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| Q | Significance | CCA | Identified by | Title | Description | |
|---|---|---|---|---|---|---|
| 05000277/FIN-2018003-03 | 2018Q3 | White | NRC identified | Reactor Core Isolation Cooling System Pressure Switch Failure Results in Condition Prohibited by TS - EA-18-108 | On April 22, 2018, during a routine surveillance test of the RCIC system, the RCIC turbine tripped approximately 28 seconds after startup, prior to the system reaching rated flow and pressure. Concurrent with the RCIC trip, an alarm was received for RCIC turbine high exhaust pressure; however, local indications did not indicate a true high pressure in the exhaust line. Therefore, the RCIC system was declared inoperable and TS 3.5.3, Condition A was entered, which requires the RCIC system to be restored to operable within 14 days. Troubleshooting determined that the B RCIC exhaust pressure switch (PS-3-13-72b) had prematurely tripped at normal operating pressure due to an age-related failure of the instrument diaphragm and O-ring. The RCIC system had been previously verified as operable during its last surveillance run on January 16, 2018. Corrective Actions: The failed pressure switch was replaced and the station performed an extent of condition review/inspection of similar pressure switch instruments. Following replacement of the switch, RCIC was retested and restored to operable on April 23, 2018. Furthermore, actions were established to modify the turbine trip logic to remove the single point trip vulnerability. Corrective Action Reference: IR 4129583 Enforcement:Violation: Peach Bottom Unit 3 TS 3.5.3 requires that the RCIC system shall be operable in Mode 1, and if RCIC becomes inoperable, it shall be returned to operable status within 14 days or the plant shall be placed in Mode 3 within the next 12 hours. Contrary to the above, based on relevant causal information, Unit 3 RCIC was likely inoperable prior to April 22, 2018, for a period greater than the TS allowed outage time of 14 days, and Unit 3 had not been placed in Mode 3. Specifically, on April 22, 2018, the Unit 3 RCIC turbine tripped during startup for a routine surveillance test due to a degraded turbine exhaust pressure switch which resulted in an inoperability time of greater than 14days. Internal inspection on the switch identified that it failed due to corrosion from water intrusion which had existed for an extended period of time. Severity/Significance: For violations warranting enforcement discretion, IMC 0612 does not require a detailed risk evaluation; however, safety significance characterization is appropriate. A Region I SRA performed a best estimate analysis of the safety significance using the Peach Bottom Unit 3 Standardized Plant Analysis Risk (SPAR) model, Version 8.51 and Systems Analysis Programs for Hands-On Integrated Reliability Evaluations (SAPHIRE), Version 8.1.8. This model was used to evaluate the internal events increase in core damage frequency (CDF) per year. The SRA performed a site visit to review Exelons fire model output to estimate the external risk contributor of the issue. The final risk evaluation estimated the total (internal and external events risk) increase in CDF to be in the mid E-6/yr range, or of low to moderate safety significance. The SRA evaluated the internal and external events risk contribution due to the inoperability of the RCIC system for an assumed 47 day exposure time. 16 The analyst used the guidance in the Risk Assessment Standardization Project (RASP) Handbook, Volume I, Section 2.4, Revision 2.0, to estimate an exposure time using a time divided by two (t/2) approach. This would represent the time from the last successful surveillance test divided by two. The approach is appropriate for periodically operated components that fail due to a degradation mechanism that gradually could affect the component during the standby period. Given this approach, the internal event contribution was calculated to estimate the internal event risk increase due to the conditional failure of the RCIC pump to successfully start. The increase for internal events was estimated at 2.5E-6/yr increase in CDF. The dominant sequence involved a loss of condenser heat sink, with operator action failure to depressurize, and HPCI system failures. The SRA noted from discussions with Exelon staff that the RCIC system was assumed to be non-recoverable given the nature of the failure. To estimate the external risk contribution, the SRA had several discussions and a site visit to review Exelons preliminary fire model outputs for the conditional failure of the RCIC system for the 47 days. The 47 days included a conservative additional day for repair time. The SRA reviewed Exelons fire risk analysis and noted that one of the dominant risk increase contributors was fire within the 13kV switchgear room. Several other fire areas were reviewed and the SRA noted that the core damage sequences appeared technically reasonable given the plant areas and values assumed for mitigating equipment. Exelons preliminary results showed an increase in external event CDF/yr for the conditional failure of RCIC for 47 days to be approximately 4.5E-6/yr. The SRA determined the results to be reasonable. Exelons model for internal events resulted in an increase in CDF/yr of 1.05E-6/yr which was considered to compare well with the NRC SPAR model. Exelon performed a review of the large early release frequency (LERF) impact and determined an overall increase in LERF due to both external and internal events for the RCIC failure for 47 days to be a nominal 6E-8/yr. Therefore, the SRA review of the dominant sequences and Exelons LERF results affirmed that LERF did not increase the risk over that determined from the increase in CDF. Basis for Discretion: The inspectors determined that the maintenance strategy for these switches was consistent with requirements and standards that existed at the time and that there was no relevant operating experience that would have reasonably necessitated consideration of additional maintenance actions. As a result, no performance deficiency was identified. The inspectors assessment considered: The industry, regulatory, and Exelon service life standards were reviewed for static O-ring pressure switches. Exelons assessment of the pressure switch service condition (critical, mild conditions, low-duty cycle) required a preventive maintenance task to perform periodic calibration and to replace the switch as-required. There was no time-based replacement task prescribed by any standard for this switch. The inspectors determined that Exelons assessment was adequate and the corresponding preventive maintenance activities met applicable standards. The subject pressure switch was installed during original construction and the calibration results of the pressure switch had been satisfactory from 2003 until the 2018 failure. The inspectors reviewed the maintenance and calibration history on the pressure switch and did not identify any adverse trends or conditions adverse to 17 quality that would have required further evaluation or replacement of the pressure switch. Industry operating experience information available to Exelon did not identify the potential for the age-related failure mode of the pressure switch o-ring and diaphragm that occurred at Peach Bottom. 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 TSs (EA-18-108). Further, because Exelons actions did not contribute to this violation, it will not be considered in the assessment process or the NRC Action Matrix | |
| 05000277/FIN-2018410-01 | 2018Q3 | Green | H.8 | NRC identified | Security | |
| 05000277/FIN-2018003-01 | 2018Q3 | Green | P.2 | NRC identified | HPCI System Exhaust Pressure Switches Exceeded Documented Qualified Life | The inspectors identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, because Exelon did not establish measures to ensure that environmental qualification requirements for qualified components were correctly translated into procedures and instructions. Specifically, the end-of-life replacement requirements for the Unit 2 HPCI exhaust pressure switches were not translated into maintenance procedures and instructions. As such, Exelon did not replace the switches prior to the end of their documented qualified life. |
| 05000277/FIN-2018003-02 | 2018Q3 | White | H.11 | NRC identified | Inadequate Corrective Actions Result in the Failure of the E-3 EDG | The inspectors identified a self-revealing preliminary White finding associated with an apparent violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, because Exelon did not perform adequate corrective actions on the E-3 EDG scavenging air check valve assembly. Specifically, Exelon did not perform an adequate repair of an interference fit pin joint during maintenance activities in April 2017 and did not correct an oil leak on the check valve dashpot assembly identified in September 2017, which resulted in the E-3 EDG failure on June 13, 2018. |
| 05000277/FIN-2018010-03 | 2018Q2 | Green | Licensee-identified | Licensee-Identified Violation | This violation of very low safety significance was identified by Exelon and has been entered into Exelons corrective action program and is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy. Violation: Peach Bottom Unit 2 and Unit 3 Renewed Facility Operating License Condition 2.C.(4) requires, in part, Exelon to implement and maintain in effect all provisions of the approved fire protection program as described in the Updated Final Safety Analysis Report. Fire Protection Program, Peach Bottom Atomic Power Station, Units 2 and 3, is incorporated by reference into the Updated Final Safety Analysis Report, as discussed in Section 10.12, Fire Protection Program. Fire Protection Program Chapter 5.1, Methodology, assumes that two or more circuit failures resulting in spurious operation of two or more valves in series at a high/low pressure interface may occur due to a postulated fire in any given area.Fire Protection Program Chapter 6.2, Analysis of High/Low Pressure Interfaces, requires Exelon to address the situations for which the isolation valves at a given interface point consists of two electrically controlled valves in series and where the potential may exist for a single fire to cause damage to cables associated with both valves. 8 Contrary to above, as ofMarch 14, 2018, Exelon identified they failed to evaluate two motor-operated valves in series, MO-2-06-2663 and MO-2-06-038B for Unit 2, and MO-3-06-3663 and MO-3-06-038B for Unit 3, where the potential may exist for a single fire to cause damage to cables associated with both valves. Specifically, a postulated fire scenario could cause spurious opening of the valves, which may potentially result in a fire-induced loss of coolant accident through the high/low pressure system interface. Exelons evaluation identified the affected valves cables are routed through Fire Area 6N for the Unit 2 valves, and Fire Area 13N for the Unit 3 valves, and thus, a possibility exists for a single fire to cause damage to cables associated with both valves. This violation is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy. Significance/Severity: The inspectors performed a Phase 2 Significance Determination Process screening for this issue, in accordance with IMC 0609, Appendix F, Fire Protection Significance Determination Process. This finding affected the post-fire safe shutdown category because of spurious operations of safe shutdown components. Based on a walkdown of Fire Areas 6N and 13N, the inspectors did not identify any credible fire ignition source scenarios that could affect both Unit 2 motor-operated valves or both Unit 3 motor operated valves. Therefore, based upon task number 2.3.5, the inspectors determined that this finding screened to very low safety significance (Green).Corrective Action References: IR 04115309, EC 623585, EC 623586 | |
| 05000277/FIN-2018010-02 | 2018Q2 | Green | NRC identified | Failure to Develop and Maintain Mitigating Strategy | The inspectors identified a Green non-cited violation of 10 CFR 50.54(hh)(2), Conditions of Licenses, and Peach Bottom Unit 2 and Unit 3 Renewed Facility Operating License Condition 2.C.(11), Mitigation Strategy License Condition, because Exelon did not develop and maintain strategies for addressing large fires and explosions that include operations to mitigate fuel damage. Specifically, Exelon did not adequately develop and maintain procedures to manually depressurize the reactor using the automatic depressurization system safety relief valves in the event of a challenge to the reactor due to a postulated large fire and/or explosion. | |
| 05000277/FIN-2018010-01 | 2018Q2 | Green | P.2 | NRC identified | Failure to Identify Time-Critical Action | The inspectors identified a green finding because Exelon did not identify, validate, and incorporate a time-critical action referenced in calculation PF-0016-025, Fire Area 025 Fire Safe Shutdown Analysis, in accordance with Sections 4.2 and 4.3 of OP-AA-102-106, Operator Response Time Program. Specifically, Exelon did not identify a 10-minute time-critical action to take the transfer/isolation switch for the high pressure coolant injection (HPCI) inboard steam isolation valve (MO-2(3)-23-015), to the EMERG position |
| 05000277/FIN-2018002-01 | 2018Q2 | Green | P.1 | NRC identified | Failure to Identify and Promptly Correct a Condition Adverse to Quality Concerning Battery Charger 2B-003-1 | The NRC identified a Green non-cited violation (NCV) of 10 Code of Federal Regulations(CFR) Part 50, Appendix B, Criterion XVI, Corrective Action, because Exelon did not identify and promptly correct a condition adverse to quality (CAQ) commensurate with its safety significance concerning the 2BD-003-1 safety-related battery charger. Specifically, Exelon did not appropriately prioritize repairs for a CAQ and, as a result, the 2BD-003-1 battery charger failed to operate when placed in service on June 5, 2018 |
| 05000278/FIN-2018001-01 | 2018Q1 | Green | Self-revealing | Untimely Corrective Actions to Address Primary Containment Isolation Valve Condition Adverse to Quality | A Green self-revealing non-cited violation(NCV)of 10 Code of Federal Regulations(CFR)50, Appendix B, Criterion XVI, Corrective Action, was identified because Exelon did not implement prompt corrective actions to address a condition adverse to quality (CAQ) on primary containment isolation valve (PCIV) SV-3-7D-3671G.Specifically, drywellair sampling valve SV-3-7D-3671G failed to perform its PCIV function on February 1, 2018, by failing to stroke closed during its surveillance test as a result of untimely corrective actions.Exelon isolated the associated piping in accordance with technical specifications(TSs) | |
| 05000277/FIN-2017403-01 | 2017Q3 | Green | H.11 | NRC identified | Security | |
| 05000278/FIN-2017003-01 | 2017Q3 | Green | Self-revealing | Instructions Not Followed for Replacement of HPSW Ventilation Switch Block | A self-revealing NCV of Technical Specification (TS) 5.4.1, Procedures,of very low safety significance (Green) was identified for Exelonnot implementing procedural instructions for the replacement of the HS-3-40H-3AV060 switch block associated with the 3AV060 high pressure service water (HPSW) ventilation fan. Exelon did not ensure that electrical connections were free of loose wire strands per their procedural standard E-1317,Wire and Cable Notes and Details, Power, Control, and Instrumentation, Revision 55, and from the vendor manual instructions. As a result,on July 10, 2017, the 3AV060 HPSW ventilation fan failed its surveillance test(ST)and rendered one subsystem of Unit 3 HPSW inoperable. Exelon entered this issue into their corrective action program (CAP) asissue reports(IR)4030367 and 4044444, straightened out the remaining loose strands, and specified additional electrical panels for an extent of condition (EOC) review.Thisfinding ismore than minor because it isassociated with the equipment performance attribute of the Mitigating Systemscornerstoneand affected the cornerstones objective to ensure the reliability, availability, and capability of systems to respond to initiating events to prevent undesirable consequences (i.e. core damage).By not implementing theE-1317 procedural instructions, the 3AV060 fan failed and affected the reliability of one HPSW subsystem.The inspectors evaluated the finding in accordance with Exhibit 2 of IMC 0609, Appendix A, SDP for Findings At-Power and determined the finding was of very low safety significance (Green) because it did notrepresent a loss of system function or represent an actual loss of function of at least a single train for longer than itsTSallowed outage time. The inspectors determined no cross-cutting aspect applied because the PD occurred in 2010 and was not indicative of current performance. | |
| 05000277/FIN-2017403-02 | 2017Q3 | Green | H.12 | NRC identified | Security | |
| 05000277/FIN-2017003-02 | 2017Q3 | Severity level IV | Licensee-identified | Licensee-Identified Violation | 10 CFR 55.25 states, in part, that if an operator develops a permanent physical or mental condition that causes the operator to fail to meet the requirements of 10 CFR 55.21, the facility licensee shall notify the Commission within 30 days of learning of the diagnosis, in accordance with 10 CFR 50.74(c),which states,that the regional administrator shall be notified if a licensed operator develops a permanent disability or illness. Contrary to these requirements, as the result of Exelons medical examination audit completed September 26, 2017, Exelon identified a change in a licensed operators medical condition that was not communicated to the NRC within the required 30 days. The results of the medical examination audit were documented in IR 4054146 and subsequent notifications were made to the NRC.This violation is subject to traditional enforcement because of the potential impact upon the regulatory process for issuing restrictions to operators licenses. The inspectors determined that this issue meets the criteria for a Severity Level IV violation using example 6.4.d.1(a) from the NRC Enforcement Policy because no incorrect regulatory decision was made as the result of the failure of the licensee to report within 30 days. This is of very low safety significance because after NRC review of the subsequent notifications, no changes to license restrictions were required. | |
| 05000277/FIN-2017201-01 | 2017Q2 | Green | NRC identified | Security | ||
| 05000277/FIN-2017201-02 | 2017Q2 | Green | NRC identified | Security | ||
| 05000278/FIN-2017002-01 | 2017Q2 | Green | H.8 | Self-revealing | Corrective Action Not Implemented Correctly for Replacement of RCIC RCR Contacts | A self-revealing non-cited violation (NCV) of 10 Code of Federal Regulation(CFR)Part 50, Appendix B, Criterion XVI, Corrective Actions, of very low safety significance (Green) was identified for Exelon not correcting a condition adverse to quality concerning reverse control relay (RCR) contacts for valves associated with the reactor core isolation cooling (RCIC) system. Specifically, Exelon specified a corrective action (CA) from an October 18, 2013, Unit 3 RCIC equipment apparent cause evaluation (EACE) to replace RCR contacts after 12 years of service, however, the CA was not correctly implemented. As a result, on January 12, 2017, an RCR contact associated with the Unit 3 RCIC suppression pool suction valve remained in service for 15 years, exhibited a high resistance failure during a surveillance which resulted in Unit 3 RCIC being inoperable. Following the failure, Exelon initiated issue reports (IRs) 03962563 and 03977949, implemented corrective actions to replace the RCR contact, restored Unit 3 RCIC operability, and risk-informed their corrective maintenance schedule for replacing all RCR contacts that currently exceeded the recommended 12-year service life.Exelons failure to recognize and correct a condition adverse to quality associated with certain RCR contacts in their Unit 3 RCIC system that had exceeded their 12-year service life, was a performance deficiency (PD) that was within their ability to foresee and correct and should have been prevented. The finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone and affected the cornerstones objective to ensure the reliability of systems to respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, not recognizing that existing RCR contacts were installed in safety-related equipment beyond their 12-year service life, resulted in the failure of the Unit 3 RCIC suppression pool suction valve. The inspectors evaluated the finding in accordance with Exhibit 2 of IMC 0609, Appendix A, SDP for Findings At-Power, and determined the finding was of very low safety significance (Green) because it did not represent a loss of system function or represent an actual loss of function of at least a single train for longer than its technical specification (TS) allowed outage time of 14 days. The inspectors determined that the finding has a cross-cutting aspect in Human Performance, Procedure Adherence, because Exelon did not validate that the correct revision of procedure WC-AA-120, Attachment 2, Preventive Maintenance (PM) Change Review Form, was used when creating a new PM to replace RCR contacts. (H.8) |
| 05000277/FIN-2017002-02 | 2017Q2 | Severity level Enforcement Discretion | NRC identified | EDG Exhaust Stacks Nonconforming Design for Tornado Missile Protection | On January 9, 2017, it was determined that PB's EDGs do not conform with the licensing basis for protection against tornado-generated missiles. The exhaust stacks for the four on-site EDGs extend approximately seven feet above the roof of the EDG building. In the event of a tornado, debris generated from the tornado could strike the exhaust stacks and, if at a sufficient mass and velocity, could crimp the exhaust stacks in a manner that would affect EDG operation.As a result of the non-conforming condition, on January 9, 2017, at 1530, all four EDGswere declared inoperable. Compensatory measures were put in place and, in accordance with NRC guidance contained in Enforcement Guidance Memorandum (EGM) 15-002, the EDGs were returned to an operable but non-conforming status.There are no actual consequences as a result of the non-conforming condition. This LER is closed.b. FindingsDescription. 10 CFR 50, Appendix B, Criterion III, Design Control, requires, in part, that measures shall be established to assure that the applicable regulatory requirements and the design basis for SSCs are correctly translated into specifications, drawing, procedures, and instructions. Contrary to the above, Exelon failed to correctly translate the design basis for protection against tornado-generated missiles into their specifications and procedures. Specifically, Exelon did not adequately protect Unit 2 and Unit 3s EDG exhaust stacks from tornado-generated missiles.Exelon documented the condition adverse to quality in their CAP under IR 3961028 and took immediate compensatory actions. The inspectors evaluated Exelons immediate compensatory measures, which included verifying that procedures are in place, equipment was appropriately staged, and training is current for performing actions in response to a tornado to preserve EDG operability. Enforcement. Because this violation was identified during the discretion period covered by EGM 15-002, Revision 1, Enforcement Discretion for Tornado Generated Missile Protection Non-Compliance, (ML16355A286) and because Exelon has implemented compensatory measures, the NRC is exercising enforcement discretion, is not issuing enforcement action, and is allowing continued reactor operation. | |
| 05000277/FIN-2017008-01 | 2017Q1 | Green | P.3 | NRC identified | Untimely Corrective Actions to Address 2C Core Spray Motor Elevated Vibrations | Green. The inspectors identified a non-cited violation of 10 CFR 50 Appendix B, Criterion XVI, Corrective Action, because Exelon did not implement corrective actions in a timely manner to correct a condition adverse to quality on the 2C core spray motor. Specifically, Exelon did not perform appropriate corrective actions to evaluate and address an increasing motor bearing vibration trend that had existed for over ten years. Consequently, motor vibration reached the fault level established in Exelons vibration analysis procedure. The finding was more than minor, because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone and adversely impacted the cornerstone objective to ensure the reliability 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, dated October 7, 2016, and Exhibit 1 of IMC 0609, Appendix A, The SDP for Findings At-Power, dated June 19, 2012, the inspectors determined this finding was of very low safety significance because the performance deficiency did not impact the design or qualification of the component, did not result in a loss of system function, did not result in the loss of function of a train greater than its Tech Spec allowed outage time, and did not represent an actual loss of function for a high safety significant component in accordance with Exelons maintenance rule program. The inspectors determined the finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Resolution, because Exelon did not take effective corrective actions in a timely manner commensurate with the safety significance of the issue. Specifically, corrective actions to address the elevated vibrations on the 2C core spray motor were not implemented before motor vibration reached the fault level and adversely impacted the long-term reliability of the motor. (P.3) |
| 05000278/FIN-2017008-02 | 2017Q1 | Green | P.3 | Self-revealing | Untimely Corrective Actions to Address Elevated Primary Containment Isolation Valve Leakage | Green. The inspectors identified a self-revealing non-cited violation of 10 CFR 50 Appendix B, Criterion XVI, Corrective Action, because Exelon did not promptly implement corrective actions to address a condition adverse to quality on two containment isolation valves. Specifically, drywell air sampling valves SV-3-7D-3671A and SV-3-7D-3671D failed to perform their primary containment isolation function on March 15 and September 26, 2016, respectively, as a result of untimely corrective actions to address elevated leakage. The valve internals were repaired, declared operable, and the issue was entered into the corrective action program (IR 3990490). The finding was more than minor, because it was associated with the barrier performance attribute of the Barrier Integrity cornerstone and adversely affected the cornerstones objective to provide reasonable assurance that the containment design barrier protect the public from radionuclide releases caused by accidents or events. In accordance with IMC 0609.04, Initial Characterization of Findings, dated October 7, 2016, and Exhibit 1 of IMC 0609, Appendix A, The SDP for Findings At-Power, dated June 19, 2012, the inspectors determined this finding was of very low safety significance, because the finding did not result in an actual open pathway in the physical integrity of the reactor containment or involve an actual reduction in the function of hydrogen igniters in the reactor containment. The inspectors determined this finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Resolution, because Exelon did not perform effective corrective actions in a timely manner commensurate with the safety significance of the issue. Specifically, corrective actions to address a CAQ on SV-3-7D-3671A and SV-3-7D-3671D were delayed which resulted in the valves failing their LLRT and being declared inoperable. (P.3) |
| 05000277/FIN-2016004-01 | 2016Q4 | Green | NRC identified | Failure to Identify and Remove FM in CAD System Piping | Green. The inspectors identified a finding of very low safety significance (Green) involving a non-cited violation (NCV) of 10 CFR 50 Appendix B Criterion XVI, Corrective Action, because Exelon did not adequately identify and correct a condition adverse to quality associated with the containment atmospheric dilution (CAD) piping system. Specifically, in 2012, Exelon did not adequately identify the source of foreign material (FM) and implement corrective actions to remove the FM from the CAD piping which resulted in the failure of the CHK-2-07C-40145 containment isolation valve to close in 2016. Exelon documented the issue in issue report (IR) 2735344 and promptly replaced the valve and restored the valve to operable. As an interim corrective action, Exelon plans to increase the local leak-rate test (LLRT) frequency and replacement of the check valve to maintain reasonable assurance of operability. Exelon is implementing a detailed troubleshooting plan to identify the source of FM and perform corrective actions to address the condition adverse to quality. The performance deficiency (PD) is more than minor because it was associated with the containment barrier performance attribute of the barrier integrity cornerstone and it adversely impacted the cornerstone objective to provide reasonable assurance that physical design barriers (containment) protect the public from radionuclide releases caused by accidents or events. The inspectors evaluated the finding using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The SDP for Findings at-Power, Exhibit 3, and the inspectors determined this finding to be of very low safety significance (Green) because the degraded condition did not represent an actual open pathway in the physical integrity of containment, and did not involve an actual reduction in function of hydrogen igniters in the reactor containment. The inspectors determined that a cross cutting aspect does not apply because the performance deficiency occurred greater than three years ago and is not indicative of current plant performance. | |
| 05000277/FIN-2016003-01 | 2016Q3 | Green | Self-revealing | Reactor Feed Pump Controller Power Supply Shelf Life Not Maintained | A self-revealing finding of very low safety significance (Green) was identified for Exelons failure to maintain the Unit 2 C reactor feed pump (RFP) Woodward controller secondary power supply in accordance with PES-S-002, Exelon Shelf Life Program. Specifically, on May 27, 2016, the Unit 2 C RFP experienced speed oscillations due to an age-related failure of the Woodward controller secondary power supply, which resulted in an automatic recirculation runback to 53 percent rated thermal power (RTP). The power supply contained an electrolytic capacitor that had exceeded its shelf life per PES-S-002. This issue was entered into Exelons corrective action program (CAP) under issue report (IR) 02691322. Exelons corrective actions included replacement of the faulted power supply and an extent of condition (EOC) review of proper expiration date entry for shelf life program components. The finding was more than minor because it was associated with the equipment performance attribute of the Initiating Events cornerstone and affected the cornerstones objective of limiting the likelihood of events that upset plant stability during power operations. The inspectors evaluated the finding in accordance with Exhibit 1 of Inspection Manual Chapter (IMC) 0609, Appendix A, SDP for Findings At-Power, and determined the finding was 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. The inspectors determined that no cross-cutting aspect was applicable to this finding because the performance deficiency (PD) was not indicative of current performance. The PD occurred between 1997 and 1999 when the power supply expiration date was incorrectly coded in Exelons work management process in accordance with PES-S-002. | |
| 05000277/FIN-2016404-01 | 2016Q3 | Green | H.14 | NRC identified | Security | |
| 05000277/FIN-2016002-03 | 2016Q2 | Green | H.2 | Self-revealing | Human Performance Event Results in Emergent Downpower | A self-revealing finding of very low safety significance (Green) was identified for the failure of Exelon operators to use human performance error reduction tools during equipment manipulation in accordance with HU-AA-101, Human Performance Tools and Verification Practices. Specifically, on March 28, 2016, an equipment operator failed to use self-check (STAR) while removing a circuit breaker from service and incorrectly tripped the E-124 480 volt supply breaker which required a rapid manual power reduction to 80 percent rated thermal power (RTP) due to lowering main condenser vacuum and a partial loss of feedwater heating. Exelon entered the issue into their corrective action program (CAP) under issue report (IR) 2646772 and performed a root cause which identified corrective actions to address the adverse human performance behaviors at the station. The finding was more than minor because it was 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, an equipment operator failed to adequately use human performance error reduction tools and opened an incorrect breaker which required a rapid downpower. The inspectors evaluated the finding in accordance with Exhibit 1 of IMC 0609, Appendix A, The SDP for Findings At-Power, dated June 19, 2012, and determined the finding was of very low safety significance (Green) because it did not result in a reactor trip and the loss of mitigation equipment relied upon for transition to a stable shutdown condition. This finding was determined to have a cross-cutting aspect in the area of Human Performance, Field Presence, because Exelon did not ensure that deviations from standards and expectations, which were identified by leaders, were corrected promptly. Specifically, Exelon identified that adverse human performance behaviors existed with certain equipment operators, however, those observations were not appropriately input into their performance management system, such that the behaviors could be addressed. Thus, these adverse behaviors were a primary contributor to this human performance error. |
| 05000277/FIN-2016002-01 | 2016Q2 | Green | H.8 | NRC identified | Improperly Stored Material in Reactor Building | The NRC identified a very low safety significance (Green) NCV of Technical Specification (TS) 5.4.1 for Exelons failure to adequately implement procedure requirements governing the storage of material in a safety-related structure. Specifically, on April 26, 2016, Exelon technicians stored ladders vertically without them being adequately tied off to prevent the ladders from falling over in accordance with MA-AA-716-026, Station Housekeeping / Material Condition Program. The inspectors identified that the ladders were stored in the PB Unit 2 reactor building (RB), such that they could fall over and impact safety-related equipment. The inspectors promptly notified Exelon, the ladders were immediately removed, and the condition was documented under IR 2661309. This finding was more than minor because it was associated with the protection against external factors attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the reliability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated the finding using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The SDP for Findings At-Power, Exhibit 2. The inspectors determined this finding to be of very low safety significance (Green) because the degraded condition was not a design deficiency that affected system operability; did not represent an actual loss of function of a system; did not represent an actual loss of function of a single train or two separate trains for greater than its TS allowed outage time; and did not represent an actual loss of function of one or more non-TS trains of equipment designated as high safety significant. The finding was determined to have a cross-cutting aspect in the area of Human Performance, Procedure Adherence, because Exelon technicians did not store ladders in safety-related buildings in accordance with station procedures, such that they could not fall over and damage safety-related equipment. |
| 05000277/FIN-2016002-02 | 2016Q2 | Green | H.1 | NRC identified | Untimely Corrective Actions to Address Condition Adverse to the Fire Protection Program Alternative Shutdown Capability | The inspectors identified an NCV of very low safety significance (Green) of PB Unit 2 and Unit 3 Facility Operating License condition 2.C.(4) for failure to implement and maintain in effect all provisions of the approved fire protection program. Exelon did not correct a condition adverse to the fire protection program alternative shutdown capability in a timely manner. Specifically, Exelon did not establish testing requirements for transfer/isolation switches since the identification of the issue on February 6, 2014, and the due date to complete this action was extended to February 24, 2018. As a result, Exelon has delayed assurance that the components credited for alternative shutdown capability would perform their fire protection design basis function. Exelon entered this issue into their CAP as IR 02669323. This performance deficiency (PD) was more than minor because it was associated with the protection against external factors (fire) 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, by failing to correct the condition, Exelon has not ensured that the control circuit for the safe shutdown components would be isolated from the effects of fire damage. The inspectors determined that the finding was of very low safety significance (Green) based on IMC 0609, Appendix F, Fire Protection SDP, task number 1.3.1, because Exelon had demonstrated reasonable expectation of functionality for these switches by having comparable switches in the test program and periodically testing those switches. The test results did not indicate any kind of significant failures of these switches. This finding was determined to have a cross-cutting aspect in the area of Human Performance, Resources, in that, Exelon extended the due date to complete the corrective action to support the completion of higher priority items, indicating lack of resources. |
| 05000277/FIN-2016403-01 | 2016Q2 | Severity level Enforcement Discretion | NRC identified | Security | ||
| 05000278/FIN-2016001-01 | 2016Q1 | Severity level IV | Licensee-identified | Licensee-Identified Violation | On September 29, 2015, Exelon identified the door to the Unit 3 condensate backwash tank room was not secure. The room is controlled as a locked HRA, and a survey of the room indicated that actual radiation levels were greater than 1.0 rem/hour. TS 5.7.2.a requires, in part, that entryways to areas exceeding 1.0 rem/hour will be locked or continuously guarded to prevent unauthorized entry. Contrary to the above, on September 29, 2015, Exelon identified an area with radiation levels greater than 1.0 rem/hour with an entryway that was not locked or continuously guarded. Traditional enforcement applies in accordance with Inspection Manual Chapter (IMC) 0612, sections 0612-09 and 0612-13; and Enforcement Policy Section 2.2.4.d; because the inspectors did not identify an associated performance deficiency. Specifically, the inspectors determined that because Exelon had an acceptable door maintenance program, conducted weekly checks of LHRA doors, and has not had previous issues with unsecured doors, that the failure of the door lock mechanism was not apparent and, therefore, was not foreseeable and preventable. The issue was considered to be a SL IV violation of TS 5.7.2.a in accordance with Enforcement Policy Section 6.1.d. In addition, IMC 0612, Appendix B, Figures 1 and 2, Issue Screening, were utilized in documenting this as a SL IV licensee-identified NCV. The licensee took immediate corrective actions to ensure the door remained locked and documented the issue in condition report 2562192, and the investigation determined that no unauthorized access to the room had occurred. | |
| 05000277/FIN-2015004-01 | 2015Q4 | Green | NRC identified | Failure to Ensure Design Basis of Emergency Diesel Generator Lubrication System | The inspectors identified a non-cited violation (NCV) of very low safety significance of 10 Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion III, Design Control, for not ensuring that the adequacy of PBAPS emergency diesel generator (EDG) lubrication oil (LO) supply was designed to withstand the effects of natural phenomena. Specifically, additional LO, evaluated by PBAPS to meet their EDG technical specification (TS) mission time of seven days of continuous operation, was housed in a non-class I structure that would be unable to withstand the effects of natural phenomena. PBAPS entered the issue into the correction action program (CAP) as issue report (IR) 02603369 and took immediate corrective actions to relocate the LO reserve inventory from their warehouse to the 135 elevation of the PBAPS radwaste building, which is a seismic class I structure The finding is considered more than minor because it is associated with the Protection Against External Factors attribute of the Reactor Safety Mitigating Systems cornerstone and adversely affected the cornerstones objective of ensuring reliability and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). The inspectors evaluated the significance of this finding using IMC 0609 Appendix A, The SDP for Findings at Power, Exhibit 2, Mitigating Systems Screening Questions. The inspectors determined that this finding was of very low safety significance (Green) because the finding is a design deficiency which did not result in an actual loss of functionality of the EDGs. This finding did not have a cross-cutting aspect because the most significant contributor of the performance deficiency (PD) occurred during the 1994 conversion to improved technical specifications (ITS) and, thus, was not reflective of current plant performance. Specifically, PBAPS current engineering change request (ECR) process would evaluate for natural phenomena considerations such as seismic, tornado, flood, etc. | |
| 05000277/FIN-2015003-01 | 2015Q3 | Green | H.12 | NRC identified | Incomplete Testing of Components from the Remote Shutdown Panels | The inspectors identified a Green NCV of Technical Specification (TS) 5.4.1.a after Exelon did not establish and implement procedures to adequately test the Unit 2 and Unit 3 remote shutdown panels (RSPs). Specifically, Exelons surveillance procedure did not test all the control circuits, as required by Surveillance Requirement (SR) 3.3.3.2.1, for the Unit 2 and Unit 3 RSPs. Exelons corrective actions included entering this issue into their CAP, the development of RSP testing procedures for the reactor core isolation cooling (RCIC), control rod drive (CRD), and emergency service water (ESW) system components, and a revision to the bases for TS 3.3.3.2 The performance deficiency (PD) was determined to be more-than-minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Additionally, examples 1.c, 4.l, and 4.m from IMC 0612, Appendix E, detail that a PD was more than minor if required TS surveillance testing is not performed and subsequent testing reveals that the equipment is out of specification or otherwise unable to perform a safety-related function. A detailed risk evaluation concluded that the issue was of very low safety significance (Green). This finding had a cross-cutting aspect in Human Performance, Avoid Complacency, because Exelon failed to recognize and plan for the possibility of latent problems. |
| 05000277/FIN-2015003-02 | 2015Q3 | Green | Licensee-identified | Licensee-Identified Violation | The following violation of very low safety significance (Green) was identified by Exelon and is a violation of NRC requirements which meets the criteria of the NRC Enforcement Policy for being dispositioned as a NCV. From 2010 to 2014, PBAPS made a total of 18 shipments of radioactive waste for disposal to the Energy Solutions Clive, UT facility, which contained category 2 levels of radioactive material quantity of concern (RAM-QC), but did not implement transportation security plan for these shipments, which is contrary to the requirements of 10 CFR 71.5 and 49 CFR 172, Subpart I, Safety and Security Plans. This PD adversely affected the Public Radiation Safety cornerstone attribute of Program and Process based on inadequate procedures associated with the transportation of radioactive materials. This issue was documented in Exelons CAP as assignment reports 02484424, 02487034, and 02490534. | |
| 05000277/FIN-2015008-01 | 2015Q2 | Green | P.1 | NRC identified | Failure to Initiate IRs for Out-of-Calibration SPVs | The inspectors identified a finding of very low safety significance (Green) because PBAPS did not initiate issue reports (IR) to identify out-of-tolerance conditions for a number of single point vulnerability (SPV) instruments. An SPV instrument is any instrument for which a single failure could initiate a plant transient or cause a plant scram. Specifically, during routine preventative maintenance (PM) calibrations, certain SPV instruments as-found data was found outside expected tolerance bands, with many being significantly outside of their bands. In most cases, IRs were not written to document these adverse conditions contrary to station guidance. The finding is determined to be more than minor because it affected the reliability of the initiating cornerstones attribute of equipment performance and affected its objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during power operations. Specifically, by not identifying and trending out-of calibration SPVs in a timely manner, a resulting transient from the loss of a single feed pump or a single reactor recirculation pump is more likely to occur. The inspectors conducted a Phase 1 screening in accordance with NRC Inspection Manual Chapter (IMC) Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings, and determined that the finding was of very low safety significance (Green) because the finding did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition (e.g. loss of condenser, loss of feed water.) A loss of a single feed pump or a single recirculation pump typically results in a power reduction but not a reactor scram. The inspectors determined that the finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Identification. In the case of the finding, PBAPS did not ensure that degraded conditions, namely, out of tolerance SPV instruments, were promptly reported and documented in the corrective action program at a low threshold. |
| 05000277/FIN-2015001-01 | 2015Q1 | Green | H.3 | NRC identified | Failure to Scope Flood Detection Level Switches into the MR | The inspectors identified a non-cited violation (NCV) of very low safety significance (Green) of 10 CFR Part 50.65, "Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," because Exelon did not include certain flood indication functions into the scope of the maintenance rule (MR). Specifically, level switches used to indicate flood levels in the Unit 2 and Unit 3 emergency core cooling system (ECCS) rooms were not included in the scope of the MR as required by 10 CFR 50.65 (b)(2)(i) as non-safety related components that are used in plant emergency operating procedures (EOPs). PBAPS entered the issue into their corrective action program (CAP) as issue reports (IRs) 02433897 and 02437502 and scoped the level switches into the MR. The finding is determined to be more than minor because it is associated with the protection against external factors attribute of the Mitigating Systems cornerstone and affected the cornerstones objective to ensure the reliability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). In the case of this finding, monitoring of components that provide alarm indication to operators during a flood hazard were not incorporated into the MR. The inspectors also reviewed IMC 0612, Appendix E, Examples of Minor Issues, and determined the issue was similar to example 7.d; in that, flood detection was not within the scope of the MR and that one of the flood detectors had experienced performance problems during preventive maintenance (PM) testing . The inspectors conducted a Phase 1 screening in accordance with IMC 0609.04, Phase 1 Initial Screening and Characterization of Findings, and determined that the finding was of very low safety significance (Green), because the finding was not a design or qualification deficiency, did not represent an actual loss of system safety function, did not represent an actual loss of safety function of a single train for greater than its Technical Specification (TS) allowed outage time, and did not screen as risk significant due to external initiating events. The inspectors determined that the finding had a cross-cutting aspect in the area of Human Performance, Change Management because PBAPS did not use a systematic process for evaluating and implementing a change. Specifically, during PBAPSs MR database update and monitoring criteria development for new functions, PBAPS did not ensure that certain level switches that provide alarms for flooding used in plant EOPs were scoped into the MR despite identifying that it was required. (H.3) |
| 05000277/FIN-2014004-01 | 2014Q3 | Green | P.2 | NRC identified | Corrective Actions Not Timely for EOC of Appendix R Broken Wires | The inspectors identified a Green non-cited violation (NCV) of the PBAPS Units 2 and 3 operating licenses, Section 2.C.4, Fire Protection, because Exelon did not have the ability to implement all provisions of their approved Fire Protection Program as described in the Updated Final Safety Analysis Report (UFSAR). Specifically, UFSAR Section 5.2.2, Appendix R, Shutdown Method D, was found degraded due to the loss of the alternate 125 volts direct current (Vdc) control power to both E-2 and E-4 alternate shutdown panels. The alternate 125 Vdc power was found degraded during a planned inspection due to broken electrical wires located in the safety-related E-23 4.16 kilovolt (kV) breaker cubicle associated with the E-2 alternate shutdown panel. The extent-of-condition (EOC) corrective actions were not timely to identify and correct similar broken wires in the E-43 4.16 kV breaker cubicle associated with the E-4 alternate shutdown panel. PBAPS entered the following issue reports (IRs) into their corrective action program (CAP): IR 01629839, 01656255, 01662555, and 01662767. Exelon completed repairs of the broken wires in both electrical breaker cubicles. The finding is more than minor because it is associated with the external events (fire) attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, following a postulated control room abandonment fire, the analyzed normal method was unavailable for closing three 4 kV circuit breakers locally with the switchgear mounted switch. Using IMC 0609, Appendix F, Fire Protection SDP, the Region I Senior Reactor Analyst (SRA) determined per Figure F.1, Phase 1 Flow Chart, and associated screening criteria 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 Problem Identification and Resolution (PI&R), Evaluation, because Exelon did not complete the EOC action in a timely manner commensurate with its safety significance. Specifically, the decision to implement corrective actions to address the EOC two months after the identification of the first breaker cubicle broken wire was not timely and commensurate with its safety significance. Additionally, the condition potentially existed for a longer period of time, but was not identified by established maintenance procedures. Even though the E-43 4.16 kV breaker wires could be checked without affecting the operability or availability of the E-4 emergency diesel generator (EDG), Exelon decided to perform the E-43 4.16 kV EDG breaker cubicle inspection during a future scheduled overhaul. Exelons corrective action procedure defines an immediate EOC concern when, as in this case, a work group evaluation (WGE) is required. |
| 05000277/FIN-2014004-02 | 2014Q3 | Green | H.5 | Self-revealing | Scaffold Obstructs A RHR Discharge Check Valve | A self-revealing finding was identified involving an NCV of very low safety significance (Green) for Technical Specification (TS) 5.4.1 Procedures, because Exelon did not correctly implement procedure MA-MA-796-024-1001, Revision 8, Scaffold Criteria for the Mid-Atlantic Stations. In addition, work order (WO) C0244158, Open/Close CHK-2- 10-48A for OPS Torus Support, instructions were not implemented as written to remove a gag (i.e., eyebolt) on the Unit 2 A residual heat removal (RHR) pump discharge check valve, CHK-2-10-48A, following restoration of the 2 A RHR system after a September 16, 2012, maintenance and fill activity. By not implementing these procedures and instructions, the eyebolt prevented full closure of CHK-2-10-48A after the 2 A RHR pump was secured. Exelon entered these issues into their CAP as IR 1680741, IR 1690648, and action request (AR) 02387793. Exelon removed the eyebolt and scaffold midrail to prevent any obstruction of movement on CHK-2-10-48A. The finding is more than minor because it affected the Mitigating Systems cornerstone attribute of equipment performance in the area of reliability and availability of the 2 A RHR train. Specifically, due to the stuck open check valve during a postulated loss of coolant accident (LOCA)/loss of offsite power (LOOP) scenario, voiding could occur and create a potential water hammer resulting in pipe support damage. This finding was determined to be of very low safety significance (Green) using IMC 0609, Appendix A, Exhibit 2, because the finding did not represent a loss of system function, did not represent a loss of a single train for greater than its allowed TS outage time, and did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding, or severe weather initiating event. Additionally, the inspectors determined that the function of 2 A RHR remained available because RHR piping would remain intact and containment cooling would not have been lost during the postulated water hammer scenario. The finding has a cross-cutting aspect in Human Performance, Work Management, because in the case of the erected scaffold, Exelon did not plan, control, and execute work activities such that nuclear safety was the overriding priority. Specifically, the work process did not coordinate effectively with different groups (i.e., operations, engineering, scaffold builders, and maintenance) and job activities to identify and preclude the scaffold from obstructing an eyebolt attached to the swing arm of the 2 A RHR pump discharge check valve. |
| 05000277/FIN-2014004-03 | 2014Q3 | Green | H.7 | NRC identified | Inadequate Evacuation Time Estimate Submittals | The inspectors identified a Green 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 PBAPS, Units 2 and 3, Emergency Plan. The station did not provide the evacuation time estimate (ETE) to the responsible offsite response organizations (OROs) by the required date. Exelon entered this issue into its CAP as IR 1525923 and IR 1578649. Additionally, Exelon re-submitted a new revision of the Peach Bottom ETE to the NRC on May 2, 2014. The performance deficiency is more than minor because it is associated with the Emergency Preparedness cornerstone attribute of procedure quality and it adversely affected the cornerstone objective of ensuring that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The finding was determined to be of very low safety significance (Green) because it was 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 the cross-cutting element of Human Performance, Documentation, because Exelon did not appropriately create and maintain complete, accurate and, up-to-date documentation. |
| 05000277/FIN-2014404-01 | 2014Q2 | NRC identified | Security | |||
| 05000277/FIN-2014007-02 | 2014Q2 | Green | P.2 | NRC identified | Non-Conservative Voltage Assumption Used to Verify MOV Capability | The team identified a Green non-cited violation of Title 10 Code of Federal Regulations 50, Appendix B, Criterion III, Design Control. Specifically, Exelon did not correctly verify the capability of alternating current motor-operated valves (MOVs) at a degraded voltage corresponding to the lowest voltage allowed by plant Technical Specification setpoints for the degraded grid voltage relays. Exelon initiated issue report 1642720 to evaluate the adequacy of their design and determined that 9 out of the 130 alternating current MOV program valves required further evaluation. The licensee performed an operability evaluation of the affected MOVs, assuming the appropriate voltage, and determined that, although significant design margin was lost, all MOVs remained operable. The finding was more than minor because the finding was associated with the Mitigating Systems Cornerstone attribute of design control and adversely affected the cornerstone objective of ensuring the capability of the 480 volt alternating current (AC) MOVs to respond to initiating events to prevent undesirable consequences. The team determined the finding was of very low safety significance because it was a design deficiency confirmed not to result in a loss of operability. The team assigned a cross-cutting aspect associated with this finding, because the deficient AC MOV operability evaluations were completed in November 2011 and were reflective of current performance. The finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation (PI.2), because Exelon did not thoroughly evaluate the issue addressed in a previous NCV contained in NRC Inspection Report 2010004, during 2011, for PBAPS such that, the resolution addressed causes and extent-of-condition commensurate with the safety significance. Specifically, the affected MOVs were not evaluated at the required voltage in operability evaluations performed following receipt of a non-cited violation. |
| 05000277/FIN-2014007-01 | 2014Q2 | Green | P.2 | NRC identified | Deficient E2 EDG Loading Calculation Design | The team identified a Green non-cited violation of Title 10 Code of Federal Regulations 50, Appendix B, Criterion III, Design Control, for failure to verify and ensure that the emergency diesel generators (EDGs) were capable of performing their design safety functions at the limits of voltage and frequency allowed by Technical Specifications (TS). Specifically, the existing EDG loading calculation permitted the E2 EDG and associated bus to be loaded up to 3100 KW at nominal frequency and voltage. At the maximum frequency and voltage values permitted by TS, the calculation-allowed maximum load would have exceeded the EDG 30-minute rating limit of 3250 KW and potentially damaged the EDG. Immediate corrective actions included evaluation of EDG loading for TS maximum voltage and frequency and changing design calculation PE-0166 to reduce the maximum permitted E2 EDG load from 3100 kW to 3052 kW at nominal voltage and frequency. Exelon entered the issue into their corrective action program (issue report 1638255) to evaluate the adequacy of the design and ensure that the allowed maximum diesel loading would not exceed the design capabilities of the diesels. The finding was more than minor because it was associated with the Mitigating Systems Cornerstone attribute of design control and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of the emergency diesels to respond to initiating events to prevent undesirable consequences. The team evaluated the finding in accordance with Inspection Manual Chapter (IMC) 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated June 19, 2012, for the Mitigating Systems Cornerstone, and IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012. The team determined the finding was of very low safety significance because it was a design deficiency confirmed not to result in a loss of EDG operability. This team assigned a cross-cutting aspect associated with this finding because the performance deficiency continued during the 2012 assessment of WCAP-17308-NP and was reflective of current performance. The team determined this finding had a crosscutting aspect in the area of Problem Identification and Resolution, Evaluation (PI.2), because engineers did not thoroughly evaluate the EDG loading issue and ensure the resolution addressed its cause commensurate with the safety significance. Specifically, Exelon relied on invalid assumptions to determine the issue was not applicable, and did not thoroughly evaluat the technical issue addressed in the WCAP. |
| 05000277/FIN-2014201-01 | 2014Q1 | Green | Licensee-identified | Security | ||
| 05000277/FIN-2014002-01 | 2014Q1 | Green | Licensee-identified | Licensee-Identified Violation | Title 10 of CFR Part 50.65 (a)(4) requires, in part, that before performing maintenance activities (including but not limited to surveillance, PMT, and corrective and preventive maintenance), the licensee shall assess and manage the increase in risk that may result from the proposed activities. Contrary to the above, on January 30, 2014, PBAPS did not initially assess an increase in plant risk resulting in an upgrade in established risk classification from yellow to orange. PBAPSs additional risk management actions, required by procedure, were delayed. On January 30, 2014, at 2:55 am, PBAPS removed their SBO line from service for planned maintenance and upgraded on-line risk to yellow for the duration of the maintenance activity. At 5:55 am, Pennsylvania-Jersey-Maryland (PJM) Interconnection issued a Maximum Emergency Generation Action for the Mid- Atlantic Region. However, as required, PBAPS was not notified at this time by a Power Team Generation Dispatch. A reactor operator monitoring PJMs website subsequently noticed the Maximum Emergency Generation Action. During a followup call to the Power Team Generation Dispatch contact, the Peach Bottom reactor operator was erroneously told that the grid emergency did not apply to nuclear power plants. In accordance with Exelons risk model and procedures, a Maximum Emergency Generation Action requires an upgrade to the next color risk category. For PBAPSs configuration with the SBO OOS, a risk upgrade from yellow to orange was required. At 7:58 am, PBAPS was notified of the Maximum Emergency Generation Action, identified that their current risk category was incorrect, upgraded the plant risk to orange, and directed the safety tagout clearance on the SBO line to be suspended until the grid emergency was lifted. PBAPS also identified that this issue was a repeat problem from a similar event on July 18, 2012. This previous event, documented in IR 1389933 and IR 1390285, was for PBAPS not being notified as required of a grid emergency by the Power Team Generation Dispatch. The inspectors determined that the finding was of very low safety significance (Green) in accordance with Flowchart 1 of Appendix K of IMC 0609, Maintenance Risk Assessment and Risk Assessment Significance Determination Process, because the incremental core damage probability deficit was significantly less than one E-6. PBAPS was in the less conservative risk category for approximately two hours. The inspectors reviewed PBAPSs planned corrective actions, which were to train power team dispatchers and revise applicable procedures to address the communication problem between generation dispatch and PBAPS. The inspectors considered the planned corrective actions appropriate. Because this finding is of very low safety significance and the issue was entered into Exelon's CAP under IRs 1614646 and 1615043, this violation is being treated as a Green NCV consistent with the NRCs Enforcement Policy. | |
| 05000277/FIN-2013005-01 | 2013Q4 | Green | Licensee-identified | Licensee-Identified Violation | TS 3.4.3 Limiting Condition for Operation requires that 11 of 13 SRVs\SVs shall be operable in reactor operating modes 1, 2, and 3. TS 3.4.3.1 surveillance requirement states that the SRVs\SVs opening lift setpoints are maintained within 1% tolerance of the design opening pressure. Contrary to the above, information received by site engineering from a laboratory performing SRV\SV as-found testing, determined that on October 1, 2013, the valve setpoint deficiencies existed with four SRVs and one SV that were in place during the Unit 3 19th operating cycle. The SRVs/SV were determined to have their as-found setpoints outside of the TS allowable 1% tolerance. The four SRVs and one SV outside of their TS allowable setpoint range were within the ASME Code allowable 3% tolerance. The cause of the SRVs/SV being outside of their allowable as-found setpoints was due to setpoint drift. The SRVs/SV were replaced with refurbished SRVs/SV for the 20th Unit 3 operating cycle. The amount of setpoint drift was within the as found Target Rock SRV values when compared to industry data. The SRVs/SV were replaced with refurbished valves that were tested and opened within the allowable 1% tolerance. The inspectors determined that the finding was of very low safety significance (Green) in accordance with Section A of Exhibit 2 in Appendix A of IMC 0609, The SDP for Findings at Power, because the SRVs safety function was not affected. Although outside the lift setpoint tolerance, the as-found SRV/SV lift pressure values would not have challenged the reactor vessel design maximum pressure rating during the most limiting postulated accident event. The inspectors reviewed PBAPSs planned corrective actions to address the SRV setpoint drift issue and a planned industry standard TS setpoint change submittal to a 3% tolerance appropriate. Because this finding is of very low safety significance, the as-found out of tolerance SRVs were replaced with SRVs that had the proper lift setpoint prior to the Unit 3 reactor plant startup, and the issue was entered into Exelon's CAP under Issue Report 1567200, this violation is being treated as a Green NCV consistent with the NRCs Enforcement Policy. | |
| 05000277/FIN-2013405-01 | 2013Q3 | Licensee-identified | Licensee-Identified Violation | |||
| 05000277/FIN-2013004-02 | 2013Q3 | Green | H.5 | NRC identified | Failure to Conspicuously Post and Lock/Guard a HRA on the Unit 3 Turbine Deck Scaffold | The inspectors identified a NCV of very low safety significance of Technical Specification (TS) 5.7.2 because Exelon did not control the access point to a Locked High Radiation Area (LHRA). The performance deficiency (PD) was related to not controlling access to a Unit 3 LHRA. The LHRA became accessible when temporary scaffold was built on the south shield wall between the electrical generator and the main turbine. On August 19, the inspectors identified a permanent ladder from the top of the north side of the shield wall to the turbine deck floor that could allow access to the LHRA. Radiation Protection (RP) procedure RP-AA-460, Controls for High and LHRA, Revision 24, provides guidance for the control of high radiation areas (HRAs). By the procedure definition of accessible area, the area was accessible after the scaffold was built, and no tools or other exceptional measures were needed to gain access. The violation was entered into Exelons corrective action program (CAP) as action request (AR) 01548397. The PD was more than minor because it is associated with the cornerstone attribute of Program and Process (RP controls), and negatively affected the Occupational Radiation Safety cornerstone objective to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear power operation. There was also an example of this PD in example 6.g. of IMC 0612, Appendix E, Examples of Minor Issues. This example concludes that the issue is more than minor because actual dose rates in excess of the posting requirements existed in the area. LHRAs are required to be posted and controlled properly to avoid unnecessary worker exposure. The finding was evaluated using the Occupational Radiation Safety SDP and was determined to be of very low safety significance (Green) because it was not related to As Low As is Reasonably Achievable (ALARA) planning, it did not involve an overexposure, did not constitute a substantial potential for overexposure, and the ability to access dose was not compromised. The finding included a cross-cutting aspect in the area of Work Controls, Human Performance component, because Exelon did not appropriately plan the work activities and identify the potential job site conditions (radiological hazards) associated with building scaffold next to a LHRA wall. |
| 05000277/FIN-2013004-01 | 2013Q3 | Green | H.4 H.5 | NRC identified | Inadequate EP Procedure Change Management Controls to Ensure Adequate EAL Classification and Assessment Capability for Effluent Parameters | The inspectors identified a Green non-cited violation (NCV) of Title 10 of the Code of Federal Regulation (CFR) 50.54(q)(2) associated with 50.47(b)(4) because PBAPS failed to control emergency planning (EP) procedure changes in a manner that would ensure timely emergency action level (EAL) classification for effluent parameters. On June 27, 2013, PBAPS issued Revision 27 to EP-AA-1007, Exelon Nuclear Radiological Emergency Plan Annex for PBAPS. One of the plan changes involved removal of the \'A\' ventilation and main stack radiation monitors from radiological effluent EAL matrix Table 3-1, and thereby rendered the B ventilation and main stack radiation monitors as the only means of EAL classification for effluent releases. On July 24, 2013, the inspectors questioned shift operations on whether the ability to make timely and accurate EAL classifications was impacted with the B reactor building (RB) ventilation stack radiation monitor inoperable. Shift operations did not have an immediate response, but later in the same shift provided a response to the inspectors that compensatory measures were required for degraded EP equipment, and the \'A\' ventilation stack radiation monitor was established as a compensatory measure for the inoperable \'B\' monitor in response to questions by the inspectors. Following the inspectors questions, PBAPS initiated issue report (IR) 1539674 to capture programmatic deficiencies that were revealed as a result of the inspectors questions. PBAPS corrective actions included a revision to the PBAPS Emergency Plan, a revision to the EP compensatory measure procedure, issuance of Operations Information Update (OIU) 13-10 to the shift managers (SMs) to clarify the purpose of the compensatory measure procedure, and an assignment to incorporate the latest revision of the compensatory measure procedure into licensed operator training program curriculum review committee (CRC). This finding was more than minor because it was associated with the procedure quality attribute of the Emergency Preparedness cornerstone, and adversely affected the associated cornerstone objective to ensure that the licensee is capable of implementing adequate measures to protect the public health and safety in the event of a radiological emergency. Using IMC 0609, Attachment 4, Phase 1 Initial Screening and Characterization of Findings, and IMC 0609, Appendix B, Emergency Preparedness SDP, the inspectors determined that this finding was of very low safety significance (Green) using Table 5.4.1. Specifically, this finding rendered an EAL ineffective such that an unusual event (UE) declaration could be delayed. The inspectors determined that this finding had a cross-cutting aspect in the area of Human Performance, Work Control, because PBAPS did not appropriately coordinate work activities by incorporating actions to address the impact of work on different job activities, and the need for work groups to communicate, coordinate, and cooperate with each other during activities in which interdepartmental coordination is necessary to assure plant and human performance (H.3(b)). Specifically, the impact of a PBAPS Emergency Plan Annex revision was not communicated properly or coordinated between the EP department and operations department, to assure that shift operations could implement compensatory measures as necessary for degraded EP equipment. |
| 05000277/FIN-2013002-01 | 2013Q1 | Green | H.14 | NRC identified | Inadequate Operability Determination in Response to Power Load Unbalance Device Failure | The inspectors identified a Green finding for PBAPS\\\'s failure to follow the operability determination (OD) process described in Procedure OP-AA-108-115, Operability Determinations. Specifically, on February 24, 2013, between 6:15 a.m. and 10:30 a.m., an immediate determination of operability was not made in a timely manner, and was not initially documented in accordance with the corrective action process (CAP), following discovery that Unit 2 was operating outside of the analyzed limits specified in the core operating limits report (COLR) with the power load unbalance (PLU) circuit out of service (OOS). Consequently, operators entered the Unit 2 minimum critical power ratio (MCPR) technical specification limiting condition for operation (TS LCO) 3.2.2, Condition A, after exceeding the two-hour required action completion time. The inspectors determined that the immediate determination of operability was not performed in a matter commensurate with the safety significance of the two-hour LCO required action completion time. The inspectors determined that this was not a violation of TSs because subsequent analysis by a third party vendor determined that MCPR thermal limits were satisfied between 85 percent and 100 percent reactor power with the PLU circuit OOS on Unit 2. This finding is more than minor because it is associated with the design control attribute of the barrier integrity cornerstone and adversely affected the cornerstone objective of providing reasonable assurance that the physical design barriers (fuel cladding) protect the public from radionuclide releases caused by events. Using IMC 0609, Attachment 4, Phase 1 Initial Screening and Characterization of Findings, and IMC 0609, Appendix A, The SDP for Findings At-Power, the inspectors determined that this issue screened to Green, because it was associated only with the fuel cladding barrier. The inspectors determined that this finding had a cross-cutting aspect in the area of Human Performance, decisionmaking, because PBAPS did not use conservative assumptions in decision making and did not adopt a requirement to demonstrate that the proposed action was safe in order to proceed rather than a requirement to demonstrate that it is unsafe in order to disprove the action. |
| 05000278/FIN-2013011-01 | 2013Q1 | Severity level IV | NRC identified | Failure to Comply with a Posted High Radiation Area Boundary | The OI investigation, which was completed on March 14, 2013, was conducted to determine whether a PBAPS instrumentation and controls (I&C) technician deliberately failed to follow posted high radiation area (HRA) requirements when he crossed a boundary to manipulate a valve. The investigation was initiated after Exelon informed the NRC, on June 28, 2012, that the PBAPS I&C technician in question had potentially willfully failed to comply with a posted HRA boundary. This was contrary to Exelon procedures which requires, as indicated in the HRA radiation work permit (RWP), a HRA briefing prior to entering a HRA. Based on the evidence gathered during the OI investigation, the NRC concluded that on June 27, 2012, the I&C technician deliberately failed to follow posted HRA requirements when he crossed a HRA boundary during a Unit 3 High Pressure Coolant Injection (HPCI) system test. Specifically, the I&C technician crossed a posted HRA boundary and entered the Unit 3 HPCI room without a HRA briefing or the proper RWP. This conclusion was based on the I&C technicians admission to OI that he had done the wrong thing when he crossed the HRA boundary without the correct RWP; his experience and training working in the RCA; and his acknowledgement that he had alternative options that he should have chosen before violating HRA boundary requirements. The I&C technicians actions caused Exelon to violate the PBAPS Unit 3 operating license. Specifically, Technical Specification 5.4.1 requires that written procedures be established, implemented, and maintained covering the applicable procedures recommended in Regulatory Guide 1.33, Appendix A, dated November 1972. Regulatory Guide 1.33, Appendix A, Section G, dated November 1972, recommends procedures for control of radioactivity, including restrictions and activities in radiation areas (G.5.a), and RWPs (G.5.e). Exelon Procedure RPAA- 460, Revision 20, Section 4.3.2, requires, in part that a HRA briefing is required to enter a HRA. Because the violation was caused by the deliberate action of the I&C technician, it was evaluated under the NRCs traditional enforcement process using the factors set forth in the NRC Enforcement Policy. After careful consideration of these factors, the NRC concluded that this violation should be classified at Severity Level (SL) IV. In reaching this decision, the NRC considered that the significance of the underlying violation was minor because, while the I&C technician crossed a posted HRA boundary, the radiological conditions at the time did not actually constitute a HRA area in accordance with the regulatory definition of a HRA. However, the NRC decided to increase the significance of this violation to SL IV since it was deliberate and the NRCs regulatory program is based, in part, on licensees and their contractors acting with integrity. In accordance with Section 2.3.2 of the Enforcement Policy, and with the approval of the Director, Office of Enforcement, this issue has been characterized as a non-cited violation (NCV 05000278/2013011-01, Failure to Comply with a Posted High Radiation Area Boundary), because: (1) Exelon placed the issue in its CAP (CR No. 1382220); (2) Exelon identified the issue and immediately conducted an investigation; (3) the violation was not repetitive as a result of inadequate corrective action; and, (4) although the violation was willful, (a) Exelon identified the violation, notified the NRC, and took significant corrective and remedial actions; (b) the violation involved the acts of an individual who was not considered a licensee official with oversight of regulated activities as defined in the Enforcement Policy; and (c) the violation did not involve a lack of management oversight and was the result of the isolated action of the employee. The NRC has concluded that information regarding the reason for the violation, the corrective actions taken and planned to correct the violation and prevent recurrence, and the date when full compliance was achieved is already adequately addressed on the docket in this letter. | |
| 05000277/FIN-2012005-02 | 2012Q4 | Green | Licensee-identified | Licensee-Identified Violation | TS LCO 3.3.1.1, Condition B, requires that with one RPS instrument function with one or more required channels inoperable, action shall be taken within six hours to place a channel or trip system in a trippedcondition within six hours. Additionally, TS LCO 3.3.4.2, Condition A, requires that with one or more required end of cycle (EOC) recirculation pump trip (RPT) instrument channels inoperable, action be taken to place the channel in a tripped condition within 72 hours if the channel is not restored to operable status. Contrary to the above, PBAPS determined that the A and B channels of the Unit 2 turbine control valve (TCV) fast closure pressure sensing instruments were inoperable for a period of time greater than allowed by TS. Specifically, the as-found trip setpoints of the A and B sensing instruments were identified to be below the allowable trip setting during surveillance testing on October 1, 2012. PBAPS Unit 2 was defueled to support the 19th RFO during performance of the ST. Both instruments were replaced and calibrated to within acceptable limits prior to reactor startup. The inspectors determined that the finding was of very low safety significance (Green) in accordance with Section C of Exhibit 2 in Appendix A of IMC 0609, The Significance Determination Process for Findings at Power, because RPS system trip capability was maintained with the C and D instrument channels. Because this finding is of very low safety significance and has been entered into Exelon\'s CAP under IR 1421069, this violation is being treated as a Green NCV consistent with the NRC Enforcement Policy. | |
| 05000277/FIN-2012005-03 | 2012Q4 | Green | Licensee-identified | Licensee-Identified Violation | 10 CFR 50.54(q) requires, in part, that a power reactor licensee follow an Emergency Plan that meets the requirements of 10 CFR 50.47(b). 10 CFR 50.47(b) requires, in part, that a standard emergency classification and action level scheme, the bases of which includes facility system and effluent parameters, is in use by Exelon. Contrary to the above, between December 2008 and November 2012, the standard emergency classification and action level scheme associated with radiological effluents at PBAPS was not updated to reflect the changes in X/Q dispersion factor that occurred during the December 2008 ODCM revision. Consequently, the effluent monitor emergency classification and action level thresholds for the reactor building exhaust vent stack were non-conservative until this condition was identified and promptly corrected by PBAPS in November 2012. The inspectors determined that the finding was of very low safety significance (Green) in accordance with NRC IMC 0609, Appendix B, Emergency Preparedness Significance Determination Process, Table 5.4-1, because the emergency action level (EAL) classification process would not be capable of classifying an Unusual Event (UE) within 15 minutes, but would still be capable of declaring all other EALs within 15 minutes. Because this finding is of very low safety significance, and has been entered into Exelon\'s CAP under IR 1439489, this violation is being treated as a Green NCV consistent with the NRCs Enforcement Policy. | |
| 05000277/FIN-2012005-01 | 2012Q4 | Green | Licensee-identified | Licensee-Identified Violation | TS 3.4.3 Limiting Condition for Operation (LCO) requires that 11 of 13 SRVs\\SVs shall be operable in reactor operating modes 1, 2, and 3. TS 3.4.3.1 surveillance requirement states that the SRVs\\SVs opening lift setpoints are maintained within + 1% tolerance of the design opening pressure. Contrary to the above, information received by site engineering from a laboratory performing SRV\\SV as-found testing, determined that on September 25, 2012, the valve setpoint deficiencies existed with six SRVs and one SV that were in place during the Unit 2 19 operating cycle. The SRVs /SV were determined to have their as-found setpoints outside of the TS allowable + 1% tolerance. The six SRVs outside of their TS allowable setpoint range were within the ASME Code allowable + 3% tolerance. The one SV outside of its TS allowable setpoint range also slightly exceeded the ASME Code allowable + 3% tolerance at a value of + 3.4%. The cause of the SRVs /SV being outside of their allowable as-found setpoints was due to setpoint drift. The SRVs /SV were replaced with refurbished SRVs/SV for the 20th Unit 2 operating cycle. The amount of setpoint drift was within the as found Target Rock SRV values when compared to industry data. The SRVs/SV were replaced with refurbished valves that were tested and opened within the allowable + 1% tolerance. The inspectors determined that the finding was of very low safety significance (Green) in accordance with Section A of Exhibit 2 in Appendix A of IMC 0609, The Significance Determination Process for Findings at Power, because the SRVs safety function was not affected. Although outside the lift setpoint tolerance, the as found SRV/SV lift pressure values would not have challenged the reactor vessel design maximum pressure rating during the most limiting postulated accident event. The inspectors reviewed PBAPSs planned corrective actions to address the SRV setpoint drift issue and considered a planned industry standard TS setpoint change submittal to a + 3% tolerance appropriate. Because this finding is of very low safety significance, the as-found out of tolerance SRVs were replaced with SRVs that had the proper lift setpoint prior to the Unit 2 reactor plant startup, and the issue was entered into Exelon\'s CAP under IR 1418320 and apparent cause evaluation 1120516, this violation is being treated as a Green NCV consistent with the NRCs Enforcement Policy. | |
| 05000277/FIN-2012004-01 | 2012Q3 | Green | H.5 | Self-revealing | Inadequate Preplanning and Performance of Maintenance/Modifications Resulted in Unavailability of RHR B Loop | The inspectors identified a Green, self-revealing non-cited violation (NCV) of Technical Specification (TS) 5.4.1, Procedures. The inspectors determined that PBAPS did not properly preplan and perform maintenance/modifications to the Unit 2 low pressure coolant injection (LPCI) swing bus B motor control cabinet (MCC) while energized. Specifically, PBAPS did not appropriately consider the potential plant impact due to sensitive energized components within the MCC that could be activated and did not utilize sufficient physical barriers to prevent such activation. Consequently, on July 25, 2012, the B loop of the residual heat removal (RHR) system was declared inoperable and unavailable after workers pulling an electrical cable into the Unit 2 energized LPCI swing bus B MCC inadvertently contacted and actuated the LPCI inboard injection valve motor relay. The motor operated valve (MOV) relay actuation caused a potential over-thrust event and had the potential to impact the valves qualification and reliability. PBAPS conducted detailed examinations and diagnostic stroke testing on the MOV assembly and concluded that the design limits of the MOV assembly were not exceeded. This finding was more than minor because it was associated with the equipment performance attribute of the Mitigating System cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined that this finding was of very low safety significance (Green) because it did not represent an actual loss of safety function of a single LPCI train for greater than its TS allowed outage time. The inspectors determined that this finding had a cross-cutting aspect in the area of Human Performance, work control, because PBAPS did not appropriately incorporate risk insights and job site conditions that could impact plant structures, systems, and components (SSCs) into its work activities. Specifically, PBAPS did not appropriately consider and reduce the potential for an over-thrust event on the B loop LPCI inboard injection valve MO-2-10-25B when performing work in the LPCI swing bus B MCC while it was energized. |