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05000277/FIN-2018003-012018Q3Peach BottomHPCI System Exhaust Pressure Switches Exceeded Documented Qualified LifeThe 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-2018003-022018Q3Peach BottomInadequate Corrective Actions Result in the Failure of the E-3 EDGThe 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-2018003-032018Q3Peach BottomReactor Core Isolation Cooling System Pressure Switch Failure Results in Condition Prohibited by TS - EA-18-108On 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-2018002-012018Q2Peach BottomFailure to Identify and Promptly Correct a Condition Adverse to Quality Concerning Battery Charger 2B-003-1The 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
05000410/FIN-2018002-012018Q2Nine Mile PointFailure to Ensure Proper Control of the Standby Gas Treatment System Damper Valve, 2GTS*V2000B, Within Procedures, Materials, and Design Control MeasuresThe inspectors identified a Green finding and associated NCVof 10 CFRPart 50, Appendix B, Criterion III, Design Control, when Exelon failed to ensure proper control of the SGTS damper valve 2GTS*V2000B within procedures, materials, and design control measures. Specifically, on April 15, 2018 operators attempted to run B SGTS for containment purge; however, no flow was observed and the system was secured. Operators discovered the 2GTS*V2000B closed due to the failure of the operating mechanism to maintain control of the valve position.
05000220/FIN-2018002-022018Q2Nine Mile PointInadequate Procedure Causes Water Hammer Condition Resulting in Isolation and Inoperability of the 12 Train of the Emergency Condenser SystemThe inspectors identified a Green finding and associated NCVof 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, when Exelon did not provide appropriate quantitative or qualitative criteria and guidance to operators in procedure N1- OP- 13 Emergency Cooling System to return an emergency condenser loop to service without inducing a water hammer condition which caused operators to re-isolate the emergency condenser loop and declare it inoperable
05000278/FIN-2018001-012018Q1Peach BottomUntimely Corrective Actions to Address Primary Containment Isolation Valve Condition Adverse to QualityA 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)
05000278/FIN-2017003-012017Q3Peach BottomInstructions Not Followed for Replacement of HPSW Ventilation Switch BlockA 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-2017003-022017Q3Peach BottomLicensee-Identified Violation10 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.
05000286/FIN-2017002-012017Q2Indian PointFailure to Maintain Flow Channeling Gate Closed in Accordance with the Containment ProcedureGreen. The inspectors identified a Green NCV of Technical Specification (TS) 5.4.1, Procedures, for Entergys failure to implement procedure OAP-007, Containment Entry and Egress. Specifically, workers transiting the inner and outer crane wall sections of containment on May 14, 2017, did not maintain flow channeling gate C secured during Mode 4 to ensure availability of the containment sumps to provide suction for the emergency core cooling system (ECCS). Entergy immediately restored gate C to an acceptable configuration, and generated condition report (CR)-IP3-2017-02737 to address this issue. This performance deficiency was more than minor because it is associated with the configuration control (shutdown equipment lineup) attribute and adversely affected the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). A detailed risk assessment was conducted and the change in core damage frequency was determined to be 2E-8, therefore, this issue represents a Green finding. The inspectors determined that this finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Resolution, because Entergy did not take effective corrective actions to address issues in a timely manner commensurate with their safety significance. Specifically, the corrective actions from the event for the prior year were ineffective at preventing this occurrence. (P.3)
05000278/FIN-2017002-012017Q2Peach BottomCorrective Action Not Implemented Correctly for Replacement of RCIC RCR ContactsA 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-2017002-022017Q2Peach BottomEDG Exhaust Stacks Nonconforming Design for Tornado Missile ProtectionOn 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-2016004-012016Q4Peach BottomFailure to Identify and Remove FM in CAD System PipingGreen. 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.
05000387/FIN-2016004-012016Q4SusquehannaFailure Rates Exceed (20%) Twenty Percent for Biennial Requalification ExamGreen. A self-revealing finding was identified associated with inadequate licensed operator performance during the annual licensed operator requalification operating test and biennial written examination. Specifically, 17 of 71 operators (23.9%) failed at least one portion of the requalification examinations. This finding is more than minor because it is associated with the Mitigating Systems cornerstone attribute of human performance and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, 17 of 71 licensed operators failed to demonstrate a satisfactory understanding of the required knowledge and abilities required to safely operate the facility under normal, abnormal, and emergency conditions. The inspectors evaluated this performance deficiency using IMC 0609, SDP, Appendix I, Licensed Operator Requalification SDP. This finding is of very low safety significance (Green) because the finding is related to requalification exam results, did not result in a failure rate of greater than 40 percent and all 17 operators were remediated and successfully retested prior to returning to licensed duties. This finding has a cross-cutting aspect in the area of Human Performance, Training, because Susquehanna did not provide adequate operator requalification training to maintain a knowledgeable, technically competent workforce. (H.7)
05000387/FIN-2016004-022016Q4SusquehannaFailure to Promptly Correct a Condition Adverse to Quality with LPCI Swing Bus Automatic Transfer SwitchesGreen. A finding of very low safety significance (Green) and associated NCV of Title 10 Code of Federal Regulations (CFR) 50, Appendix B, Criterion XVI, Corrective Action, was self-revealed when Susquehanna failed to assure that conditions adverse to quality were promptly identified and corrected on two separate occasions. Both examples resulted in the failures of safety-related automatic transfer switches (ATSs) associated with the low pressure coolant injection (LPCI) swing buses. Corrective actions included enhancing the work instructions for all applicable ATSs based off original equipment manufacturer (OEM) input and scheduling the enhanced work instructions to be performed on the four swing bus ATSs during their next scheduled bus outages. Inspectors determined that the finding was more than minor because it was associated with the Equipment Performance attribute of the Reactor Safety 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 (i.e., core damage). In both examples, the failure to correct conditions adverse to quality resulted in the loss of power to the LPCI swing bus and inoperability of the respective division of LPCI. In accordance with IMC 0609.04, Initial Characterization of Findings, dated June 19, 2012, inspectors and Exhibit 2 of IMC 0609, Appendix A, The SDP for Findings At-Power, dated June 19, 2012, inspectors determined that the finding was of very low safety significance (Green). Specifically, though a single train was inoperable for greater than its technical specification (TS) allowed outage time, in consultation with regional senior reactor analysts, inspectors determined it did not represent an actual loss of function. The finding is related to the cross-cutting area of Problem Identification and Resolution, Evaluation, because Susquehanna did not thoroughly evaluate issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, Susquehanna either failed to evaluate deficiencies encountered during maintenance or failed to ensure that corrective actions aligned with and corrected the identified causes. (P.2)
05000388/FIN-2016004-032016Q4SusquehannaRefuel Floor Radiation Monitor Inoperable Due to being Improperly CalibratedGreen. A finding of very low safety significance (Green) and NCV of TS 5.4.1, Procedures was self-revealed when Susquehanna incorrectly calibrated the Unit 1 B refuel floor high exhaust duct high radiation monitor on November 15, 2014. This impacted the initiation capability of secondary containment isolation and control room emergency outside air supply system (CREOASS) and resulted in Susquehanna exceeding the allowed outage time for TSs 3.3.6.2, Secondary Containment Isolation, and 3.3.7.1, CREOASS Instrumentation. Upon identification of the issue, Susquehanna properly calibrated the radiation monitor to restore its operability. This finding is more than minor because it is associated with the Human Performance (Routine OPS/Maintenance Performance) attribute of the Barrier Integrity cornerstone and affected the cornerstone objective of providing reasonable assurance that physical design barriers (Secondary Containment and Control Room Ventilation) protect the public from radionuclide releases caused by accidents or events. Specifically, incorrectly calibrating the radiation monitor resulted in both systems being inoperable for almost two years. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 3 of IMC 0609, Appendix A, The SDP for Findings At-Power, both dated June 19, 2012, the inspectors determined that this finding is of very low safety significance (Green) because the performance deficiency was only associated with the radiological barrier function of the Control Room and Secondary Containment. This finding had a cross-cutting aspect in the area of Human Performance, Avoid Complacency because Susquehanna did not recognize and plan for the possibility of mistakes, latent problems, or inherent risk, even while expecting successful outcomes. Specifically, Susquehanna personnel did not consider the potential undesired consequences of their actions before performing work and implement appropriate error-reduction tools (e.g. self-check, peer-check). (H.12)
05000387/FIN-2016004-042016Q4SusquehannaAuxiliary Bus Load Shed when a Daisy Chained Neutral was Interrupted during MaintenanceGreen. A finding of very low safety significance (Green) for failure to develop an adequate work plan for replacement of a voltage potential indicating light on a breaker on the Unit 2 B auxiliary bus was self-revealed when the Unit 2 B reactor recirculation pump (RRP) tripped, along with other non-safety related loads on November 14, 2016, resulting in a rapid unplanned power change and transition to single loop operation. Specifically, operations and maintenance personnel did not recognize that disconnecting the neutral wires from the light socket would interrupt power to all of the degraded voltage relays for the auxiliary bus. Therefore, the relays de-energized when the maintenance was performed, tripping all the breakers on the bus. Susquehannas immediate corrective actions included stabilizing the plant, entering single loop operations, and entering the issue into their corrective action program (CAP). Additionally, Susquehanna performed a maintenance department stand down to communicate immediate lessons learned from the event while a more thorough causal analysis was conducted. The performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Initiating Events cornerstone and affected its objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, implementation of work instructions resulted in the trip of the Unit 2 B RRP, B and D circulating water (CW) pumps, B and D condensate pumps, and the B service water (SW) pump, which caused an automatic trip of the C reactor feed pump and runback of the A RRP, resulting in a rapid power reduction to 32 percent rated thermal power (RTP). The inspectors evaluated the finding in accordance with IMC 0609, Appendix A "The SDP for Findings At-Power," dated June 19, 2012, Exhibit 1 for the Initiating Events cornerstone and determined the finding was of very low safety significance (Green) because it did not cause a reactor trip. This finding was determined to have a cross-cutting aspect in the area of Human Performance, Work Management because Susquehanna did not implement a process of planning work activities such that nuclear safety is the overriding priority, including the identification and management of risk commensurate with the work. Specifically, Susquehanna did not recognize the risk of interrupting a daisy chained neutral when planning a minor maintenance work order and did not recognize the impact of the work activity in the field. (H.5)
05000387/FIN-2016004-052016Q4SusquehannaLERs Associated with Reactor Coolant Pressure Boundary LeakageEnforcement. TS 3.4.4, "RCS" requires RCS leakage be limited to no pressure boundary leakage in Mode 1. Contrary to this, pressure boundary leakage from a LPRM instrument housing and from socket weld #8 occurred between plant start-up in December 2015 and plant shutdown on June 6, 2016, and existed while in Mode 1. The inspectors determined that these violations of TS 3.4.4 are more than minor, but not the result of performance deficiencies. Specifically, for the first event, though leakage likely existed during the previous refueling outage when personnel were performing unrelated maintenance and inspection activities, it was likely too small to reasonably identify and correct. Similarly, for the 2016 leak identified in weld #8, the leakage causes were not within Susquehannas ability to foresee as they had replaced the weld with the industry recommended 2 x 1 taper configuration and used qualified procedures and personnel. The Susquehanna staff had also measured the susceptibility of the attached piping for vibrational inputs. In accordance with the NRC Enforcement Policy guidance and IMC 0612, these violations are being treated under the traditional enforcement process and best characterized as a Severity Level (SL) IV (very low safety significance) violation, similar to example d.1 in NRC Enforcement Policy, Section 6.1, Reactor Operations. Although a performance deficiency was not identified, to verify that the issue was of very low safety significance, the inspectors considered risk insights obtained by using IMC 0609, SDP, Appendix A, Exhibit 1, Initiating Events Screening Questions. The inspectors determined that these TS violations would screen to Green (very low safety significance) because the boundary leakage would not have exceeded the leak rate for a small loss of coolant accident (LOCA) and would not affect any LOCA accident mitigating systems or components. Therefore, the inspectors considered that the SL IV characterization was appropriate. The licensee entered these issues into the Susquehannas CAP as CR-2016-14544 and CR-2016-14366. Because these issues are of very low safety significance, it has been determined that it was not reasonable for Susquehanna to be able to foresee and prevent, and as such no performance deficiencies exist. The NRC has decided to exercise enforcement discretion in accordance with Sections 2.2.4 and 3.5 of the NRC Enforcement Policy and refrain from issuing enforcement action for the violation of TS (EA-16-283). Further, because Susquehanna's actions did not contribute to this violation, it will not be considered in the assessment process or the NRC's Action Matrix.
05000277/FIN-2016003-012016Q3Peach BottomReactor Feed Pump Controller Power Supply Shelf Life Not MaintainedA 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-2016002-032016Q2Peach BottomHuman Performance Event Results in Emergent DownpowerA 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-2016002-012016Q2Peach BottomImproperly Stored Material in Reactor BuildingThe 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-2016002-022016Q2Peach BottomUntimely Corrective Actions to Address Condition Adverse to the Fire Protection Program Alternative Shutdown CapabilityThe 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.
05000278/FIN-2016001-012016Q1Peach BottomLicensee-Identified ViolationOn 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-2015004-012015Q4Peach BottomFailure to Ensure Design Basis of Emergency Diesel Generator Lubrication SystemThe 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.
05000272/FIN-2015009-012015Q3SalemLicensee-Identified Violation10 CFR 50.65 (a)(2) states, in part, that monitoring as specified in 10 CFR 50.65 (a)(1) is not required where it has been demonstrated that the performance or condition of an structure, system, or component is being effectively controlled through the performance of appropriate preventive maintenance, such that the structure, system, or component remains capable of performing its intended function. Paragraph (a)(1) requires, in part, that licensee shall monitor the performance of SSC within the scope of the rule against licensee-established goals in a manner sufficient to provide reasonable assurance the SSC are capable of fulfilling their intended safety functions. Contrary to the above, PSEG failed to recognize that the plant level performance goal of only one reactor trip per year had been exceeded in the spring of 2014. Following the failure, PSEG failed to consider performing an evaluation under 10 CFR 50.65(a)(1) for establishing goals and monitoring against the goals until February 2015. PSEGs nuclear oversight organization identified this deficiency and documented it in Notification 20678696. The inspectors evaluated this finding using IMC 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix A, Exhibit 1, Initiating Events Screening Questions, and screened it to Green as 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.
05000387/FIN-2015003-012015Q3SusquehannaRHR Shutdown Cooling Procedure Not Maintained Consistent with Technical Specification RequirementsInspectors identified a finding of very low safety significance (Green) and associated NCV of SSES Unit 1 and 2 TS 5.4.1, Procedures, because Susquehanna did not maintain the procedure for operation of the residual heat removal (RHR) shutdown cooling (SDC) system consistent with the requirements in TS 3.4.8, RHR Shutdown Cooling- Hot Shutdown. As TS 3.4.8 requires two RHR SDC loops to be operable and, if no reactor recirculation pumps (RRPs) are running, one of the loops to be in-service in Mode 3 below the RHR cut in permissive pressure (98 psig), inspectors determined that OP-1(2)49-002, RHR Shutdown Cooling, was not maintained appropriately because a change to the procedure precluded operation of the system between 40 psig and the RHR cut in permissive pressure (98 psig). Susquehanna entered the issue into the corrective action program (CAP) as CR-2015-22882 and CR-2015-24137 and revised the procedure to remove the requirement that precluded operation of the SDC system between 40 psig and the RHR cut in permissive pressure This finding is more than minor because it is associated with the procedure quality attribute of the Mitigating Systems cornerstone and adversely affected its objective to ensure the availability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the 40 psig procedural limit impacted the availability and capability of RHR to be placed in SDC between 98 psi, the cut-in permissive for the system, and 40 psig. In accordance with Exhibit 2 of IMC 0609, Appendix A, The SDP for Findings At-Power, 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 TS allowed outage time, and did not screen as potentially risk-significant due to a seismic, flooding, or severe weather initiating event. This finding had a cross-cutting aspect in the area of Human Performance, Change Management because Susquehanna did not use a systematic process for evaluating and implementing change so that nuclear safety remains the overriding priority (H.3). Specifically, implementation of Susquehannas procedure change process did not ensure that the RHR SDC procedure was maintained consistent with the requirements of plant TSs.
05000387/FIN-2015003-022015Q3SusquehannaC EDG Rendered Inoperable by Switch Manipulation during Training SimulationA self-revealing finding of very low safety significance (Green) and associated NCV of 10 CFR 50 Appendix B, Criterion V, Instructions, Procedures and Drawings, was identified when Susquehanna inadvertently operated the C emergency diesel generator (EDG) mode switch during the performance of a job performance measure (JPM). Specifically, the student performing the JPM operated plant equipment that was contrary to the quality assurance program requirement to only simulate equipment operation. Susquehanna entered the issue into the CAP as CR-2015-19578, the C EDG mode switch was restored to the Remote position, and the operating crew performed a walk-down of the C EDG to confirm proper standby alignment, restoring operability of the EDG. Inspectors determined that the finding was more than minor because it was associated with the Human Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the reliability of systems that respond to initiating events to prevent undesirable consequences. Specifically, improper manipulation of the C EDG mode switch while simulating a task resulted in an inoperable condition since the EDG would not have auto started, if required. In accordance with Exhibit 2 of IMC 0609, Appendix A, The SDP for Findings At-Power, dated 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 TS allowed outage time, and did not screen as potentially risk-significant due to a seismic, flooding, or severe weather initiating event. This finding had a cross-cutting aspect in the area of Human Performance, Avoid Complacency because Susquehanna did not implement appropriate error reduction tools (H.12). Specifically, personnel did not implement appropriate human error prevention tools (e.g. self-check, stop-think-act-review) in accordance with station processes.
05000387/FIN-2015003-032015Q3SusquehannaSecondary Containment Inoperability due to Improperly Controlled Access to the Reactor Building RoofA self-revealing finding of very low safety significance (Green) and associated NCV of SSES Unit 1 and 2 TS 5.4.1, Procedures, was identified because Susquehanna incorrectly implemented procedures for maintaining secondary containment integrity. Specifically, on July 27, 2015, maintenance technicians rendered secondary containment for both units inoperable for approximately 44 minutes when a secondary containment boundary door was opened to access the reactor building roof. Susquehanna entered the issue into the CAP as CR-2015-20857 and CR-2015-24442, restored the boundary, and verified the integrity of secondary containment. The finding was more than minor because it was associated with the Human Performance (Routine OPS/Maintenance Performance) attribute of the Barrier Integrity cornerstone, and affected the cornerstone objective of providing reasonable assurance that physical design barriers (Secondary Containment) protect the public from radionuclide releases caused by accidents or events. Specifically, opening the secondary containment barrier did not maintain reasonable assurance that the secondary containment would be capable of performing its safety function in the event of a reactor accident. The inspectors evaluated the finding in accordance with IMC 0609, Appendix A, "The SDP for Findings At-Power," Exhibit 3, for the Barrier Integrity cornerstone, dated June 19, 2012. The inspectors determined the finding was of very low safety significance (Green) because only represented a degradation of the radiological barrier function of secondary containment provided by the standby gas treatment (SBGT) system. This finding had a cross-cutting aspect in the area of Human Performance, Teamwork because Susquehanna did not effectively communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety was maintained (H.4). Specifically, when the work plan was changed to accessing the reactor building roof through secondary containment, the change was not effectively communicated to operations department personnel to ensure the secondary containment impairment was appropriately controlled.
05000388/FIN-2015003-042015Q3SusquehannaLicensee-Identified ViolationOn May 28, 2015, patrols by Susquehanna personnel identified that the panel doors on 2C664, located in the Unit 2 upper relay room, were left open and unattended following troubleshooting by station personnel. Susquehannas Unit 2 operating license condition 2.C.(6), requires, in part, that Susquehanna implement and maintain in effect all provisions of the approved fire protection program as described in the fire protection review report for the facility. Unit 2 Technical Requirements Manual (TRM) directs compensatory measures as required by the fire protection program for fire suppression system impairments. TRM bases for 3.7.3.4, halon systems, states the opening of any relay room panel door causes the affected Halon system to become inoperable if the panel door is unattended. TRM 3.7.3.4 requires a continuous fire watch with backup fire suppression equipment be established for inoperable halon systems. Contrary to the above, panel doors were left open and unattended without a continuous fire watch for a period of five hours and 35 minutes. Susquehanna entered this issue into the CAP as CR-2015-15709. The inspectors determined that the finding was more than minor because it was associated with the protection against external factors attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). The inspectors determined through a review of IMC 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix F, Fire Protection Significance Determination Process, issued September 20, 2013, the finding to be of very low safety significance (Green) based on the reactor maintaining the ability to reach and maintain safe shutdown condition. Specifically, safe shutdown path 3 systems and components would be available for safe shutdown in the event of a fire in the Unit 2 upper relay room.
05000277/FIN-2015003-012015Q3Peach BottomIncomplete Testing of Components from the Remote Shutdown PanelsThe 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-2015003-022015Q3Peach BottomLicensee-Identified ViolationThe 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.
05000387/FIN-2015003-052015Q3SusquehannaDegraded Condition Due to Reactor Coolant Pressure Boundary Leakage Caused by an Inadequate WeldOn December 13, 2014, after operators shutdown Unit 1 for a planned forced outage due to increased unidentified leakage, Susquehanna staff identified a small leak by observation on the Unit 1 "B" RRP Lower Seal Cavity Vent 34 piping, a location that Susquehanna determined was part of the reactor coolant pressure boundary. Susquehanna determined that this leakage constituted a violation of the Unit 1 TS, Section 3.4.4 titled "Reactor Coolant System (RCS)" that requires RCS leakage to be limited to no pressure boundary leakage. Susquehanna staff performed a progressive non-destructive examination of the leak site to further characterize the flaw in the socket weld. The evaluation concluded the flaw was a linear defect approximately 0.74 in length through the center of the socket weld and that the apparent cause was a weld defect created during the vendors original manufacturing process. The defective area was removed by progressive grinding with confirmatory liquid penetrant test examinations to confirm the defect length. The weld was repaired and inspected satisfactorily prior to plant startup from the Unit 1 forced outage. The inspector reviewed LER 2014-011 and CR-2014-37848, Revision 1 that documented the related apparent cause evaluation and failure modes analysis for this condition. The inspector also reviewed weld repair work package No. 140469 and the final nondestructive examination reports of the repair area. The inspector determined Susquehanna staffs leak review and analysis process concluded that the leakage, located in the center of the vendor supplied fillet weld was most likely the result of an original weld flaw. The leak area was removed by grinding and the area was repaired by welding to the required 2 to 1 dimensional fillet weld configuration. Although this leak did not appear to be fatigue or system vibration driven, the inspector noted Susquehanna staffs implementation of mitigation methods to reduce socket weld fatigue susceptibilities, including a 2 to 1 fillet weld configuration and instrumented hammer testing of the completed pipe assembly to confirm the absence of resonant frequency vibration modes, were applied. LER 2014-011 documents the licensees violation of TS 3.4.4, limiting reactor pressure boundary leakage during plant operations to zero. This TS violation occurred sometime between unit start-up in July 2014 and plant shutdown on December 13, 2014. The inspector determined that this violation of TS 3.4.4 was more than minor, but not the result of a performance deficiency. In accordance with the NRC Enforcement Policy guidance and IMC 0612, this violation is being treated under the traditional enforcement process and best characterized as a Severity Level IV (very low safety significance) violation, similar to example d.1 in NRC Enforcement Policy, Section 6.1, Reactor Operations. In addition, using IMC 0609, Significance Determination Process, Appendix A, Exhibit 1, Initiating Events Screening Questions, this TS violation screens to Green (very low safety significance) because the boundary leakage did not exceed the capacity of the control rod drive system or TS unidentified leakage (less than 5 gpm) and actual leakage did not adversely impact any LOCA mitigating systems or components. Because this issue is of very low safety significance and it has been determined that it was not reasonable for Susquehanna staff to be able to foresee and prevent this leakage, and as such no performance deficiency exists, the NRC has decided to exercise enforcement discretion in accordance with Sections 2.2.4 and 3.5 of the NRC Enforcement Policy and refrain from issuing enforcement action for the violation of TS (EA-15-149). Further, because Susquehanna's actions did not contribute to this violation, it will not be considered in the assessment process or the NRC's Action Matrix. The inspectors did not identify any new issues during the review of the LER. This LER is closed.
05000387/FIN-2015003-062015Q3SusquehannaDegraded Condition Due to Reactor Coolant Pressure Boundary Leakage Caused by Vibration and Stiff Pipe ConnectionOn April 11, 2015, during the initial drywell walk down following shutdown for the spring 2015 Unit 2 refueling outage, Susquehanna staff identified a small leak on the Unit 2 "A" RRP seal piping at a seal flange weld associated with the pressure and vent piping to the upper seal chamber (Connection #2). Water was observed spraying out of the top of the pipe in a fan pattern. Susquehanna staff determined that this leakage constituted a violation of the Unit 2 TS, Section 3.4.4, titled "RCS" that requires RCS leakage be limited to no pressure boundary leakage. Based on the unidentified leakage rate of 0.25 gpm measured during plant operation and visual inspection of the leak area, the leak most likely existed during plant operation. Susquehanna staff removed the seal flange and piping assembly containing the leak area and provided that material for metallurgical investigation. A qualified replacement assembly was installed with a modified pipe routing to provide for increased pipe flexibility and was inspected prior to and during startup from the Unit 2 refuel outage. The inspector reviewed LER 2015-004, CR-2015-09907, the apparent cause evaluation completed for this leak, the metallurgical and structural analysis inputs, and CR-2015-009953 regarding the extent of condition evaluation scope. The leakage cause was attributed to the short rigid piping span between the pump and H5005 support with atypical vibration between the two. Using a modified pipe routing to increase pipe flexibility, the seal flange with pipe assembly was replaced. The modified pipe configuration was instrumented and hammer tested to confirm its vibrational frequency characteristics. LER 2015-004 documents the licensees violation of TS 3.4.4, limiting reactor pressure boundary leakage during plant operations to zero. This TS violation occurred sometime between Unit 2 start-up in 2013 and plant shutdown on April 10, 2015. The inspector determined that this violation of TS 3.4.4 was more than minor, but not the result of a performance deficiency. In accordance with the NRC Enforcement Policy guidance and IMC 0612, this violation is being treated under the traditional enforcement process and best characterized as a Severity Level IV (very low safety significance) violation, similar to example d.1 in NRC Enforcement Policy, Section 6.1, Reactor Operations. In addition, using IMC 0609, Significance Determination Process, Appendix A, Exhibit 1, Initiating Events Screening Questions, this TS violation screens to Green (very low safety significance) because the boundary leakage did not exceed the capacity of the control rod drive system or TS unidentified leakage (less than 5 gpm) and actual leakage did not adversely impact any loss-of-coolant accident mitigating systems or components. Because this issue is of very low safety significance and it has been determined that it was not reasonable for Susquehanna staff to be able to foresee and prevent this leakage, and as such no performance deficiency exists, the NRC has decided to exercise enforcement discretion in accordance with Sections 2.2.4 and 3.5 of the NRC Enforcement Policy and refrain from issuing enforcement action for the violation of TS (EA-15-189). Further, because Susquehanna's actions did not contribute to this violation, it will not be considered in the assessment process or the NRC's Action Matrix. The inspectors did not identify any new issues during the review of the LER. This LER is closed.
05000277/FIN-2015008-012015Q2Peach BottomFailure to Initiate IRs for Out-of-Calibration SPVsThe 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.
05000387/FIN-2015002-012015Q2SusquehannaFailure to Assess a Non-Conforming Condition for its Impact on Component OperabilityThe inspectors identified a finding of very low safety significance (Green) and associated NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, when Susquehanna staff did not assess component operability following identification of a potentially non-conforming condition. Specifically, Susquehanna did not assess for operability a potential non-conforming condition associated with inadequate testing of the primary containment airlock inboard equalizing valve which was identified during the review of industry operating experience. Susquehannas corrective actions to restore compliance included entering this issue in their CAP as CR-2015-15187, performing a prompt operability determination of the Unit 1 primary containment airlock inboard equalizing valve, including completion of the requirements in SR 3.0.3 for a missed surveillance, and performing testing on the Unit 2 valve which adequately demonstrated that the PCIV was operable prior to entering into a mode of TS applicability. The inspectors determined that the finding was more than minor because it was associated with the SSC and Barrier performance attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective of providing reasonable assurance that the physical design barriers (containment) protect the public from radionuclide releases caused by accidents or events. Specifically, inadequate actions to evaluate the impact of the condition adverse to quality on the operability of the Unit 1 PCIV resulted in a reasonable doubt of operability of the barrier. In accordance with IMC 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Exhibit 2 of IMC 0609, Appendix A, The SDP for Findings At-Power, dated June 19, 2012, the inspectors determined that this finding is of very low safety significance (Green) because the performance deficiency did not represent an actual open pathway in the physical integrity of reactor containment and heat removal components or involve the actual reduction in function of hydrogen igniters in containment. This finding has a cross-cutting aspect in the area of Human Performance, Avoid Complacency, because Susquehanna did not perform a thorough review of the work and planned activity but rather relied on past successes and assumed conditions. Specifically, the control room staff did not assess the condition for operability believing that it was similar to previous CRs documenting a review of operating experience.
05000387/FIN-2015002-022015Q2SusquehannaEntry into a High Radiation Area without Radiological BriefingA self-revealing finding of very low safety significance (Green) and associated NCV of SSES Unit 2 TS 5.7.1 was identified because Susquehanna did not comply with a radiological posting barrier and other protective measures for HRA entry. Specifically, on October 10, 2014, two workers entered the turbine building roof, a posted HRA, but the workers were not on the proper RWP and were not briefed on the radiological conditions prior to entry. Upon receiving a dose rate alarm, the workers exited the HRA and reported the issue to radiation protection personnel. Susquehanna entered the issue into the CAP as condition report CR-2014-31911. The inspectors determined that Susquehannas inadequate adherence to a high radiation area (HRA) posting, which requires a HRA RWP and a radiological briefing prior to entry, was a performance deficiency that was within Susquehannas ability to foresee and correct and should have been prevented. The inspectors determined that the finding was more than minor because it adversely affected the human performance attribute of the Occupational Radiation Safety cornerstone objective. Specifically, the individual violated the RWP and briefing requirements designed to protect the worker from unnecessary radiation exposure. The issue was also similar to example 6.h in IMC 0612, Appendix E. Using IMC 0609, Appendix C, Occupational Radiation Safety SDP, dated August 19, 2008, the finding was determined to be of very low safety significance (Green) because it did not involve: (1) as low as is reasonably achievable (ALARA) occupational collective exposure planning and controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. This finding has a cross-cutting aspect of Human Performance, Challenge the Unknown, because the workers did not stop when faced with uncertain conditions. Specifically, the workers did not use a questioning attitude during the pre-job brief or when they encountered the HRA posting on the access to the turbine building roof.
05000387/FIN-2015002-032015Q2SusquehannaIncorrect Implementation of the Ventilation Filter Testing ProgramThe inspectors identified a finding of very low safety significance (Green) and associated NCV of 10 CFR 50, Appendix B, Criterion XI, Test Control, because Susquehanna did not ensure representative samples were obtained from Engineered Safety Feature (ESF) filter ventilation systems and did not establish written test procedures. Specifically, subsequent to refilling charcoal test canisters for the activated charcoal absorber of both trains of the SBGT System, new charcoal was added to the activated charcoal absorber which was not exposed to the same service conditions as the bulk of the absorber section as required by TS 5.5.7, Ventilation Filter Testing Program, and written test procedures were not established for this activity. As corrective action for the identified issue, Susquehanna replaced the charcoal in the A and B trains of SBGT and the A and B trains of CREOASS activated charcoal absorber beds and test canisters between January and February 2015 and initiated condition reports CR-2014-39116 and CR-2015-01443. The inspectors determined that the finding was more than minor because it was associated with the Procedure Quality Attribute of the Barrier Integrity Cornerstone and it adversely affected the cornerstone objective to provide reasonable assurance that physical barriers protect the public from radionuclide releases caused by accidents or events. Specifically, since 2001, work instructions did not prevent the contamination of test canisters with charcoal that was not representative of the in-service conditions of the adsorber bed and the introduction of new charcoal into the test canisters likely provided better results during periodic surveillance testing which were not representative of actual conditions. In accordance with IMC 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Exhibit 3 of IMC 0609, Appendix A, The SDP for Findings At-Power, dated June 19, 2012, the inspectors determined that the finding was of very low safety significance (Green) because it only represented a degradation of the radiological barrier function provided for the control room and SBGT system. This finding has a cross-cutting aspect in the area of Human Performance, Documentation, because the activities for sampling the activated charcoal beds were not governed by comprehensive, high-quality programs, processes, and procedures nor were the design documentation, procedures, and work packages complete, thorough and accurate.
05000387/FIN-2015002-042015Q2SusquehannaMultiple Violations of Work Hour Limitations by Licensed OperatorsThe inspectors identified a finding of very low safety significance (Green) and associated NCV of 10 CFR 26.205, Work Hours, because Susquehanna did not ensure that the working hours of licensed operators were maintained within regulatory limits. Specifically, numerous instances of violations were identified in the operations department in which individuals exceeded the required work hour limits while performing duties subject to work hour controls. In review of the issue, the inspectors identified that Susquehanna inappropriately excluded some works hours performing non-covered work from the total accumulated work hours, which allowed individuals to perform covered work while in excess of the work hour limits without meeting the requirements for applying a waiver. Susquehanna entered the issue into the CAP as CR-2015-15708 and initiated action to evaluate the extent of the matter and communicate the issue with the operations department, reinforce the standards for work hour tracking with station personnel, and ensure work hours are appropriately tracked. The inspectors determined that the finding was more than minor because Susquehanna inadequately implemented the requirements of 10 CFR 26.205 and NDAP-QA-0025 routinely. Therefore, if the performance deficiency were left uncorrected, the continued process of not including all hours accumulated toward work hour limits and allowing workers to exceed work hour limits, had the potential to lead to a more significant safety concern. The finding was also similar to IMC 0612, Appendix E, "Examples of Minor Issues," Example 9.a. In accordance with IMC 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Exhibits 1 and 2 of IMC 0609, Appendix A, The SDP for Findings At-Power, dated June 19, 2012, the inspectors determined that this finding is of very low safety significance (Green) because no transients, loss of function of a mitigating system, or mismanagement of reactivity occurred as a result of licensed operators performing covered work while not in compliance with the work hour limits specified in 10 CFR 26.205. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Identification, because Susquehanna did not identify the issues completely, accurately, and in a timely manner. Specifically, Susquehanna identified violations of work hour limits on multiple occasions but the CRs were not in sufficient detail to ensure they were appropriately prioritized and assigned for resolution. Individuals did not recognize that work performed doing non-covered work was to be counted as hours accumulated towards the work hour limitations and thus discounted the violations as erroneous.
05000388/FIN-2015002-052015Q2SusquehannaLoss of Main Condenser Vacuum When Transitioning Steam Seals to Auxiliary SteamA self-revealing finding of very low safety significance (Green) and associated NCV of SSES Unit 2 TS 5.4.1, Procedures, was identified because Susquehanna incorrectly implemented procedures for operation of the auxiliary steam and main turbine steam sealing systems. Specifically, on April 10, 2015, while Unit 2 was being shut down for a RFO, operators secured main turbine steam seals resulting in degraded main condenser vacuum. The degraded main condenser vacuum resulted in a main turbine trip, which caused an automatic reactor scram from approximately 37% power. Susquehanna restored main condenser vacuum by reestablishing steam seals, performed off-normal and emergency operating procedures to stabilize the plant post-scram and entered the issue into the corrective action program (CAP) as CR-2015-09890. The finding was more than minor because it was associated with the Human Performance attribute of the Initiating Events cornerstone and affected its objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, not understanding the impact of securing auxiliary steam to the main turbine steam seals resulted in the degradation of main condenser vacuum, automatic trip of the main turbine and associated reactor scram. The inspectors evaluated the finding in accordance with IMC 0609, Appendix A "The SDP for Findings At-Power," Exhibit 1, for the Initiating Events cornerstone, dated June 19, 2012. The inspectors determined the finding was of very low safety significance (Green) because it did not cause a reactor trip and the loss of mitigation equipment. Specifically, though a reactor scram occurred, operators were able to restore main condenser vacuum prior to MSIV closure and the main condenser and reactor feed pumps remained functional during the event. This finding has a cross-cutting aspect in the area of Human Performance, Avoid Complacency, because Susquehanna did not implement appropriate error reduction tools. Specifically, operators did not effectively implement human error prevention tools (e.g. pre-job briefing, stop-think-act-review) in accordance with station processes.
05000277/FIN-2015001-012015Q1Peach BottomFailure to Scope Flood Detection Level Switches into the MRThe 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)
05000293/FIN-2014008-032014Q4PilgrimParallel White Unplanned Scrams per 7000 Critical Hours PI FindingThe inspectors identified deficiencies regarding Entergys execution of corrective actions documented in the RCEs, as well as understanding of some of the causes of the issues. Specifically, inspectors identified several examples in the RCEs where corrective actions were not completed as intended or were closed prematurely. Additionally, for one of the RCEs, inspectors determined that Entergy failed to investigate a deficient condition sufficiently to ensure they fully understood all of the causes of the event. Inspectors determined that the specific deficiencies in execution of CAP procedures discussed in the findings in sections 02.02.e.1 and 02.03.f.1 of this report were indicative of the CAP implementation weakness that Entergy identified as part of their common cause and safety culture evaluations. Correspondingly, inspectors determined that corrective actions specified in the common cause and safety culture evaluations were not effective at ensuring that all the causes of the performance issues were understood and that corrective actions taken were adequate to address the identified root and contributing causes. Taken collectively, the issues associated with the two White PIs represent a significant weakness, as discussed in IP 95002.
05000293/FIN-2014008-042014Q4PilgrimParallel White Unplanned Scrams with Complications PI FindingThe inspectors identified deficiencies regarding Entergys execution of corrective actions documented in the RCEs, as well as understanding of some of the causes of the issues. Specifically, inspectors identified several examples in the RCEs where corrective actions were not completed as intended or were closed prematurely. Additionally, for one of the RCEs, inspectors determined that Entergy failed to investigate a deficient condition sufficiently to ensure they fully understood all of the causes of the event. Inspectors determined that the specific deficiencies in execution of CAP procedures discussed in the findings in sections 02.02.e.1 and 02.03.f.1 of this report were indicative of the CAP implementation weakness that Entergy identified as part of their common cause and safety culture evaluations. Correspondingly, inspectors determined that corrective actions specified in the common cause and safety culture evaluations were not effective at ensuring that all the causes of the performance issues were understood and that corrective actions taken were adequate to address the identified root and contributing causes. Taken collectively, the issues associated with the two White PIs represent a significant weakness, as discussed in IP 95002.
05000293/FIN-2014008-012014Q4PilgrimFailure to Fully Derive the Causes of a Manual ScramOn August 22, 2013, Pilgrim station experienced a failed splice on a nonsafety- related power supply to a level control valve, which caused an electrical transient that directly led to the automatic trip of all three RFPs when combined with a latent issue related to a modification. Entergy performed a RCE of the event and identified the failed splice as a direct cause. Inspectors determined that Entergy failed to investigate the cause of the electrical transient in accordance with station CAP procedures sufficiently to ensure all of the root and contributing causes of the event were understood. Their investigation determined that the splice failed because it was improperly fabricated when it was installed in 1999 as a part of a modification package on balance of plant valves. The splice was inside a flexible conduit, had two splices on parallel wires places right next to each other instead of being staggered, and the splice had not been properly crimped. Entergy performed an extent of condition review to verify there were no other splices installed in flexible conduit by the same modification package. Entergy used multiple causal evaluation methods as part of their RCE. Three of these (event and causal factor charting, failure modes analysis, and the why staircase analysis) discussed the failed splice. Entergys CAP procedures state that neither the failure modes analysis nor the why staircase are acceptable stand-alone methods of evaluation. The failure modes analysis method is described in EN-LI-118-08, Failure Modes Analysis, Revision 2. This procedure states that the output of the failure modes analysis will only be the direct cause, so it must be used with another method. The why staircase method is described in EN-LI-118-11, Why Staircase, Revision 0. This procedure states that for human performance problems, the method may only get to the general area of the cause and most likely will require further analysis to establish the exact cause. In the why staircase for this RCE, Entergy stopped at failure to follow requirements of design change and (procedure) 3.M.3-51, which is a human performance issue. The event and causal factor charting method is described in EN-LI-118-01, Event and Causal Factor Charting, Revision 2, and is the only method of the three intended for stand-alone use. This procedure directs them to continue to investigate and develop the chart until one of the following limits is reached: (1) the cause is outside the control of Entergy, (2) the correction of the cause is determined to be cost prohibitive, (3) the primary effect is fully explained, or (4) there are no other causes that explain the effect being evaluated. Entergy terminated their chart at maintenance work practices, which does not meet any of the criteria listed in EN-LI-118-01. Ultimately, Entergy did not identify corrective actions to correct the cause of the failed splice and inspectors could not verify that actions taken to ensure other splices were not improperly installed were sufficient. The inspectors questioned why Entergy had not continued to investigate what caused the splice to be improperly fabricated in accordance with station procedures such that either appropriate corrective actions could be planned or justification as to why no corrective actions were required could be provided. In response to inspectors questions, Entergy entered the issue into the CAP as CR-PNP-2014-5796 and initiated additional causal analysis to determine why the splice was improperly fabricated.
05000293/FIN-2014008-022014Q4PilgrimFailure to Complete Several Corrective Actions as Required by Program RequirementsIn 2013, four reactor scrams occurred at Pilgrim which resulted in two PIs in the Initiating Events cornerstone crossing the Green to White threshold. To address these risk significant performance issues, both individually and collectively, Entergy performed four RCEs for the individual scram events which occurred on January 10, February 8, August 22, and October 14. Additionally, Entergy performed a RCE to assess the commonalities between the four scram events and CCA to assess if any safety culture aspects caused or significantly contributed to the events. EN-LI-102, Corrective Action Program, Revision 23, provides instructions for the administration of Entergy corrective action process, including the identification, reporting, evaluation, and correction of a broad range of problems, areas for improvements, and standards performance deficiencies. Issues addressed in the corrective action process must include Adverse Conditions and Conditions Adverse to Quality, and can include minor problems that may be precursors to more significant events, areas for improvement and standards performance deficiencies identified during assessments and other activities. To that end, EN-LI-102 contains instructions for review and approval of corrective action development, response and documentation, and due date extensions. Section 5.6(4) of EN-LI-102 states that corrective action response must address the intent of the action and must not indicate correction or implementation based on future action. In review of the corrective action plans and status of completed or scheduled corrective actions to address the risk-significant performance issues, inspectors identified multiple deficiencies in implementing the CAP procedure. As documented in section 4OA4.02.03.a.1, a.2, a.4, and d.2 of this report, inspectors identified that some of the corrective actions specified in the RCEs were not completed in accordance with CAP requirements. Specifically, inspectors identified: Several of the human performance related corrective actions from the RCE of a scram on January 10, 2013, during surveillance testing had been cancelled or closed. To determine whether the closure of the corrective actions was appropriate, inspectors reviewed a sampling of recently performed surveillances and observed performance of maintenance in the field. During this review, inspectors noted that several procedures did not have critical steps annotated as such, one procedure directed work to be performed following system restoration, and identified examples of technicians proceeding with testing when challenged with test equipment challenges or unexpected system response. Ultimately, inspectors determined that observations of maintenance execution did not support closure or cancellation of corrective actions identified in the RCE and determined that the numerous procedure deficiencies and human performance issues identified by inspectors represented conditions adverse to quality that were reasonably within Entergys ability to identify and correct by execution of corrective actions identified in the RCE; Despite corrective actions to upgrade severe weather procedures to address deficiencies revealed during a winter storm on February 8, 2013, inspectors identified that the procedure changes did not fully meet the intent of the corrective actions because there were no substantive changes to the procedures for pre-storm actions. Additionally, inspectors determined that the inadequate guidance for pre-storm actions represented a condition adverse to quality that was reasonably within Entergys ability to identify and correct by execution of corrective actions identified in the RCE; An action to send a transformer insulator that faulted offsite for vendor analysis was not completed as required by CAP requirements. Despite being selfidentified by Entergy in preparation for the inspection, this deficiency still existed at the time of the inspection; and One of two effectiveness reviews for a RCE was not completed as required by CAP requirements. Entergy entered the issues into the CAP as CR-PNP-2014-5909, CR-PNP-2014-5976, CR-PNP-2014-5977, CR-PNP-2014-5682, CR-PNP-2014-5625, CR-PNP-2014-5826, CR-PNP-2014-5735, and CR-PNP-2014-06067 and took action to address the identified deficiencies. In particular, for the first example, Entergy revised the RCE to include additional corrective actions for procedural reviews prior to performance of work and enhanced oversight of maintenance activities. For the second example, Entergy made numerous additional changes to severe weather procedures. As discussed in section 4OA4.02.05.b.1, Entergys site safety culture review adequately identified the components of nuclear safety culture that caused or significantly contributed to the four scram events. In particular, inspectors noted Entergy identified that implementation of the stations CAP has not been effective in ensuring adequate corrective actions are taken to address issues in a timely manner, and Entergy identified corrective actions to improve performance in this area. As discussed in section 4OA4.02.05.a.5, inspectors identified examples of approved corrective actions intended to address challenges in CAP implementation not being completed as prescribed. For these corrective actions, inspectors discussion with Entergy personnel revealed that the performance improvement department determined that the actions were not producing the desired results, and decided to implement CAP training instead. However, Entergy did not present these change to the CARB for approval of the revision as required by EN-LI-102. This was entered into the CAP as CR-PNP-2014-06040. Ultimately, inspectors determined that the specific deficiencies in execution of corrective actions identified by inspectors were symptomatic of the identified challenges in CAP implementation, and correspondingly determined that corrective actions specified in the common cause and safety culture evaluations were not effective at ensuring that performance issues identified by the numerous RCEs were corrected.
05000277/FIN-2014004-012014Q3Peach BottomCorrective Actions Not Timely for EOC of Appendix R Broken WiresThe 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-2014004-022014Q3Peach BottomScaffold Obstructs A RHR Discharge Check ValveA 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-2014004-032014Q3Peach BottomInadequate Evacuation Time Estimate SubmittalsThe 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-2014002-012014Q1Peach BottomLicensee-Identified ViolationTitle 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.
05000388/FIN-2014002-022014Q1SusquehannaReactor Scram due to Loss of Reactor Feed PumpsA finding of very low safety significance (Green) for failure to implement work instructions for an engineering change to the Integrated Control System (ICS) was self-revealed when Unit 2 lost control of reactor vessel level on September 14, 2013, requiring insertion of a manual scram. The cause of the loss of level control was determined to be a coding error in the ICS that resulted in the improper transition of feedwater control modes during a reactor shutdown. PPLs immediate corrective actions included entering the issue into their corrective action program (CAP) as condition report 1746169, correcting the coding error, and performing and extent of condition review of the ICS code to ensure no additional errors were present. The performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Initiating Events cornerstone and affected its objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the failure to implement work instructions associated with the engineering change resulted in an ICS logic code error which caused a loss of reactor feed requiring a manual reactor scram. The inspectors evaluated the finding in accordance with IMC 0609, Appendix A, The SDP for Findings At-Power, Exhibit 1 for the Initiating Events cornerstone. The inspectors determined the finding was of very low safety significance (Green) because it did not cause both a reactor trip and the loss of mitigation equipment. This finding was determined to have a cross-cutting aspect in the area of Human Performance, Work Management because PPL did not implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority, including the identification and management of risk commensurate to the work. Specifically, the work instructions associated with the engineering change lacked the specificity commensurate with the complexity of the work such that it could be accomplished without error.
05000387/FIN-2014002-032014Q1SusquehannaLicensee-Identified Violation10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires in part, that activities affecting quality shall be prescribed by documented instructions and procedures of a type appropriate to the circumstances. Contrary to the above, prior to July 6, 2012, it was identified that PPL had not incorporated adequate written guidance in TP-264-032, Core Flow Calibration, Revision 5, to require iteration of the procedural steps used for the calibration check if flow instrumentation summer gains were adjusted. This resulted in core flow for the A recirculation loop being adjusted to approximately 2.4 Mlb/hr below actual loop flow and the B recirculation loop being adjusted to approximately 0.2 Mlb/hr below actual loop flow. PPL entered the issue into the corrective action program as CR 1708878. Inspectors determined this finding to be of very low safety significance (Green) in accordance with IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process (SDP) for Findings At-Power, because none of the logic questions under the barrier integrity cornerstone applied, indicating the issue screened to Green. Inspectors reviewed PPLs technical evaluation and determined that there was adequate margin in the thermal limit calculations to ensure that no safety or operating limits were exceeded. This issue was discussed in further detail within Section 4OA3 of this report.
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