude of the decrease of margin, there was a reasonable doubt on the operability of the component. The inspectors evaluated the finding using 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix A, Exhibit 2, Mitigating System Screening Questions. The inspectors determined that this finding is a deficiency that affected the design or qualification of a mitigating structure, system or component (SSC), where the SSC maintained its operability or functionality. Therefore, inspectors determined the finding to be of very low safety significance (Green). The finding is not assigned a cross-cutting aspect because it is not reflective of current performance. Specifically, the last time Exelon had an opportunity to evaluate this issue was in 2010 when Exelon identified that the EMRV solenoid voltage had low margin. | |
| 05000219/FIN-2015002-02 | 2015Q2 | Green | H.9 | Self-revealing | Failure Rates Exceed Twenty Percent for Annual Requalification Exam | A self-revealing finding was identified associated with inadequate licensed operator performance during licensed operator requalification exams in accordance with TQ-AA-150, Operator Training Program. Specifically, two of seven crews failed the simulator scenario portion of the requalification examinations. As an immediate corrective action, the crews that failed were restricted from licensed duties. Exelon entered this issue into the corrective action program, and facility training staff remediated the crews (the crews were retrained and successfully retested), and those crews were returned to licensed duties. 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 of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, two of seven crews failed to demonstrate a satisfactory understanding of the knowledge and abilities required to safely operate the facility under normal, abnormal, and emergency conditions. The inspectors determined the finding to be of very low safety significance (Green) because it is related to requalification exam results, did not result in a failure rate of greater than forty percent, and the two crews were remediated (i.e., the crews were retrained and successfully retested) prior to returning to shift. This finding has a cross-cutting aspect in the area of Human Performance, Training, because Exelon staff did not provide adequate operator requalification training to maintain a knowledgeable, technically competent workforce. |
| 05000219/FIN-2015008-01 | 2015Q2 | Severity level IV | NRC identified | Use of an Analytical Method to Determine the Core Operating Limits Without Prior NRC Approval | The NRC identified a Severity Level lV non-cited violation (NCV) of Technical Specification (TS) 6.9.1.f.2 in that Exelon did not obtain NRC approval prior to using a specific analytical method to determine the core operating limits. Specifically, Exelon used an analytical method (TRACG04P) to determine the core operating limits (the average power range monitor protection settings which were identified in the Core Operating Limits Report (COLR)); however, that particular analytical method was not previously reviewed and approved by the NRC prior to Exelons use. Exelon submitted a corrective action issue report (IR) to evaluate the condition (IR2482042). The team determined that Exelon did not comply with TS 6.9.1.f.2 requirements in that Exelon used an analytical method to determine the core operating limits without prior NRC approval. The team determined that this was a performance deficiency that was within Exelons ability to foresee and correct. Because the issue had the potential to affect the NRCs ability to perform its regulatory function, the team evaluated this performance deficiency in accordance with the traditional enforcement process. Using the Enforcement Manual, the team characterized the violation as Severity Level IV because the underlying analytical method required NRC approval prior to use. Because this violation involves the traditional enforcement process and does not have an underlying technical violation that would be considered more-than-minor within the Reactor Oversight Process (ROP), the team did not assign a cross-cutting aspect to this violation in accordance with IMC 0612, Power Reactor Inspection Reports, Section 07.03.c. | |
| 05000219/FIN-2015002-03 | 2015Q2 | Green | H.11 | Self-revealing | Reactor Water Cleanup Procedure Not Followed Resulting in a Level Transient | A self-revealing NCV of Technical Specification 6.8.1(a), Procedures and Programs, was identified because Exelon did not follow procedure 303, Reactor Cleanup Demineralizer System, during the system restoration on March 26, 2015. Specifically, during startup from a forced outage (1F36), Exelon did not follow procedure 303, which required correct valve lineups for system restoration of reactor water cleanup (RWCU) after system isolation. This resulted in decreasing reactor water level, which was automatically terminated by a second RWCU isolation. Exelon entered this issue into the corrective action program. Planned corrective actions include enhancing operator training in system knowledge and procedure compliance and revising startup procedures. This finding is determined to be more than minor because it is associated with the human performance attribute of the Initiating Events cornerstone and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, Exelon did not properly lineup the RWCU system after isolation, which resulted in a water level transient and challenging the critical safety function of inventory control. This finding is determined to be of very low safety significance (Green), because it did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. This finding has a cross-cutting aspect in the area of Human Performance, Challenge the Unknown, because Exelon did not recognize and plan for the possibility of mistakes, or implement appropriate error reduction tools. Specifically, the operators did not stop and fully communicate plant condition after the initial RWCU isolation. Consequently, operators opened the RWCU system inlet valve due to the increasing water level without following procedure guidance. |
| 05000219/FIN-2015002-04 | 2015Q2 | Green | H.11 | Self-revealing | Reset of the Automatic Voltage Regulator Controller Led to an Automatic Reactor Scram | A self-revealing finding was identified because Exelon did not properly screen work in accordance with MA-AA-716-010, Maintenance Planning. Specifically, on September 12, 2014, Exelon did not screen the automatic voltage regulators (AVR) human machine interface (HMI) post-maintenance test per the maintenance planning procedure. As a result, on October 12, 2014, Exelon personnel performing the post-maintenance test did not have a work order, which would have included plant configurations and limitations associated with the test. This led to an automatic reactor scram. Exelon entered this issue into the corrective action program. Planned corrective actions include reinforcing with work planners that a work order is required for similar work activities. This finding was determined to be more than minor because it is associated with the human performance attribute of the Initiating Events cornerstone and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during plant operation. Specifically, resetting the three AVR controllers caused an automatic plant scram. This finding is determined to be of very low safety significance (Green), because it did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. This finding has a cross-cutting aspect in the area of Human Performance, Challenge the Unknown, because Exelon did not recognize and plan for the possibility of mistakes, or implement appropriate error reduction tools. Specifically, on October 12, 2014, Exelon personnel did not stop when faced with the uncertain situation of the HMI screen that did not respond as expected. |
| 05000219/FIN-2015002-01 | 2015Q2 | Green | P.2 | NRC identified | Inadequate Assessment of 4k Emergency Switchgear Roll-Up Door Degraded Floor Gasket | The inspectors identified a finding associated with Exelon procedure, OP-AA-108-115, Operability Determinations, because Exelon did not adequately assess a degraded floor gasket for the D emergency 4 kilovolt (kV) switchgear roll-up door. Specifically, Exelon did not adequately assess the flood and fire functionality of the degraded gasket, which is credited to provide protection to safety-related D emergency 4kV switchgear during a postulated internal flood event and to contain the carbon dioxide (CO2) gaseous suppression system during a postulated fire within the D switchgear room. Exelon entered this issue into the corrective action program. Planned corrective actions include reinforcing the operability determination procedure and enhancing operator training in fire and flood functionality of gaskets. Additional corrective actions included repairing the gasket and performing a detailed analysis of the ability of degraded gasket to meet its flooding and fire function. This finding is more than minor because it is associated with the protection against external factors attribute of the Mitigating Systems cornerstone, and affected the cornerstone objective of ensuring the reliability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the degraded floor gasket could have resulted in increased water level in the D emergency 4kV switchgear room during a postulated internal flood due to a fire water pipe rupture, therefore affecting the reliability of the D emergency 4k switchgear to perform its safety function. In addition, the degraded floor gasket could have resulted in CO2 leakage out of the D emergency 4k switchgear room during a postulated fire in that room, therefore affecting the reliability of the D emergency 4k switchgear gaseous suppression system to perform its safety function. The inspectors determined that this finding is of very low safety significance (Green) because it is a deficiency that affected the design or qualification of a mitigating structure, system, or component (SSC), where the SSC maintained its operability or functionality. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, because Exelon did not thoroughly evaluate issues to ensure that resolutions address the causes and extent of conditions commensurate with their safety significance. Specifically, Exelon staff did not thoroughly evaluate the issue associated with the degraded floor gasket for fire and flood functionality. |
| 05000219/FIN-2015008-02 | 2015Q2 | Green | NRC identified | Untimely Corrective Actions to Restore Design Conformance of Two SDV Vent & Drain Valves Pressure Regulator Valves | The NRC identified an NCV of Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion XVI, Corrective Action, for failure to promptly correct a condition adverse to quality. Specifically, corrective actions to restore design conformance of scram discharge volume (SDV) vent and drain valve pressure regulator valves V-6-961 and V-6-962 were not taken at the first opportunity of sufficient duration which was refueling outage 25 (1R25). Additionally, justification of the basis for deferral of corrective actions beyond the restart from 1R25 on October 2014, was not documented, reviewed, or approved by site management and/or oversight organizations as required by station procedure OP-AA-108-115, Section 4.5.5. Consequently, two non-conforming pressure regulator valves which perform a safety-related function remained installed following plant startup from 1R25, without appropriate evaluation and approval. Immediate corrective action included licensee determination that V-6-961 and 962 and the associated SDV vent and drain valves (V-15-119 and 121) remained operable, but non-conforming. Exelon entered the issue into their corrective action program as IR 2482851. The finding was more than minor because it was associated with the design control and barrier performance attributes of the Barrier Integrity cornerstone and adversely affected the cornerstone objective of ensuring the operational capability of the containment barrier to protect the public from radionuclide releases caused by accidents or events. Additionally, the finding was similar to example 5.c in Appendix E of Inspection Manual Chapter (IMC) 0612, because the control rod drive system was returned to service following 1R25 with two non-conforming (non-safety-related) pressure regulator valves installed in a safety-related application. The team determined the finding was of very low safety significance because it did not affect the reactor coolant system (RCS) boundary; did not affect the radiological barrier function of the control room, auxiliary building, or spent fuel pool systems or boundaries; and did not represent an actual open pathway in containment or involve a reduction in the function of hydrogen igniters. The team assigned a cross-cutting aspect in the area of Human Performance, Consistent Process (aspect H.13) because the organization did not use a consistent systematic approach to evaluate component operability after Exelon upgraded the classification of three pressure regulator valves from a non-safety to a safety-related status. | |
| 05000219/FIN-2015001-05 | 2015Q1 | Green | Licensee-identified | Licensee-Identified Violation | Technical Specification 6.8.1b states that, Written procedures shall be established, implemented, and maintained covering surveillance and test activities of equipment that affects nuclear safety and radioactive waste management equipment. Contrary to the above, from August 2012 through September 2013, Exelon took no action following receipt of ten lubricating oil analysis report results taken from two emergency diesel generator No. 2 sample locations which indicated silver content at 1.0 ppm, which exceeded procedural action levels. Specifically, Exelon maintenance procedure MA-AA-716-230-1001, Oil Analysis Interpretation Guideline, Section 3 governs safety system oil analyses and describes actions to be taken when equipment wear metals exceed specific thresholds, as obtained through monthly oil analysis. Section 3 of procedure MA-AA-716-230-1001 lists potential actions to be taken when oil analysis results indicate silver content above 0.3 and 0.7 ppm respectively. These actions include resampling immediately to verify abnormal results, performing confirmatory testing using more accurate methods if required, reviewing all vibration and thermography data immediately for adverse trends, and contacting the equipment manufacturer for additional assistance. Exelon identified this issue on October 21, 2013, during the performance of a 24-month lubricating oil system inspection on the emergency diesel generator No. 2 when silver metal shavings were found in the main lubricating oil filter housing and in the sump below cylinder #15. The inspectors determined that the failure to identify an out-of-specification lubricating oil sample result on numerous occasions was a performance deficiency that was within Exelons ability to foresee and correct. The inspectors determined that the issue adversely impacted the reliability of the safety-related emergency diesel generator in that the wrist pin bearing was degraded and had partially failed. The inspectors determined that the issue was of very low safety significance (Green) because it did not: affect design or qualification; represent a loss of system or function; exceed technical speciation allowed outage times; and involve external events. Exelon entered this issue into the corrective action program as IR 1575045. | |
| 05000219/FIN-2015001-04 | 2015Q1 | Green | Licensee-identified | Licensee-Identified Violation | Technical Specification 3.5.B, Secondary Containment, requires in part, that secondary containment integrity be maintained at all times when the reactor vessel head and the drywell head are in not in place. Technical Specification 1.14, Secondary Containment Integrity, requires in part, that the standby gas treatment system is operable. Technical Specification 4.5.G.3 specifies that with the trunnion room door open and the trunnion room is isolated from secondary containment in support of outage activities, testing of the standby gas treatment system to be performed to demonstrate the capability to maintain 14 inch of water vacuum under calm wind conditions and a standby gas treatment system filter train flow rate of not more than 4000 cfm. Contrary to Technical Specification 3.5.B, on September 20, 2014, with the reactor vessel head and drywell head removed for the refueling outage, Exelon determined that they did not have secondary containment integrity when performing testing to demonstrate standby gas treatment system capability in accordance with Technical Specification 4.5.G.3 and subsequently found that the outer railroad air lock personnel access hatch had not been closed properly, which prevented a proper vacuum from being achieved. Exelon entered this issue into the corrective action program as IR 2383852. Using guidance in IMC 0609, Appendices G and H, the inspectors determined that this finding was of very low safety significance (Green) because the decay heat values were low and the reactor water level inventory was above that required to move irradiated fuel. | |
| 05000219/FIN-2015001-03 | 2015Q1 | Severity level IV | NRC identified | Incomplete 50.72 and 50.73 Reports Associated with Secondary Containment Integrity | The inspectors identified a Severity Level IV NCV of 10 CFR 50.9(a) in that Exelon did not provide complete information in reports submitted per 10 CFR 50.72 and 10 CFR 50.73. Specifically, a licensee event report (LER) submitted on November 18, 2014, did not discuss a separate, partially opened secondary containment door that was discovered during the same time frame, which could have prevented the fulfillment of the safety function of secondary containment, and therefore was required to be discussed in the original LER. Exelon entered this issue into their corrective action program as IR 2440641. Planned corrective actions include revising the original LER to add a discussion of the partially opened secondary containment door. The inspectors determined that not providing a complete report in accordance with 10 CFR 50.9(a) is a performance deficiency that was reasonably within Exelons ability to foresee and correct and should have been prevented. Because the issue had the potential to affect the NRCs ability to perform its regulatory oversight function, the inspectors evaluated this performance deficiency in accordance with the traditional enforcement process. In accordance with Section 2.2.2.d of the NRC Enforcement Policy, the inspectors determined that the performance deficiency identified with the reporting aspect of the event is a Severity Level IV violation because it is of more than minor concern with relatively inappreciable potential safety significance and is related to findings that were determined to be more than minor issues. In accordance with IMC 0612, Appendix B, this issue was not assigned a cross-cutting aspect. | |
| 05000219/FIN-2015001-02 | 2015Q1 | Green | H.5 | NRC identified | Inadequate Post Maintenance Testing for Emergency Service Water Pump Breaker | The inspectors identified an NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings for Exelons failure to develop an adequate post maintenance test to determine operability of the A emergency service water pump breaker. Specifically, the corrective maintenance work performed on April 16, 2013, did not correct the cause of the failure and Exelon did not perform an adequate post maintenance test to verify conditions had been corrected. As a result, the emergency service water system was returned to service even though it did not meet all the requirements for operability. The issue was not identified and resolved until a subsequent surveillance test on April 17, 2013, which identified a failed breaker. Exelon entered this issue into their corrective action program (IR 2471069). Planned corrective actions include revising work order activities to specify the correct post maintenance test. This 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 and reliability of the systems that respond to initiating events. Specifically, the inadequate post maintenance test for A emergency service water pump breaker on April 16, 2013, led to the A emergency service water pump failing to perform its function during the subsequent surveillance testing on April 17, 2013. The inspectors assessed this finding in accordance with the IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The inspectors concluded that this finding did not represent an actual loss of function of the emergency service water system for greater than its technical specification allowed outage time (15 days). Therefore, the inspectors determined that this finding is of very low safety significance (Green). The inspectors determined that this finding had a cross-cutting aspect in the area of Human Performance, Work Management, in that Exelons work planning and executing of work activities did not include documented instructions for performing an adequate post maintenance test. |
| 05000219/FIN-2015001-01 | 2015Q1 | Green | H.8 | NRC identified | Post Maintenance Test Results Were Not Evaluated to Assure that Technical Specifications Requirements Were Satisfied | The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, when Exelon did not document and adequately evaluate test results to assure that test requirements had been satisfied. Specifically, Exelon did not perform the proper post maintenance test procedure to assure that the requirements of Technical Specification 4.5.G.3 were satisfied following installation of a temporary modification to secondary containment. Exelon entered this issue into the corrective action program for resolution as issue report (IR) 2440643. Corrective actions include revising the process to perform the correct post maintenance test to ensure Technical Specification 4.5.G.3 is met. This finding is more than minor because it is associated with the configuration control (Standby Gas Trains) attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The inspectors evaluated the finding using IMC 0609.04, Initial Characterization of Findings, issued June 19, 2012, and IMC 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process: Phase 1 Initial Screening and Characterization of Findings, issued May 9, 2014. Because the finding degraded the ability to close or isolate secondary containment, the inspectors were required to further assess the finding using IMC 0609, Appendix H, Containment Integrity Significance Determination Process, issued May 6, 2004. The inspectors determined that this finding is of very low safety significance (Green) because the decay heat values were low, given that the unit had been shut down for approximately three days, and reactor water level was greater than that required for movement of irradiated fuel assemblies within the reactor pressure vessel. This finding has a cross-cutting aspect in the area of Human Performance, Procedure Adherence, because Exelon personnel did not perform the post maintenance test specified by the work order. |
| 05000219/FIN-2014005-01 | 2014Q4 | Green | P.2 | NRC identified | Reactor Head Cooling Spray Piping Flange Misalignment | The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, because Exelon did not promptly correct a condition adverse to quality associated with the reactor head cooling (RHC) spray line 2-inch upper flange which was installed in a configuration that exceeded the allowable acceptance criteria. Specifically, Exelon staff identified a misaligned flange condition in IR 845395 but did not correct the deficiency by evaluation, repair or replacement during the 1R22 refueling outage in 2008 or subsequently during the 1R23 and 1R24 refueling outages. Exelon staff completed corrective actions to replace the flange during the 1R25 refueling outage after the NRC inspector questioned the acceptability of this condition. Exelon staff entered this issue into their corrective action program as IR 2385501. The finding is more than minor because it is associated with the equipment performance attribute of the Initiating Events cornerstone and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, misalignment of the RHC spray line flange was greater than that provided in Oyster Creek pipe specifications and resulted in additional stresses in the flange weld. This condition was identified by Exelon staff as a possible contributor to the occurrence of a through wall crack and leak in the N7B upper flange socket weld joint that was identified and repaired in November 2012, but the misalignment was not corrected at that time. The inspectors screened this issue using IMC 0609.04, Phase 1 Initial Screening and Characterization of Findings, and Exhibit 1 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, and determined this finding was of very low safety significance (Green). The inspectors determined that this finding had a Problem Identification and Resolution cross-cutting aspect because Exelon did not evaluate and take timely corrective actions to address the long-standing repetitive flange alignment issue of the reactor head cooling spray piping flange connection to reactor pressure vessel head N7B nozzle. |
| 05000219/FIN-2014005-04 | 2014Q4 | Green | H.8 | NRC identified | Procedures Not Implemented During Plant Shutdown | The inspectors identified an NCV of Technical Specification 6.8.1, Procedures and Programs, because Oyster Creek operators did not adequately implement procedures when performing a plant shutdown. Specifically, the operators did not ensure that all personnel on shift had received Just-in-Time-Training for their role in the shutdown; operators did not perform a reactivity Heightened Level Awareness brief for the shutdown, and did not insert source range monitors (SRMs) in accordance with procedure. These performance deficiencies contributed to two unanticipated criticalities during the shutdown. Exelon entered this issue into their corrective action program as IR 2412093 and conducted a root cause analysis. This finding is more than minor because it affected the procedure quality attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the reliability and capability of systems that respond to initiating events. Specifically, Exelon did not implement procedures during the plant shutdown which contributed to two unanticipated returns to criticality which required operator action to mitigate. The inspectors screened this issue using IMC 0609.04, Initial Characterization of Findings, Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, and IMC 0609 Appendix M, Significance Determination Process Using Qualitative Criteria. Inspectors determined this finding was of very low safety significance (Green). This finding has a cross-cutting aspect in the area of Human Performance, Procedure Adherence, because licensed operators did not implement processes, procedures and work instructions during the plant shutdown. |
| 05000219/FIN-2014005-03 | 2014Q4 | Green | H.7 | NRC identified | Plant Shutdown Procedure was Inadequate for Soft Shutdown | The inspectors identified a non-cited violation (NCV) of Technical Specification 6.8.1, Procedures and Programs, because Exelon did not adequately establish and maintain the plant shutdown procedure. Specifically, the procedure was not adequate in that it did not contain precautions concerning rod insertion when reactor power is below the point of adding heat; operational limitations on plant cooldown when power is below the point of adding heat; and contingency actions for re-criticality during shutdown. Exelon entered this issue into their corrective action program as IR 2412093 and conducted a root cause analysis. This finding is more than minor because it affected the procedure quality attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the reliability and capability of systems that respond to initiating events. Specifically, the plant shutdown procedure did not contain precautions to continuously insert control rods when reactor power is less than the point of adding heat, did not define operational considerations for limiting reactor cooldown, and did not contain contingency actions for return to criticality during shutdown. The inspectors screened this issue using IMC 0609.04, Initial Characterization of Findings, Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, and IMC 0609 Appendix M, Significance Determination Process Using Qualitative Criteria. Inspectors determined this finding was of very low safety significance (Green). This finding has a cross-cutting aspect in the area of Human Performance, Documentation, because Exelon did not ensure that the shutdown procedure contained adequate controls for soft shutdown. |
| 05000219/FIN-2014010-01 | 2014Q4 | Green | P.5 | Self-revealing | Failure to Evaluate a Temporary Configuration Change | A self-revealing finding (FIN) of very low safety significance was identified for Exelons failure to implement the temporary configuration change program when a temporary repair was performed on condenser bellows expansion joint Y-1-26. The temporary repair impacted the design function of Y-1-26 and led to failure of the downstream side of the bellows, causing a loss of condenser vacuum and manual reactor scram on July 11, 2014. Exelon replaced both the expansion joint Y-1-26 and the 2nd stage reheater steam supply relief valve V-1-132 on July 11, 2014, during forced outage 1F35. Exelon entered this issue into the corrective action program (IR 2422831). This finding was more than minor because it was associated with the Design Control attribute of the Initiating Events cornerstone, and adversely affected the 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 inspectors determined that this finding was of very low safety significance (Green) using Exhibit 1 of NRC IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, because the finding did not cause both a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition (e.g. loss of condenser, loss of feed water). The inspectors determined that this finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Operating Experience, because Exelon did not systematically and effectively evaluate relevant internal operating experience related to a similar condenser bellows expansion joint failure in 1986. |
| 05000219/FIN-2014005-02 | 2014Q4 | White | NRC identified | Inadequate Review of Change in Maintenance Process Results in Inoperable Emergency Diesel Generator | The inspectors identified a preliminary White finding and an associated apparent violation of 10 CFR 50, Appendix B, Criterion III, Design Control, because Exelon staff did not review the suitability of the application of a different maintenance process at Oyster Creek that was essential to a safety-related function of the emergency diesel generators (EDG). Specifically, in May 2005, Exelon staff changed the method for tensioning the cooling fan belt on the EDG from measuring belt deflection to belt frequency and did not verify the adequacy of the acceptance criteria stated for the new method. As a result, Exelon staff did not identify that the specified belt frequency imposed a stress above the fatigue endurance limit of the shaft material, making the EDG cooling fan shaft susceptible to fatigue and subsequent failure on July 28, 2014. As a consequence, Exelon also violated Technical Specification 3.7.C, because the EDG No. 2 was determined to be inoperable for greater than the technical specification allowed outage time. Exelons immediate corrective actions included entering the issue into their corrective action program as issue report (IR) 1686101, replacing the EDG No. 2 fan shaft, examining the EDG No.1 fan shaft for extent of condition, and performing a failure analysis to determine the causes of the broken shaft. This finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). 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 screened the finding for safety significance and determined that a detailed risk evaluation was required because the finding represented an actual loss of function of a single train for greater than its technical specification allowed outage time. The detailed risk evaluation concluded that the increase in core damage frequency was 5.1E-6, or White (low to moderate safety significance). This finding does not have an associated cross-cutting aspect because the performance deficiency occurred in 2005 and is not reflective of present performance. | |
| 05000219/FIN-2014009-01 | 2014Q4 | Yellow | NRC identified | Inadequate Application of Materials, Parts, Equipment, and Processes Associated with the Electromatic Relief Valves | The NRC identified a preliminary Yellow finding and associated apparent violation of 10 CFR 50, Appendix B, Criterion III, Design Control, and Technical Specification 3.4.B, Automatic Depressurization System, because the station did not establish adequate measures for selection and review for suitability of application of materials, parts, equipment, and processes that are essential to the safety-related functions of the electromatic relief valves (EMRVs). The violation was also preliminarily determined to meet the IMC 0305, Section 11.05, criteria for treatment as an old design issue. Specifically, on June 20, 2014, during refurbishment of EMRVs that were removed from the plant during the 2012 refueling outage, Exelon personnel identified deficiencies with the B and D EMRVs. As part of the planned EMRV actuator testing and refurbishment activities, Exelon personnel conducted bench testing on June 26, 2014. Both valves did not stroke satisfactorily and resulted in two inoperable EMRVs for greater than the Technical Specification allowed outage time of 24 hours. Exelons immediate corrective actions included placing this issue into the corrective action program as issue report 1679428 and redesigning the EMRV actuators to ensure the spring is on the outside of the guide bushing, therefore removing the possibility of the spring entering the guide bushing area and subsequently jamming the actuator causing valve failure. All of the actuators were replaced with redesigned actuators during the refueling outage in October 2014. In addition, Exelon issued a 10 CFR Part 21 report to inform the industry of the deficient EMRV actuator design. This finding is more than minor because it adversely affected the design control quality 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. Specifically, the design deficiency of the EMRVs and the inadequate maintenance process led to the inability of the B and D EMRVs to perform their safety function. The inspectors screened this issue for safety significance in accordance with IMC 0609, Appendix A, Exhibit 2, and determined a detailed risk evaluation was required because the EMRVs were potentially failed or unreliable for greater than the Technical Specification allowed outage time. As described in Attachment 3 to this report, a detailed risk evaluation concluded that the increase in core damage frequency (CDF) related to failure of the B and D EMRVs is in the mid E-5 range; therefore, this finding was preliminarily determined to have a substantial safety significance (Yellow). Due to the nature of the failures, no recovery credit was assigned. The dominant sequences included loss of main feedwater with failures of the isolation condensers, and failure to depressurize. This finding does not represent an immediate safety concern because Exelon replaced all of the actuators with the redesigned actuators during the refueling outage in October 2014. Further, the NRC is considering treatment of this finding as an old design issue because the condition existed since the original installation of the EMRVs, and is not indicative of current licensee performance. Additional details are discussed in Attachment 1. The inspectors determined that this finding did not have a cross-cutting aspect because the most significant contributor to the performance deficiency was not reflective of current licensee performance. Specifically, the inspectors determined that the performance deficiency existed since original installation of the EMRVs. Although an opportunity to identify this issue following original installation occurred in 2006 when Quad Cities changed the EMRV actuator design due to similar issues, the inspectors could not conclude that the issue would have likely been identified during that period since a Part 21 Report was not issued to inform the industry and NRC of the design change and industry operating experience focused on plants that completed or were scheduled to complete an extended power uprate. | |
| 05000219/FIN-2014405-01 | 2014Q3 | Green | Licensee-identified | Licensee-Identified Violation | ||
| 05000219/FIN-2014405-02 | 2014Q3 | Green | Licensee-identified | Licensee-Identified Violation | ||
| 05000219/FIN-2014004-01 | 2014Q3 | Green | H.7 | NRC identified | Inadequate Evacuation Time Estimate Submittals | The inspectors identified an NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50.54(q)(2), 10 CFR 50.47(b)(10), and 10 CFR Part 50, Appendix E, Section IV.4, for failing to maintain the effectiveness of the Oyster Creek emergency plan as a result of failing to provide the station evacuation time estimate (ETE) to the responsible offsite response organizations (OROs) by the required date. Exelon entered this issue into its corrective action program as issue reports 1525923 and 1578649. Additionally, Exelon resubmitted a new revision of the Oyster Creek ETE to the NRC on April 4, 2014, and the NRCs review of that ETE is documented in Section 1EP4 of this report. The performance deficiency is more than minor because it is associated with the Emergency Preparedness cornerstone attribute of procedure quality and adversely affected the cornerstone objective of ensuring that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The ETE is an input into the development of protective action strategies prior to an accident and to the protective action recommendation decision making process during an accident. Inadequate ETEs have the potential to reduce the effectiveness of public protective actions implemented by the OROs. The finding is determined to be of very low safety significance (Green) because it is a failure to comply with a non-risk significant portion of 10 CFR 50.47(b)(10). The cause of the finding is related to a cross-cutting aspect of Human Performance, Documentation, because Exelon did not appropriately create and maintain complete, accurate, and up-to-date documentation. |
| 05000219/FIN-2014003-01 | 2014Q2 | Green | P.1 | NRC identified | Failure to Identify and Correct High Oil Level in D Emergency Service Water Pump Upper Motor Bearing ( | The NRC inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, because Exelon did not promptly identify and correct a condition adverse to quality. Specifically, Exelon did not identify and correct a high oil level condition caused by water intrusion in the D emergency service water pump upper motor bearing resulting in an inoperable D emergency service water pump. Following identification of the high level by the inspections, Exelon entered this issue into their corrective action program as issue report 1645010. Exelons corrective action included sealing joints on top of the motor that are susceptible to water intrusion. The inspectors determined that inadequate identification and resolution of the condition adverse to quality into the corrective action program is a performance deficiency that was within Exelons ability to foresee and correct. This finding is more than minor because it is associated with the configuration control of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the performance deficiency affected the reliability of an emergency service water pump to perform its safety function. This issue was also similar to Example 3j of NRC IMC 0612, Appendix E, Examples of Minor Issues, because the condition resulted in reasonable doubt of the operability of emergency service water system. The inspectors determined that this finding was a deficiency affecting the design or qualification of a mitigating structure, system, or component (SSC), where the SSC maintained its operability or functionality. Therefore, inspectors determined the finding to be of very low safety significance (Green). The finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, because Exelon did not identify the issue associated with the high oil level in the emergency service water pump upper motor bearing oil in a timely manner in February and April 2014. |
| 05000219/FIN-2014404-01 | 2014Q2 | Green | H.1 | NRC identified | Security | |
| 05000219/FIN-2014002-01 | 2014Q1 | Green | H.11 | NRC identified | Untimely Performance of a 50.65 a(4) Risk Evaluation during a Maximum Emergency Generation Action | The inspectors identified a Green non-cited violation of 10 CFR Part 50.65(a)(4), Requirements for monitoring the effectiveness of maintenance at nuclear power plants, because Exelon did not reassess and manage risk after the grid operator declared a maximum emergency generation action, prior to performing maintenance on the B control rod drive pump on January 30, 2014. The inspectors identified that Exelon assessment of risk was green; however, if the emergency generation action had been included in the assessment, the risk would have been yellow requiring Exelon to perform compensatory actions to limit the risk to the unit. Exelon entered this issue into their corrective action program as issue report 1614625. The inspectors determined that Exelons failure to assess and manage risk prior to performing maintenance on the B control rod drive pump after the grid operator declared a maximum emergency generation was a performance deficiency that was reasonably within Exelons ability to foresee and correct. This finding is more than minor because it is associated with the configuration control 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. The inspectors used NRC inspection manual chapter 0609, appendix K, flowchart 2, Assessment of Risk Management Actions, to determine the significance of this finding. The inspectors determined that the finding is of very low safety significance (Green). This finding has a cross-cutting aspect in the area of Human Performance because operators did not stop when faced with uncertain conditions and evaluate and manage risks before proceeding as scheduled. Specifically, the operators continued maintenance without reassessing risk after the inspectors questioned the rationale for not entering the grid emergency procedure (H.11). |
| 05000219/FIN-2014002-02 | 2014Q1 | Green | Self-revealing | Corrective Action to Prevent Recurrence Ineffective to Preclude Repetition of a Significant Condition Adverse to Quality | A self-revealing Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified when the corrective action to prevent recurrence of a significant condition adverse to quality did not preclude repetition of the event. Specifically, Exelon generated a corrective action to prevent recurrence during a root cause evaluation for a reactor scram caused by spiking on intermediate range monitor nuclear instruments that occurred in May 2004. In October 2013 another scram caused by spiking on the intermediate range nuclear instrument occurred, which Exelon subsequently determined to be a repeat of the May 2004 event. Exelon entered this issue into their corrective action program as issue report 1567196. The inspectors determined that Exelons failure to preclude repetition of a significant condition adverse to quality was a performance deficiency that was reasonably within Exelons ability to foresee and correct. This performance deficiency is more than minor because it is associated with the equipment performance attribute of the Initiating Events cornerstone and adversely affected the 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 this finding was determined using NRC IMC 0609, appendix A, exhibit 1. This finding screened as very low safety significance (Green), because the finding did not contribute to both the likelihood of a reactor trip and likelihood that mitigation equipment or functions would not be available. The finding does not have a cross cutting aspect as it is not reflective of current performance because the root cause and associated corrective actions to prevent reoccurrence were from 2004. | |
| 05000219/FIN-2013004-01 | 2013Q3 | Green | P.2 | NRC identified | Physical Change To Security Feature Causes Flood Control Feature To Be Ineffective | The inspectors identified a Green, non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion III, Design Control, in that, Exelon did not ensure applicable regulatory requirements and design basis for the emergency diesel generators were correctly translated into instructions. Specifically, the inspectors determined that Exelon did not ensure that the design basis for flood control features was maintained and correctly translated into specifications, drawings, procedures and instructions for the installation of a security wall modification around the emergency diesel generator building which affected the probable maximum precipitation flood protection features of the building. Exelon entered this issue into the corrective action program for resolution as IR 1546148 and implemented corrective actions which included removing soil and re-grading the area adjacent to the building to improve the storm water runoff flow patterns. The performance deficiency was more than minor because the finding affected the protection against external factors attribute of the mitigating systems cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined this finding involved the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding or severe weather initiating event; however, it did not involve the assumption that the protected equipment or safety function was completely failed or unavailable, and did not involve the total loss of any safety function, identified by Exelon through a PRA, IPEEE or similar analysis that contributes to external event initiated core damage accident sequences. Therefore, the inspectors determined the finding to be of very low safety significance (Green). This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, because Exelon, beginning in 2009 and as recent as September 2012, did not thoroughly evaluate the problem such that the resolution addressed the cause of an issue that potentially impacted nuclear safety. |
| 05000219/FIN-2013003-03 | 2013Q2 | NRC identified | Difference In Interpretation Of Guidance Contained NEI 99-02 Submitted Into TheFrequently Asked Question Process | Exelon planned a downpower to 70% power for end of cycle control rod conditioning starting at 2300 on September 28, 2012 with a target completion time of 0700, September 29, 2012. At 1855 on September 28, 2012, operators lowered power to 85% in order to remove the A-north condenser from service to allow maintenance workers to repair a pre-existing patch on the A-north condenser circulating water discharge piping which had exhibited an increased leakage rate earlier in the day. At 2305 on September 28, 2012, operators lowered power to 70% to perform the planned control rod conditioning evolution. At 0033, on September 29, 2012, operators commenced raising power to full power following completion of the control rod conditioning. At 0115 on September 29, 2012, Exelon decided to stop the power ascension and hold power at 80% to assess repair options for the A-north circulating water piping. At 0302 on September 29, 2012, operators lowered power to 70% to set conditions for maintenance workers to perform repairs. Repairs to the A-north circulating water piping were completed by station personnel at 1539 and operators returned the plant to full power at 1843 on September 29, 2012. NEI 99-02 Regulatory Assessment Performance Indicator Guideline, Revision 6 defines an unplanned power change as follows: Unplanned changes in reactor power are changes in reactor power that are initiated less than 72 hours following the discovery of an off-normal condition, and that result in, or require a change in power level of greater than 20% of full power to resolve. Unplanned changes in reactor power also include uncontrolled excursions of greater than 20% of full power that occur in response to changes in reactor or plant conditions and are not an expected part of a planned evolution or test. The inspectors determined that this downpower should count towards the unplanned power changes performance indicator for the following reasons: Power was initially lowered in response to an off-normal condition, an increase in leakage from a patch on the A-north condenser circulating water discharge piping. Power was lowered 4 hours earlier than the planned control rod conditioning starting time and repairs required an additional 8 hours to complete beyond the planned completion time for the control rod conditioning. Power was lowered to 70% to resolve the off-normal condition. Power ascension following the control rod conditioning was interrupted and a second downpower evolution was conducted to perform the repairs on the circulating water piping. Exelon determined that this downpower should not count towards the unplanned power changes performance indicator for the following reasons: The initial unplanned downpower was 15% and would not be subject to the performance indicator criteria of 20%. The downpower to 70% for the control rod conditioning was planned and was extended to conduct repairs to the A-north circulating water piping. The inspectors and Exelon could not resolve the differences in the interpretation of the NEI guidance and the issue will be submitted into the frequently asked question process for resolution. | ||
| 05000219/FIN-2013003-02 | 2013Q2 | Green | H.7 | NRC identified | Alarm Response Procedures did not implement Technical Specification Require | The inspectors identified a Green NCV of technical specification (TS) 6.8.1(a), Procedures and Programs, because Exelons alarm response procedure RAP-9XF2d, BUS C UV, (125 VDC system) was not adequately established and maintained. Specifically, the alarm response procedure allowed operator action that was not consistent with applicable TS 3.7.D, Station Batteries and Associated Battery Charger, requirements. Exelon entered this issue into the corrective action program for resolution as IR 1512551 and initiated a procedure change request. The inspectors determined this finding was more than minor because the finding affected the procedure quality attribute of the Mitigating System cornerstone objective to ensure the reliability and capability of systems that respond to initiating events. 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 that 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 technical specification allowed outage time, and did not screen as potentially risk-significant due to a seismic, flooding, or severe weather initiating event. Therefore, the inspectors determined the finding to be of very low safety significance (Green). This finding has a cross-cutting aspect in the area of Human Performance, Resources, because Exelon did not ensure that an alarm response procedure was adequate and accurately reflected technical specification requirements. |
| 05000219/FIN-2013003-01 | 2013Q2 | Green | P.1 | NRC identified | Degraded Emergency Diesel Generator Bypass Sight Glass not identified in the Corrective Action Program | The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, because Exelon did not promptly identify a condition adverse to quality. Specifically, from December 10, 2012 to April 4, 2013, Exelon did not identify that the fuel bypass sight glass on the #1 emergency diesel generator (EDG) was partially full. A partially full fuel bypass sight glass indicates that the bypass relief valve is degraded, challenging the operability of the emergency diesel generator because fuel could have bypassed the fuel injectors. Exelon entered this issue into the corrective action program for resolution as issue report (IR) 1497683 and subsequently replaced a degraded relief valve in the bypass sight glass. This finding is more than minor because it is associated with the design control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the reliability and capability of systems that respond to initiating events to prevent undesirable consequences. This issue was also similar to Example 3j of NRC IMC 0612, Appendix E, Examples of Minor Issues, because the condition resulted in reasonable doubt of the operability of the #1 emergency diesel generator and additional analysis was necessary to verify operability. The inspectors evaluated the finding using exhibit 2, Mitigating System Screening Questions in Appendix A to IMC 0609, Significance Determination Process. The inspectors determined that this finding was a deficiency affecting the design or qualification of a mitigating SSC, where the SSC maintained its operability or functionality. Therefore, inspectors determined the finding to be of very low safety significance (Green). The finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, because Exelon did not identify the issue associated with the degraded emergency diesel generator bypass sight glass in a timely manner. |
| 05000219/FIN-2013002-02 | 2013Q1 | Green | Licensee-identified | Licensee-Identified Violation | Technical specification 3.17.A, Control Room Heating, Ventilating, and Air Conditioning (HVAC) System , requires that The control room HVAC system shall be operable during all modes of operation. With one control room HVAC system determined inoperable, technical specification 3.17.B requires Exelon to verify once per 24 hours the partial recirculation mode of operation for the operable system, or place the operable system in the partial recirculation mode. On October 29, 2012, the B control room HVAC system experienced flow oscillations, was taken out of service and A control room HVAC system was placed in service. On November 1, 2012, Exelon operators initially declared the B control room HVAC inoperable due to the observed flow oscillations; however, Exelon operators, in error, subsequently determined that the B control room HVAC was not required to be operable in the cold shutdown mode of operation. Exelon discovered the error on November 9, 2012 during troubleshooting activities on the B control room HVAC system. Contrary to technical specification 3.17.B, Exelon did not verify or place the A control room HVAC system in the partial recirculation mode daily from October 29, 2012 until November 9, 2012. Exelon entered this issue into the corrective action program as IR 1438918. The inspectors determined that the finding was of very low safety significance (Green) in accordance with NRC IMC 0609, Appendix G, Shutdown Operations Significance Determination Process, Appendix 1, Checklist 7, because the finding did not affect core heat removal guidelines, did not affect inventory control guidelines, did not affect AC power guidelines, did not affect containment control guidelines, did not affect reactivity control guidelines and did not require a quantitative assessment. | |
| 05000219/FIN-2013002-01 | 2013Q1 | Green | P.1 | NRC identified | Emergency service water non-conformance not identified in the corrective action program | The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, when Exelon did not promptly identify or correct a condition adverse to quality. Specifically, Exelon did not identify and correct misaligned flanges between the D emergency service water pump and the associated discharge pipe during inspection activities in 2008 and subsequent maintenance activities in 2013. The misalignment of the flanges caused the expansion joint being installed in a configuration which exceeded design criteria of the emergency service water pump. Exelon entered this issue into the corrective action program for resolution as issue report (IR)1481670. This finding is more than minor because it is associated with the design control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. This issue was also similar to Example 3j of NRC IMC 0612, Appendix E, Examples of Minor Issues, because the condition resulted in reasonable doubt of the operability of emergency service water system and additional analysis was necessary to verify operability. The inspectors evaluated the finding using IMC 0609, Appendix A, and determined this finding was a deficiency affecting the design or qualification of a mitigating SSC, where the SSC maintained its operability or functionality. Therefore, inspectors determined the finding to be of very low safety significance (Green). This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, because Exelon did not identify the issue associated with the non-conforming emergency service water expansion joint in a timely manner. |
| 05000219/FIN-2012005-02 | 2012Q4 | Green | H.2 | Self-revealing | Inadequate Application of Strippable Coating to the Refueling Cavity Liner and the Failure to Configure a Valve in the Leakage Collection System Resulting in Increased Potential for Corrosion on the Exterior of the Drywell Liner Surface in the Sand Beds | A self-revealing NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified because Exelon procedures and work orders were not effective in preventing refueling cavity leakage from overflowing onto the exterior surface of the drywell liner during the refueling outage (1R24) in November 2012. The performance deficiencies that contributed to the finding were inadequate oversight of the contractors applying a strippable coating to the reactor cavity liner and a valve configuration control error on a temporarily installed leakage collection system. Upon discovery, Exelon took immediate corrective actions to open the leakage collection system filter inlet valve and restore reactor cavity liner leakage flow to the reactor building equipment drain tank. This finding is associated with the barrier integrity cornerstone and is more than minor because, if left uncorrected, this condition would have the potential to lead to a more significant safety concern. Specifically, the continued wetting of the metallic drywell liner surface could provide an environment conducive to corrosion. This finding is not more than very low safety significance because Exelon performs periodic inspections of drywell liner and exterior surface coating to ensure that liner corrosion is monitored and controlled. The inspectors completed the Phase 1 Initial Screening and Characterization of Findings, of Attachment 0609.04 of Inspection Manual Chapter (IMC) 0609, and screened the finding to Green, very low safety significance. Exelon has entered this condition into the corrective action process under IR 1440116. This finding has a crosscutting aspect in the area of Human Performance, Work Practices, for not ensuring supervisory and management oversight of work activities, including contractors and plant personnel, such that nuclear safety is supported regarding the application of the strippable coating on the reactor cavity liner. |