Source: https://kanterella.com/wiki/Special:Ask/-5B-5BCategory:Finding-5D-5D-20-5B-5BSite.Reactor-20type::GE-2D2-5D-5D-20-5B-5BIdentified-20by::!Licensee-2Didentified-5D-5D/-3FStart-20date/-3FSite/-3FTitle/-3FDescription/mainlabel%3D/order%3Ddescending/sort%3DStart-20date/offset%3D0/format%3Dtable/searchlabel%3D-5BGE-2D2-20events-5D
Timestamp: 2020-07-04 14:25:34
Document Index: 764905826

Matched Legal Cases: ['art 50', 'art 50', 'art 50', 'art 50', 'art 1', 'art 2', 'art 50', 'art 50', 'art 50', 'art-20', 'art-20']

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05000219/FIN-2018410-03 30 June 2018 23:59:59 Oyster Creek Security
05000410/FIN-2018001-02 31 March 2018 23:59:59 Nine Mile Point Potential Inadequate 50.59 Evaluation for TS 3.3.1.1 Bases Change On February 23, 2018, Exelon personnel performed a 50.59 Screening for a change to Unit 2 TS Bases 3.3.1.1, Reactor Protection System (RPS) Instrumentation, for MSIV and TSV surveillance testing. Exelon personnel performed this activity to address operating experience associated with the use of a test box that prevents a scram signal during RPS surveillance testing for TS 3.3.1.1 Function 5 MSIV Closure and Function 8 TSV Closure. TS Bases B 3.3.1.1, C.1, Revision 1 was revised to state, in part, For Function 5 (MSIV Closure), this would require both trip systems to have at least one channel associated with the MSIVs for each main steam line in one Trip Logic Channel (not necessarily the same main steam lines for both trip systems), Operable or in trip (or the associated trip system in trip). For Function 8 (Turbine Stop Valve Closure), this would require both trip systems to have the channels for one Trip Logic Channel, Operable or in trip (or the associated trip system in trip).The inspectors questioned whether the change to TS Bases B 3.3.1.1 resulted in a change to the implementation of TS 3.3.1.1. A licensee is permitted to make changes to their Technical Specification Bases documents without NRC review and approval. However, in certain cases, such as a change to the Technical Specification Bases that would change how the associated Technical Specification is applied, NRC review and approval would be required.Planned Closure Action(s): The inspectors sent written questions to request assistance from NRR to determine whether this change to the TS Bases reasonably would have required NRC review and approval. The inspectors are opening a URI to determine if this is violation of 10 CFR 50.59 and if it is more than minor. Licensee Action: Documented NRCs concern as AR 04055602. Exelons position is the change would not affect how TS 3.3.1.1, or its note, is applied and therefore NRC review was not required.Corrective Action Reference: AR04055602 NRC Tracking Number: 05000410/2018001-02
05000219/FIN-2018001-01 31 March 2018 23:59:59 Oyster Creek Untimely Licensee Event Report for Reportable Conditions Associated with the No. 2 Emergency Diesel Generator The inspectors identified a non-cited, Severity IV violation of 10 CFR 50.73(a)(1) for a failure to submit a licensee event report (LER) within 60 days after the discovery of an event requiring a report. Specifically, on October 9, 2017, Exelon determined that the No. 2 emergency diesel generator was inoperable for longer than the allowed outage time, which is reportable as a condition prohibited by technical specifications. Exelon did not submit an LER for this event until January 3, 2018
05000219/FIN-2018001-02 31 March 2018 23:59:59 Oyster Creek Enforcement Action (EA)-18-007: No. 2 Emergency Diesel Generator Ring Lug Failure On October 9, 2017, during a routine surveillance load test, the No. 2 emergency diesel generator failed approximately 5 minutes into the run due to a broken ring lug on a current transformer. Laboratory analysis of the broken ring lug determined that the ring lug failed due to fatigue cracking that was initiated due to stresses caused by bending and twisting of the electrical lug. Exelon last conducted a load surveillance on the No. 2 emergency diesel generator on September 25, 2017. Corrective Actions: Corrective actions included replacement on the broken ring lug on the No. 2 emergency diesel generator, extent of condition inspections on the No. 1 and No. 2 emergency diesel generators for additional bent or twisted ring lug connectors, and revision to the electrical ring lug installation and emergency diesel generator procedures to include inspection for bent or twisted ring lugs. Corrective Action Reference(s): Issue report 4060815 Enforcement:Violation: Oyster Creek Technical Specification 3.7.C.2.b states, in part, that if one diesel generator becomes inoperable during power operation, the reactor may remain in operation for a period not to exceed 7 days. Contrary to the above, on October 9, 2017, it was recognized that one diesel generator was inoperable for greater than the technical specification allowed outage time of 7 days, and Oyster Creek continued power operation. Specifically, on October 9, 2017, No. 2 emergency diesel generator failed to run during a routine surveillance test due to a broken ring lug on a current transformer, which resulted in a total inoperability time of 6.5 months.Severity/Significance: For violations warranting enforcement discretion, Inspection Manual Chapter 0612 does not require a detailed risk evaluation, however, safety significance characterization is appropriate. A Region I Senior Reactor Analyst (SRA) performed a best estimate analysis of the safety significance using the Oyster Creek Standardized Plant Analysis Risk (SPAR) model, Version 8.50 and Systems Analysis Programs for Hands-On Integrated Reliability Evaluations (SAPHIRE). The evaluation estimated the total (internal and external events risk) increase in core damage frequency (CDF) to be in the mid to high E-6/yr range, or a low to moderate safety significance. The SRA evaluated the internal events risk contribution due to the inoperability of the No. 2 emergency diesel generator for an approximate 6.5 month exposure time. The exposure time relative to when the No. 2 emergency diesel generator was no longer capable of meeting its 24 hour mission time is uncertain due to the effect of vibration induced fatigue, and therefore the method prescribed within the RASP handbook guidance was used. 9 The analyst used the guidance in Section 2.5 of the Handbook, Revision 2.0, to estimate the exposure time of 6.5 months based on the cumulative 24 hour summation of the No. 2 emergency diesel generator surveillance test proven run time. This approach is appropriate for periodically operated components that degrade during operation (i.e. vibration induced fatigue only occurs while the emergency diesel generator is in-service/operating). Given this approach, the dominant internal events, loss of offsite power were evaluated for the estimated internal risk increase. This contribution was estimated at 2E-6/yr increase in CDF. The dominant sequences involved loss of offsite power events with a concurrent failure of the No. 1 emergency diesel generator, failure of the combustion turbines, and failure to recover offsite power or recover an emergency diesel generator prior to core damage.The SRA performed various modeling changes after a review of revised calculations for DC battery life:Analysis noted that Oyster Creek Generating Station recirculation pump seals are similar in design to those tested in reports generated for Nine Mile Point Unit 1 with the use of CAN2A seals. Therefore, the failure probability of the seals in the station blackout sequence wasadjusted from 0.1 to 5E-2 similar to Nine Mile Point Unit 1 SPAR model 8.50.The failure to load shed action (DCP-XHE-XM-LSHED) in the model was calculated using the SPAR-H method and revised to 1.2E-2 versus being assumed to always fail (TRUE).Failure probabilities for 1, 2, or 3 stuck open electromatic relief valves were revised to be consistent with the previous model version 8.22 because of the isolation condenser design at Oyster Creek Generating Station which limits cycling and significantly reduces the probability of a failed open electromatic relief valve due to isolation condensers controlling pressure.The depressurization function using electromatic relief valves, if required, was calculated through SPAR-H to be 1E-2 for sequences where total seal failure is assumed (DEPSEALFAIL) (conservatively assumed limited time available).The diesel driven firewater pumps are both available and were set to calculated fault tree failure probabilities instead of always failed in the previous model. These are 2,000 gallons per minute pumps with a large supply of water and relatively simple operator actions to inject to the reactor pressure vessel. The firewater was assumed to fail at 0.1 when a total recirculation seal failure occurs due to assumed time constraints.The offsite power and the emergency diesel generator required recovery time events were increased to 24 hours for events where DC load shedding was successful, without seal failures and isolation condenser success along with diesel driven firewater success.The SRA noted the No. 2 emergency diesel generator was recoverable. In fact, the diagnosis of the failed condition was performed in a nominal 8-10 hours from the failure. Therefore, a probability of failure to recover event for the conditional case was developed. The SRA used SPAR-H as simple guidance, which conservatively supported a reasonable assumption of a 0.10 conditional probability of failure to recover the emergency diesel generator within 24 hours. The base case utilized a calculation within SPAR of 0.33 failure to recover probability for 24 hour sequences. To estimate the external risk contribution, the SRA identified that the most significant external risk contribution was from fire events. Seismic, external flooding, and high wind events were not significant contributors for the issue. From discussions with Oyster Creek Fire probabilistic risk analysts and a review of this failure condition, the increase in CDF due to the failed No. 2 emergency diesel generator for the assumed 6.5 month exposure time was estimated at 4.5E-6/yr ((8.5E-5/yr-4.5E-5/yr) x (6.5/12 months) x 0.2).The DC safety-related battery life would be at least a nominal 14 hours and longer if DC bus stripping occurred, this allows for extended isolation condenser or electromatic relief valve function, with injection from diesel driven firewater. Given the time considerations and characteristics of the failure, an assumed recovery at a failure probability of 0.2 (slightly higher than internal due to less time) was applied for the No. 2 emergency diesel generator, which was a best estimate determined through SPAR-H insights. The dominant fire sequence was a fire affecting the A and B 4kV switchgear rooms, where combustion turbine support would be lost, with failure of the No. 1 emergency diesel generator breaker to close, and failure of locally operating the isolation condenser due to eventual loss of power. The SRA noted that FLEX credit was not quantified and would result in a lower risk estimation likely in the low E-6/yr range. Combining internal and external risk contributions, the total increase in CDF was 6.5E-6/yr, or low to moderate safety significance. The SRA determined that Exelon uses a Large Early Release Frequency (LERF) factor value of 8E-2. This value takes into consideration operator action for those relevant high pressure vessel breach scenarios (fuel-coolant interaction, liner-melt-through, and direct containment heating). This also credits procedure strategies where other mitigating actions are taken such as flooding the drywell. The SRA review of the dominant sequences and time to core damage affirmed that LERF did not increase the risk over that determined from the increase in CDF.Basis for Discretion: The inspectors determined that the ring lug failure was not within Exelons ability to foresee and prevent. As a result, no performance deficiency was identified. The inspectors assessment considered:1. Exelons review of emergency diesel maintenance performed in 2015 checked allconnections of the current transformer for tightness. The inspectors did not identify any gaps or deficiencies in the 2015 inspections. Inspectors also reviewed completed biennial inspections of the connection dating back to 1991 and did not identify any gaps.2. At the time of the failure, the current transformer connections did not have a time directed replacement frequency recommended by the Emergency Diesel Generator Owners Group. The inspectors did not identify any additional vendor or industry recommendations specific to the failed component or considerations specific to the failed component that existed prior to the failure.3. Industry operating experience information available to Exelon did not identify the potential for the fatigue cracking of the bent wire ring lug that was experienced.4. The bent ring lug failure was not the result of a failure on the part of Exelon staff; no standards existed on bending of the lug during installation and is considered skill of the craft.The NRC determined that it was not reasonable for Exelon to have been able to foresee and prevent this violation of NRC requirements, and as such, no performance deficiency existed. Therefore, the NRC has decided to exercise enforcement discretion in accordance with Sections 2.2.4 and 3.10 of the NRC Enforcement Policy and refrain from issuing enforcement action for the violation of technical specifications (EA-18-007). Further, because Exelons actions did not contribute to this violation, it will not be considered in the assessment process or the NRC Action Matrix. Exelons equipment corrective action program evaluation report (ECAPE) determined that the ring lug failed on the No. 2 emergency diesel generator as a result of fatigue cracking, which was initiated due to excessive stress caused by bending and twisting of the ring lug beyond limits specified in industry guidelines. The inspectors noted that the ECAPE did not provide supporting information regarding how the ring lug was bent and twisted beyond industry guidelines. Specifically, industry guidance states that ring lugs can be bent up to 90 degrees. The broken ring lug found in the No. 2 emergency diesel generator was bent at approximately 45-55 degrees per the ECAPE, which was within industry guidelines. Additionally, the ECAPE did not include specific guidance on twisting allowances for ring lugs. Exelon documented the inspectors observation in Issue Report 4089829. As a result of the inspectors observation, Exelon revised the ECAPE to say the ring lug failed on the No. 2 emergency diesel generator as a result of fatigue cracking, which was initiated due to excessive stress caused by bending and twisting of the ring lug.
05000220/FIN-2018001-01 31 March 2018 23:59:59 Nine Mile Point Potential Failure to Submit an 8-Hour Event Notification for a Valid Actuation of HPC On March 18, 2018,at 1:18 a.m., during the Unit 1maintenance outage while the unit was in cold shutdown, operators received multiple low level alarms on the GEMAC 11 and 12 level indications. Operators responded by adjusting reactor water cleanup reject flow and the feedwater minimum flow control valve to raise reactor water level. Upon the operators making the adjustment to reactor water level, the feedwater low flow control valve was slow to respond, but eventually opened more rapidly, and the increased flow from feedwater resulted in a rapid rise in reactor water level. At 1:28 a.m., indicated reactor water level rose to the 95-inch trip setpoint for the 11 and 12 Yarway level indication instruments, resulting in a turbine trip and HPCI initiation signal. The HPCI pumps were tagged out and thus did not inject, and the turbine was offline for the shutdown. The 11 and 12 Yarway level indication instruments provide reactor protection system logic inputs for reactor vessel water level; however, the Yarway level indication instruments are not density compensated. Therefore, under cold shutdown conditions, actual reactor vessel water level was lower than indicated water level on the 11 and 12 Yarways. During cold shutdown conditions, the GEMAC level instruments, which are calibrated to cold shutdown conditions, provide an accurate indication of actual reactor vessel water level. The GEMAC instruments both indicated well below the trip setpoint of 95 inches (indicated ~72 inches) when the turbine trip and HPCI initiation signal were received. Exelon determined that this event was not reportable under 10 CFR 50.72.Title 10 CFR 50.72(b)(3)(iv)(A) states, Any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B) of this section, except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation. (B) The systems to which the requirements of paragraph (b)(3)(iv)(A) of this section apply are: 10 (5) BWR reactor core isolation cooling system; isolation condenser system; and feedwater coolant injection system. Planned Closure Action(s): The inspectors requested the 10 CFR 50.72 subject matter experts at the Office of Nuclear Reactor Regulation (NRR) and Office of General Council (OGC) to review whether this was a valid actuation and thus reportable. The inspectors are opening an unresolved item (URI) to determine if a performance deficiency exists.Licensee Action(s): Licensee entered the concern into their corrective action program, and communicated with NRC Region I and NRR Staff. Exelons position is that the event was not reportable. Corrective Action Reference:IR 04116336 NRC Tracking Number: 05000220/2018001-01
05000410/FIN-2017004-04 31 December 2017 23:59:59 Nine Mile Point Ineffective Correction Action Results in Failure of Instrument Air System The inspectors documented a self-revealing Green finding (FIN) of CNG-CA-1.01-1000, Corrective Action Program, Revision 01100, because Nine Mile Point Nuclear Station (NMPNS) failed to implement corrective actions at NMPNS Unit 2 to remove and replace all un-annealed red brass piping for the instrument air system during the April 2008 refueling outage. Specifically, on July 13, 2017, Unit 2 experienced a rupture of un-annealed red brass instrument air pipe which resulted in a feedwater pump trip and a reactor recirculation pump runback to 49 percent. Exelons corrective actions for the July 13, 2017 failure of un-annealed red brass instrument air piping included wrapping the instrument air piping with a material that both supports the piping and prevents potential stress corrosion cracking. Exelon has developed work orders to replace the piping in the upcoming outage in spring 2018. Exelon also improved staff training for accountability and work checking to verify that generated work orders are completed and closed out. Exelon entered this issue into the corrective action program (CAP) as issue report (IR) 04031685, and performed a corrective action program evaluation (CAPE). This finding is more than minor because it is associated with the design control attribute of the Initiating Events cornerstone and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. Specifically, NMPNS staff failed to complete corrective actions to replace Unit 2 un-annealed red brass instrument air piping, which was susceptible to stress corrosion cracking, resulting in a feedwater pump trip and a reactor recirculation runback to 49 percent on July 13, 2017. The inspectors evaluated the finding using IMC 0609, Attachment 4, Initial Characterization of Findings, issued on October 7, 2016, and Exhibit 1 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2012. The inspectors determined that the finding was of very low safety significance (Green) because it did not result in the complete or partial loss of a support system that contributes to the likelihood of, or cause, an initiating event and affected mitigation equipment. The inspectors determined that this finding did not have a cross-cutting aspect because the performance deficiency occurred greater than 3 years ago; therefore, it is not considered to be indicative of current plant performance.
05000410/FIN-2017004-03 31 December 2017 23:59:59 Nine Mile Point Inadequate Operability Determination forImpairedInternal Flood Barrier An NRC-identified Green NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified when Exelon failed to perform an adequate operability determination in accordance with OP-AA-108-115, Operability Determinations, Revision 20, upon identification of Unit 2 degraded internal flood barriers that support operability of emergency core cooling system (ECCS) equipment. Specifically, from November 21 until December 10, 2017, Exelon failed to properly evaluate the excavation of internal flood barriers and concluded there was a reasonable expectation for operability of the supported ECCS systems. Exelon entered this issue into the CAP as IR 04082686. Corrective actions included conducting a detailed evaluation of operability for the supported safety-related systems, additional training associated with TS 3.0.9, including a focus on the need for risk assessments when entering TS 3.0.9, and a procedure change to CC-AA-201, Plant Barrier Control Program, and CC-NM-201-1001, Plant Barrier Control Program Implementation, which is the NMPNS specific procedure to address the vulnerabilities associated with impairing multiple required barriers. This finding is more than minor because it is associated with the human performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, from November 21 until December 10, 2017, Exelon failed to adequately evaluate the operability of a degraded internal flooding barrier and the potential impact on operability of the supported ECCS system equipment. The inspectors identified that the internal flood barrier was excavated such that there was not sufficient material to ensure adequate flood protection, and resulted in a reasonable doubt for the operability of the supported ECCS systems. This finding is also similar to example 3.j and 3.k of IMC 0612 Appendix E, Examples of Minor Issues, issued August 11, 2009, because the condition identified by the inspectors resulted in a reasonable doubt for the operability of the ECCS supported systems and additional analysis was necessary to verify operability. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2012, the inspectors determined that the finding was of very low safety significance (Green), because the finding was not a design or qualification deficiency, did not represent a loss of system safety function, and did not screen as potentially risk significant due to vulnerability to external initiating events. This finding has a cross-cutting aspect in the area of Human Performance, Work Management, because Exelon failed to implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. As a result, Exelon personnel failed to recognize that work activities that impaired internal flood barriers on both Division I and II low pressure ECCS pump rooms were executed simultaneously, which led to an unplanned entry into TS Limiting Condition for Operation (LCO) 3.0.9. (H.5)
05000220/FIN-2017004-02 31 December 2017 23:59:59 Nine Mile Point Inadequate Fill and VentProcedure for Control Room Chiller Results in Unplanned LCO Entry An NRC-identified Green NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified for Exelons failure to ensure that activities affecting quality were prescribed in a manner appropriate to the circumstances for the Unit 1 control room chiller system. Specifically, Exelon procedure N1-OP-49, Control Room Ventilation System, Revision 03800, Section H.5, Venting of Control Room Chiller Circulating Water Pump 11 and 12 Discharge Piping, led personnel to inadequately fill and vent the 12 control room chiller during system restoration from maintenance, while in a single chiller lineup. As a result, on October 15, 2017, control room chiller 12 tripped on low flow, and due to a prior trip of 11A control room chiller compressor, an unplanned 7-day LCO in accordance with TS 3.4.5.e, Control Room Air Treatment System, was entered due to an insufficient number of available chiller compressors to provide adequate control room cooling. Exelon entered this issue into the CAP as IR 04090200. Corrective actions included generating a procedure change to correct N1-OP-49 Section H.5, which provides instruction for filling and venting when in a single chiller lineup This finding is more than minor because it is associated with the procedure quality attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, Exelon failed to prescribe an adequate fill and vent procedure for the Unit 1 control room chillers which led to a trip of the 12 chiller on low flow while troubleshooting of chiller compressor 11A was on-going, resulting in an unplanned TS LCO entry. The inspectors evaluated this finding using IMC 0609.04, Initial Characterization of Findings, and Exhibit 3 of IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The performance deficiency did not represent a degradation of the radiological barrier function provided for the control room. Additionally, the performance deficiency did not represent a degradation of the barrier function of the control room against smoke or a toxic atmosphere. Therefore, this finding was determined to be of very low safety significance (Green). This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, because between 2014 and 2017 the inspectors noted over 20 issue reports documenting issues affecting reliability of the control room chiller system. Exelon failed to thoroughly evaluate the issues associated with the chillers to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, Exelon failed to effectively evaluate previous chiller trips and to prevent additional trips of the chiller system such as the one that occurred on October 15, 2017. (P.2) (Section 1R12.b.2)
05000220/FIN-2017004-01 31 December 2017 23:59:59 Nine Mile Point Main Control Room Annunciators 10 CFR 50.65(a)(2) Demonstration Not Met An NRC-identified Green non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) 50.65 (a)(2), was identified because Exelon did not adequately demonstrate that the performance of the Unit 1 main control room (MCR) annunciators was effectively controlled through performance of appropriate preventive maintenance. Specifically, Exelon did not identify and properly account for functional failures of the MCR annunciators in June 2017, and therefore did not recognize that the annunciator system exceeded its performance criteria and required a Maintenance Rule (a)(1) evaluation. On December 7, 2017, Exelon entered the issue into their CAP as IR 04081698 and performed a review of the events identified by the inspectors that were applicable to the maintenance rule annunciator system. Corrective actions included Exelon determining that the events were functional failures, and initiated an (a)(1) evaluation based on the MCR annunciator system functional failures exceeding the designated performance criteria of an allowable one functional failure per 24 months.This finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and affects the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Specifically, following the two failures of the main control annunciator panel in June 2017, Exelon did not identify the failures as functional failures, and consequently, did not establish goals and monitoring criteria in accordance with 10 CFR 50.65(a)(1). In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The SDP for Findings At-Power, issued June 19, 2012, the inspectors determined that the finding was of very low safety significance (Green) because the finding did not affect the design or qualification of a mitigating structure, system, or component (SSC), did not represent a loss of system and/or function, did not involve an actual loss of a function of at least a single train or two separate safety systems for a greater time than allowed by technical specifications (TS), and did not represent an actual loss of function of one or more non-TS trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program. T his finding has a cross-cutting aspect in the area of Human Performance, Consistent Process, in that Exelon failed to use a consistent, systematic approach to make decisions. Specifically, Exelon did not ensure their review process for issues entered into the CAP was effectively implemented to ensure proper evaluations for all applicable maintenance rule systems affected by a n SSC failure. (H.13)
05000219/FIN-2017003-01 30 September 2017 23:59:59 Oyster Creek Inadequate Augmented Offgas System Procedure Resulted in a Manual Scram A self -revealing NCV of Technical Specification 6.8.1, Procedures and Programs, was identified because Exelon did not adequately establish and maintain the augmented offgas (AOG) system operation procedure as required by NRC Regulatory Guide 1.33, Quality Assurance Requirements (Operation), Appendix A, Section 7, Procedures for Control of Radioactivity. Specifically, Exelon procedure 350.1, Augmented Offgas System Operation, did not include adequate guidance for placing the AOG system into a recycle or shutdown configuration following a system trip. Without this guidance, Operations personnel failed to ensure the correct configuration of the AOG system following a partial trip of the system which resulted in degraded main condenser vacuum and a subsequent manual reactor scram on July 3, 2017. This issue was entered into the corrective action program as issue report 4028402. The corrective actions included placing the AOG system in the correct configuration and revising the AOG system operation procedure to provide guidance for verifying proper alignment of the AOG system when the system is in recycle or shutdown. The inspectors determined the performance deficiency was more than minor because it was associated with the Initiating Events cornerstone attribute of Procedure Quality and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the failure to establish an adequate procedure for verifying proper alignment of the AOG system following a full or partial trip of the system resulted in the AOG inlet valve being left in the open position, which allowed demineralized water to be siphoned from the flame arrestor tank and slowly fill the offgas hold- up pipe. This caused a degradation of main condenser vacuum and resulted in operators inserting a manual reactor scram on July 3, 2017. The inspectors evaluated the finding using IMC 0609, Attachment 4, Initial Screening and Characterization of Findings, and IMC 0609, Appendix A, Exhibit 1, Initiating Event Screening Questions. The inspectors determined the finding was a transient initiator that did not contribute to both the likelihood of a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of a trip to a stable shutdown condition, and therefore was of very low safety significance (Green). The finding had a cross- cutting aspect in the area of Human Performance, Avoid Complacency , because Exelon failed to recognize and plan for the possibility of mistakes or latent errors and implement appropriate error reduction tools by verifying the AOG system was properly aligned following a system trip ; instead , Operations personnel relied upon using a procedure that did not contain adequate guidance to place the AOG system in the correct configuration following a system trip (H. 12)
05000219/FIN-2017404-01 30 September 2017 23:59:59 Oyster Creek Security
05000219/FIN-2017002-01 30 June 2017 23:59:59 Oyster Creek Inadequate Assessment of Degraded Fuel Oil Filter Impact to Emergency Diesel Generator Operability The inspectors identified a finding associated with Exelon procedure OP-AA-108-115, Operability Determinations, because Exelon did not adequately assess the No. 2 emergency diesel generator operability with a degraded fuel oil filter. Specifically, Exelon did not adequately assess the capability of the emergency diesel generator to perform its function during its credited duration time of 72 hours. Exelon entered this issue into the corrective action program for resolution as issue report (IR) 3999576 and IR 3990799 and subsequently replaced the fuel oil filter. The finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone objective to ensure the reliability and capability of systems that respond to initiating events to prevent undesirable consequences. This issue was also similar to Example 3j of IMC 0612, Appendix E, Examples of Minor Issues, because the condition resulted in reasonable doubt of the operability of the No. 2 emergency diesel generator and additional analysis was necessary to verify operability. The inspectors evaluated the finding using Exhibit 2, Mitigating System Screening Questions, in Appendix A to IMC 0609, Significance Determination Process. The inspectors determined that this finding was a deficiency affecting the design or qualification of a mitigating structure, system, or component (SSC), where the SSC maintained its operability or functionality. Therefore, inspectors determined the finding to be of very low safety significance (Green). The finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, because Exelon did not thoroughly evaluate the issue associated with the degraded fuel oil filter and its impact to the No. 2 emergency diesel generator operability (P.2).
05000410/FIN-2017002-01 30 June 2017 23:59:59 Nine Mile Point Inadequate Extent of Condition Results inUnplanned Yellow Risk Condition The inspectors identified a Green non-cited violation (NCV) of Title 10 of the Code of Federal Regulation (10 CFR) 50.65(a)(4), Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, when Exelon did not assess and manage the increase in risk for online maintenance activities. Specifically, on May 24, 2017, the inspectors identified a planned surveillance activity which caused unavailability of the A residual heat removal (RHR) system minimum flow valve that was not recognized by the Exelon staff as a, Yellow, elevated risk activity in accordance with their EOOS (Equipment Out of Service) probabilistic risk assessment (PRA) model. Exelon staff generated issue report (IR) 04015294 to address the failure to recognize the Yellow, elevated risk activity and failure to review adequate extent of condition. Corrective actions include evaluating PRA to assess if risk can be reduced to Green with compensatory actions and providing training to operations to enhance PRA modeling of system availability. Following review of the PRA model, Exelon plans to evaluate all surveillance procedures as part of extent of condition that could impact availability of the A RHR minimum flow valves.This performance deficiency is more than minor because it affected the human performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, on May 24, 2017, the inspectors identified a planned activity that resulted in an unplanned Yellow risk activity during planned maintenance that resulted in unavailability of a component used to support the A RHR system. Additionally, this issue is similar to Example 7.f of IMC 0612, Appendix E, Examples of Minor Issues, issued August 11, 2009, because the overall elevated plant risk placed the plant into a higher licensee-established risk category. The inspectors determined that this finding is of very low safety significance (Green). Because the incremental core damage probability deficit was less than 1E-6 and the incremental large early release probability was less than 1E-7, this finding was determined to be of very low safety significance (Green). The cause of the finding has a cross-cutting aspect in the area of Human Performance, Teamwork, because Exelon staff did not effectively communicate internally to ensure that corrective actions were being addressed to resolve concerns with risk associated with A RHR minimum flow valve availability. Specifically, Exelon staff incorrectly believed that integrated risk management guidance corresponded to PRA availability. Thus, it was assumed risk would remain Green during surveillance and maintenance activities that resulted in the A RHR minimum flow valve being unavailable; and a failure to recognize future maintenance activities that resulted in risk being Yellow. (H.4)
05000220/FIN-2017001-01 31 March 2017 23:59:59 Nine Mile Point Deficient Design Control of Outboard MSIV Pilot Valve Instrument Air Supply Green. The inspectors identified a Green NCV of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion III, Design Control, for Exelons failure to correctly translate the design basis into the NMPNS Unit 1 instrument air system to ensure the Unit 1 outboard main steam isolation valves (MSIVs) were capable of performing their design function. Specifically, the NMPNS Unit 1 Updated Final Safety Analysis Report (UFSAR) states, Reliable operation of instrument air end users and in-line components is dependent on the filtration and removal of particulates greater than 40 microns. Additional filtration for various components exists where the 40 micron limit is not satisfactory. The MSIV pilot valves at Unit 1 have a tighter clearance than the 40 micron limit. However, contrary to the UFSAR, NMPNS did not install additional filtration upstream of the pilot valves. As a result, during a surveillance test conducted on December 10, 2016, foreign material in the instrument air system potentially contributed to the failure of an outboard MSIV. Exelons immediate corrective actions included entering this issue into its corrective action program (CAP) as issue report (IR) 03959732, performing an air purge of the instrument air system to remove foreign material from the system, and replacing the current style pilot valves with new style valves with larger clearances during the spring 2017 refueling outage. The performance deficiency was determined to be more than minor because it was associated with the design control attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents for events. Specifically, Exelon failed to install additional filtration in the instrument air system upstream of the outboard MSIV pilot valve in accordance with the Unit 1 UFSAR even though the internal clearance of the pilot valve was significantly less than the 40 micron particulate limit. Additionally, example 3.j from IMC 0612, Appendix E, Examples of Minor Issues, provides a similar scenario to this issue. Example 3.j details that a performance deficiency is more than minor if the error results in a condition where there is a reasonable doubt of the operability of a system or component. This performance deficiency is more than minor because without the additional filtration defined in the UFSAR there 4 existed a reasonable doubt of operability for the Unit 1 outboard MSIVs. The finding was evaluated in accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 3 of IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, and determined to be of very low safety significance (Green). The finding has a cross-cutting aspect in the area of Human Performance, Documentation, because Exelon failed to create and maintain complete, accurate, and up-to-date documentation pertaining to instrument air sampling for high particulate. Specifically, Exelon failed to develop and implement a surveillance testing program for the instrument air system that would alert personnel that particulate greater than 5 microns could jeopardize the operability of the outboard MSIVs. (H.7)
05000220/FIN-2017001-02 31 March 2017 23:59:59 Nine Mile Point Failure to Identify and Correct a Non- Conforming Condition in Safety-Related UPSs Green. The inspectors documented a self-revealing Green NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions, for the failure to identify and correct a non-conformance (an inadequate capacitor) in safety-related uninterruptable power supplies (UPSs) 162 and 172. Between 2008 and 2017, this non-conformance led to multiple component failures, loss of vital power supplies, plant transients, and in one case, loss of the emergency condenser safety function. Specifically, in 2003, during a preventative maintenance activity, NMPNS installed a commercially dedicated capacitor (part number C-805) that was not rated for the normal service temperature for the application. This resulted in chronic overheating, reduction of service life, and in seven cases failures (internal shorts of C-805) which resulted in the loss of the associated safety-related UPS. Upon identification, Exelon entered each failure into the CAP conducted an apparent cause evaluation (ACE) following the 2016 and 2017 failures, and developed corrective actions to replace the underrated capacitors. The performance deficiency was determined to be more than minor because it affected the equipment performance attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge the critical safety functions during shutdown as well as power operations. Specifically, the underrated capacitors failure resulted in the loss of a vital alternating current (AC) bus, a support system and in one case the unplanned loss of a safety function required to bring and maintain the plant in safe shutdown. In accordance with IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, a detailed risk assessment was required. Using the NMPNS Unit 1 Standardized Plant Analysis Risk (SPAR) Model Version: 8.21, model date January 28, 2010, a Region I senior reactor analyst ran a zero maintenance condition assessment with basic events for emergency condenser (EC) motor operated valve (MOV) 39-09R and EC MOV 39-10R, normally closed condensate return isolation valves, failed for a duration of one hour. The results were a CDP of 1.37E-08. The dominant risk sequences involved loss of feedwater and loss of offsite power. As a result, the finding is of very low safety significance (Green). The performance deficiency for this finding occurred in 2008. Because the performance deficiency occurred greater than 3 years ago and is not indicative of current performance based upon the corrective actions taken following the 2016 failure, there is no cross-cutting aspect assigned to this finding.
05000219/FIN-2016004-01 31 December 2016 23:59:59 Oyster Creek E EMRV Failureto Stroke Due to Incorrect Reassembly The NRC identified a preliminary White finding and associated apparent violation of Technical Specification 6.8.1, Procedures and Programs, and Technical Specification 3.4.B, Automatic Depressurization System, because Exelon failed to implement a procedure related to the maintenance of safety related equipment. Specifically, Exelon personnel did not follow electromatic relief valve (EMRV) reassembly instructions that required personnel to reinstall previously removed lock washers from the E EMRV cut-out switch lever. The incorrect reassembly caused excessive friction between the solenoid frame and the cut-out switch lever, which led to the E EMRVs failure to perform its safety function. This resulted in one inoperable EMRV for greater than the Technical Specification allowed outage time. The issue was entered into the corrective action program as issue report 2722109, and Exelons immediate corrective actions include installing new cut-out switch lever plates with increased clearances, replacing star lock washers with split ring lock washers for additional clearance, and verifying the five EMRV solenoid actuators being installed into the drywell following the most recent refueling outage were correctly assembled. The finding is more than minor because it adversely affects the human performance quality attribute of the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the missing lock washers due to the incorrect EMRV lever plate reassembly caused excessive friction between the solenoid frame and the cut-out switch lever, causing the cut-out switch lever to become bound in the energized position. This led to the E EMRVs failure to perform its safety function. The inspectors screened this issue for safety significance in accordance with Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, and determined a detailed risk evaluation was required because the E EMRV had potentially failed or was unreliable for greater than the Technical Specification allowed outage time. A detailed risk evaluation concluded that the increase in core damage frequency (CDF) related to the failure of the E EMRV is 5.4E-6/year; therefore, this finding was preliminary determined to have a low to moderate safety significance (White). Due to the nature of the failure, no recovery credit was assigned. The dominant core damage sequences involve loss of main feedwater events with operator errors resulting in failure to make-up to the 4 isolation condensers or otherwise maintain reactor vessel level and the loss of reactor pressure vessel depressurization capability (due to common cause failure of the remaining four EMRVs). The finding has a cross-cutting aspect in the area of Human Performance, Procedure Adherence, because Exelon personnel did not follow station processes. Specifically, Exelon did not follow written instructions when reassembling the E EMRV. The missing lock washers resulted in excessive friction between the solenoid frame and cut-out switch lever, causing the cut-out switch lever to become bound in the energized position, which led to the E EMRVs failure to perform its safety function. (H.8)
05000220/FIN-2016404-01 31 December 2016 23:59:59 Nine Mile Point Security
05000410/FIN-2016002-03 30 June 2016 23:59:59 Nine Mile Point Failure to Understand Radiological Conditions Results in Unintended Exposure A self-revealing NCV of TS 5.4.1 Procedures was identified when a worker performed a radiological work activity without notifying radiation protection personnel and, as a result, did not comply with procedure RP-AA-1008, Unescorted Access to and Conduct in Radiologically Controlled Areas, Revision 5, in being briefed on the necessary radiological work controls and conditions for performance of the Unit 2 reactor seal cleaning work activity. Specifically, on April 11, 2016, a worker entered the Unit 2 reactor cavity to perform inspection of the reactor seal that was highly contaminated. Although not previously discussed with radiation protection staff, the worker cleaned the highly contaminated reactor seal with rags and carried the highly contaminated rags (5 rem/hr) in his hand out of the reactor cavity, which resulted in unplanned radiation exposure to the workers hand. Exelons immediate corrective actions included reinforcing the need to properly communicate radiological work activities with radiation protection, and require workers to carry WOs with them to improve communications with radiation protection. Exelon entered the issue into the corrective action program (CAP) as IR 02654591. The failure of the worker to discuss the full scope of the radiological work activity with radiation protection staff, who were subsequently not effectively briefed on the expected radiological work conditions and requisite radiological controls needed for the work activity, is a performance deficiency that was reasonably within Exelons ability to foresee and correct. The finding was determined to be more than minor because it affected the human performance attribute of the Occupational Radiation Safety cornerstone objective. Using IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, the finding was determined to be of very low safety significance (Green) because it did not involve: (1) as low as 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. The finding is self-revealing because Exelon was made aware of the situation when an air monitor alarmed. The finding had a cross-cutting aspect of Human Performance, Team Work, since individuals and work groups did not communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety was maintained. Specifically, the worker did not adequately communicate to radiation protection staff, the reactor seal cleaning activity to be performed. As a result, radiation protection personnel did not prescribe sufficient radiological controls for this high-contamination work activity, and led to an unintended exposure to the workers hand.
05000220/FIN-2016403-01 30 June 2016 23:59:59 Nine Mile Point Security
05000410/FIN-2016002-01 30 June 2016 23:59:59 Nine Mile Point Ineffective Corrective Action Results in Water Intrusion to Battery Switchgear Room The inspectors identified a Green finding (FIN) of PI-AA-125, Corrective Action Program, Revision 3, when Exelon failed to implement adequate corrective actions in March 2003, to prevent water intrusion into the Unit 2 normal switchgear building area at elevation 237. Specifically, on May 4, 2016, the inspectors observed water leaking into the normal switchgear room through a wall on elevation 237. The leakage was through a section of the wall that contained electrical junction boxes that were not sealed. The water progressed under inverter 2BYS-SWG001B, which led to the potential for a reactor scram from an electrical fault associated with uninterruptible power supply battery breakers. Previously, a reactor scram had occurred at Unit 2 on March 4, 2014, when the inverter was lost because of an electrical fault, as such this was a known initiating event single point vulnerability . Corrective actions included entering the issue into the corrective action program (CAP) (IR 02664534), generating work order (WO) C93414574 to seal or repair the wall, and installing temporary barriers to redirect any water away from the switchboard. The WO is scheduled to be performed in October 2016 with an action to assess moving the work to the refueling outage if needed to remove the electrical junction boxes to apply coating to the wall. The finding is more than minor because it is associated with the Protection Against External Factors attribute of the Initiating Events cornerstone and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. Specifically, Exelon did not ensure the surface area behind the electrical junction boxes was coated to prevent water intrusion into the normal switchgear room at elevation 237. The water intrusion through this area of the wall had the potential to cause an electrical fault on 2BYS-SWG001B resulting in a reactor scram similar to the reactor scram in March 2014. The inspectors evaluated this finding using IMC 0609.04, Initial Characterization of Findings, and Exhibit 1 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power. The inspectors determined that this finding was of very low safety significance (Green) because it did not represent the potential for both a reactor scram and a loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The inspectors did not assign a cross-cutting aspect to this finding because the performance deficiency occurred greater than three years ago; therefore, it is not considered to be indicative of current plant performance.
05000410/FIN-2016002-02 30 June 2016 23:59:59 Nine Mile Point Failure to Identify Wide Range Level Indication Impacts Operability of HPCS and RCIC The inspectors identified a Green NCV of Unit 2 Technical Specification (TS) 3.5.1, Emergency Core Cooling (ECCS) Systems-Operating, and TS 3.5.3, Reactor Core Isolation Cooling (RCIC) System, for failure to ensure all necessary attendant instrumentation required for the systems to perform their specified safety functions were capable of performing their related support function in all require modes of applicability. Specifically, the inspectors identified the Unit 2 wide range level indication to be inaccurate during Mode 2 and at 200 pounds per square inch gauge (psig) reactor pressure, a mode of applicability requiring both high-pressure core spray (HPCS) and RCIC to be operable. This resulted in a high level trip signal being locked preventing HPCS or RCIC from auto initiating, rendering the systems inoperable. Upon identification, Exelon generated issue report (IR) 02667837 to address the inspectors concern regarding the wide range level indication. An action was created to evaluate the impact of the wide range level discrepancy with regard to its impact on safety-related functions to supply water in the TS Mode of Applicability. Exelon also plans to assess the need for a TS amendment. The performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, Exelon failed to recognize that the wide range level indication did not provide accurate indication at low reactor pressures and temperatures, preventing automatic safety-related functions associated with high drywell pressure automatic initiation signals and manual start functions. This would require operators to manually open the HPCS and RCIC injection valves during these conditions should a loss of offsite power or loss-of-coolant accident occur. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, the inspectors determined that the finding was of very low safety significance (Green), because the finding was not a design or qualification deficiency, did not represent a loss of system safety function, and did not screen as potentially risk significant due to seismic, flooding, or severe weather initiating event. The finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Identification. Exelon personnel had many opportunities, including during the reactor startup in May of 2016, to question operability of the instrumentation that provides input for automatic initiation and isolation signals. As a result, Exelon personnel failed to identify that the wide range level indication did not support operability of the HPCS and RCIC systems during reactor startup on May 5, 2016. (P.1)
05000219/FIN-2016002-01 30 June 2016 23:59:59 Oyster Creek Inadequate Maintenance Procedure associated with Reactor Recirculation Pump Seal A self-revealing NCV of Technical Specification 6.8.1, Procedures and Programs, was identified because Exelon did not adequately establish and maintain the reactor recirculation pump (RRP) reassembly maintenance procedures as required by NRC Regulatory Guide 1.33, Appendix A, Section 9, Procedures for Performing Maintenance. Specifically, the RRP reassembly procedure, 2400-SMM-3226.03, Reactor Recirculation Pump Mechanical Seal Rebuild Using CAN-2A Parts, did not provide critical dimensional checks for the locking plate and seal adjusting cap. This led to the incorrect reassembly of the D RRP. Exelon entered this issue into their corrective action program as issue report 2663436. The corrective actions included repairing the D RRP and revising RRP maintenance procedures to include critical dimensional information. This finding is more than minor because it is associated with the procedure quality attribute of the Initiating Events cornerstone and affected the objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown and power operation. Specifically, the incorrect reassembly of the D RRP created a leakage path, which led to an unexpected increase in reactor coolant system (RCS) unidentified leakage. As a result, the operators inserted a manual scram on April 30, 2016. The inspectors evaluated the finding using IMC 0609, Attachment 4, Initial Screening and Characterization of Findings, and IMC 0609, Appendix A, Exhibit 1, Initiating Event Screening Questions. The inspectors determined that this finding is a transient initiator that did not contribute to both the likelihood of a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition, and therefore was of very low safety significance (Green). The inspectors determined that there was no cross-cutting aspect associated with this finding since it was not representative of current Exelon performance. Specifically, in accordance with IMC 0612, the causal factors associated with this finding occurred outside the nominal three-year period of consideration and were not considered representative of present performance.
05000410/FIN-2016001-01 31 March 2016 23:59:59 Nine Mile Point Inadequate Procedure Leading to Failure to Manage Elevated Risk during Preventive Maintenance The inspectors identified a non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) 50.65(a)(4), Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, when Exelon did not assess and manage the increase in risk for online maintenance activities. Specifically on February 12, 2016, Exelon did not assess and manage risk during Unit 2 planned testing associated with the A residual heat removal (RHR) system heat exchanger (HX). The inspectors identified that although the testing would render the A RHR minimum flow valve 2RHS*MOV4A unavailable, this was not considered as part of the planned maintenance window, which resulted in an increase in risk during the unavailability of 2RHS*MOV4A. When properly calculated, plant risk should have been indicated as Yellow for the day and not Green. Exelon generated issue report (IR) 02625546 to document the inspectors concern regarding the status of the availability associated with the A RHR minimum flow valve during test setup for the A RHR HX. Exelon corrective actions included evaluating the risk management activities to be implemented when the minimum flow valves are subject to maintenance or testing activities to ensure future work is properly screened. This finding is more than minor because it is associated with the configuration control attribute of the Mitigating Systems cornerstone and adversely affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, Exelons failure to plan for the unavailability of the A RHR minimum flow valve resulted in Unit 2 being placed in an unplanned elevated risk category (i.e., Yellow) without ensuring adequate compensatory measures were established and briefed to ensure maximum availability, reliability, and capability of the system. This issue is similar to Example 7.f of IMC 0612, Appendix E, Examples of Minor Issues, because the overall elevated plant risk placed the plant into a higher licensee-established risk category. The inspectors evaluated the finding using Phase 1, Initial Screening and Characterization worksheet in Attachment 4 and IMC 0609, Significance Determination Process. For findings within the Initiating Events, Mitigating Systems, and Barrier Integrity cornerstones, Attachment 4, Table 3, Paragraph 5.C, directs that if the finding affects the licensees assessment and management of risk associated with performing maintenance activities under all plant operating or shutdown conditions in accordance with Baseline Inspection Procedure 71111.13, Maintenance Risk Assessment and Emergent Work Control, the inspectors shall use IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, to determine the significance of the finding. The inspectors used Flowchart 1, Assessment of Risk Deficit, to analyze the finding and calculated incremental core damage probability using Equipment Out Of Service (EOOS), Exelons risk assessment tool. The inspectors determined that had this condition existed for the full duration of the Technical Specification (TS) limiting condition for operation (LCO), the incremental conditional core damage probability would have been 3.46E-9. Because the incremental core damage probability deficit was less than 1E-6 and the incremental large early release probability was less than 1E-7, this finding was determined to be of very low safety significance (Green). This finding has a cross-cutting aspect in the area of Human Performance, Work Management, because Exelon did not properly implement a process of planning, controlling, and executing the work activity such that nuclear safety was the overriding priority. Specifically, Exelon did not ensure risk was properly assessed during the planning process in accordance with WC-AA-101-1006, On-Line Risk Management and Assessment, Revision 001, prior to testing the A RHR HX, which caused unavailability of the A RHR minimum flow valve during certain periods of the test.
05000410/FIN-2016001-02 31 March 2016 23:59:59 Nine Mile Point 50.65(a)(4) Risk Evaluation Not Properly Performed Prior to Residual Heat Removal Heat Exchanger Testing The inspectors identified a Green non-cited (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) 50.65(a)(4), Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, for Exelons failure to take risk management actions (RMAs) as required by procedure OP-AA-108-117, Protected Equipment Program, Revision 004, during a Unit 2, Division III, emergency switchgear electrical maintenance window on January 27, 2016. Specifically contrary to procedure OP-AA-108-117, during planned maintenance, Exelon failed to post the unit coolers in the A and B RHR pump and HX rooms, the C RHR pump room, and their associated breakers as protected equipment although their inoperability would have resulted in both trains of the standby gas treatment system (SBGT) being inoperable which would require entry into Technical Specification (TS) Limiting Condition for Operation (LCO) 3.0.3 and a short term shutdown action statement. Upon identification, Exelon generated IR 02617915 to document this issue. Corrective actions included creating an action item to evaluate Attachment 3 of N2-OP-52 and to determine the relevance of the TS LCO 3.0.3 entry requirement. The inspectors determined the performance deficiency to be more than minor because it was associated with the structure, system, and component (SSC) and barrier performance attribute of the Barrier Integrity cornerstone and adversely affected the associated cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, contrary to OP-AA-108-117, Exelon personnel failed to include the unit coolers for the Unit 2 RHR pump and HX rooms and their associated breakers, whose unavailability would have resulted in the inoperability of both trains of SBGT and necessitated entry into LCO 3.0.3. Additionally, Examples 7.e, 7.f, and 7.g from IMC 0612, Appendix E, Examples of Minor Issues, provided similar scenarios to this issue. Example 7.e details that a performance deficiency is more than minor if a failure to include accurate TS requirements in a risk assessment and if done properly, would have required RMAs, or additional RMAs under applicable plant procedures. The inspectors evaluated the finding using Phase 1, Initial Screening and Characterization worksheet in Attachment 4 to IMC 0609, Significance Determination Process. For findings within the Initiating Events, Mitigating Systems, and Barrier Integrity cornerstones, Attachment 4, Table 3, Paragraph 5.C, directs that if the finding affects the licensees assessment and management of risk associated with performing maintenance activities under all plant operating or shutdown conditions in accordance with Baseline Inspection Procedure 71111.13, Maintenance Risk Assessment and Emergent Work Control, the inspectors shall use IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, to determine the significance of the finding. The inspectors used Flowchart 2, Assessment of RMAs, to analyze the finding and calculated incremental core damage probability using EOOS, Exelons risk assessment tool, and found the result to be less than 1E-6. The inspectors determined that had this condition existed for the full duration of the TS LCO, the incremental core damage probability would have been 6.8E-7. Because the incremental core damage probability deficit was less than 1E-6 and the incremental large early release probability was less than 1E-7, this finding was determined to be of very low safety significance (Green). This finding has a cross-cutting aspect in the area of Human Performance, Procedure Adherence, because Exelon failed to follow processes, procedures and work instructions. Specifically, Exelon failed to follow procedure OP-AA-108-117, which led to the failure to protect the unit coolers for the RHR pump rooms, HX rooms, and associated breakers which could have led to a TS LCO 3.0.3 entry.
05000219/FIN-2016001-01 31 March 2016 23:59:59 Oyster Creek Failure to Identify a Slower than Normal Scram Time of a Control Rod Drive The inspectors identified an NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, because Exelon did not promptly identify and correct a condition adverse to quality. Specifically, Exelon did not identify that the scram time test result for control rod drive 18-47 was beyond the analyzed scram time, which resulted in a degraded control rod drive. Exelon entered this issue into their corrective action program. Immediate corrective actions included fully inserting the control rod drive and developing a casual analysis to determine the degraded condition. The performance deficiency is more than minor because it is associated with the configuration control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the performance deficiency affected the reliability of control rod drive 18-47 to perform its safety function due to a slower than normal scram time. The inspectors evaluated the finding using IMC 0609, Attachment 4, Initial Screening and Characterization of Findings. The inspectors determined that this finding is a deficiency that affected the design or qualification of a mitigating structure, system, or component (SSC), when the SSC maintained its operability or functionality. Therefore, the inspectors determined the finding to be of very low safety significance (Green). The finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Identification, because Exelon did not identify issues completely, accurately, and in a timely manner in accordance with the program. Specifically, Exelon did not identify that the actual scram time of control rod drive 18-47 was beyond the analyzed scram time, resulting in a degraded control rod drive.
05000220/FIN-2016001-03 31 March 2016 23:59:59 Nine Mile Point Inadequate Tagout Resulting in Reactor Building Closed-Loop Cooling Drain Down Event A self-revealing Green non-cited violation (NCV) of Technical Specification (TS) 6.4.1, Procedures, was identified when a Unit 1 Exelon operator did not maintain proper configuration control of a plant system during a system tagout for planned maintenance. Specifically, on January 25, 2016, a Unit 1 non-licensed operator manipulated a reactor building closed-loop cooling (RBCLC) system drain valve out of sequence while performing a tagout for the #13 shutdown cooling (SDC) HX for planned maintenance. This resulted in unintentional draining of the operating RBCLC system, annunciation of multiple alarms in the main control room, and operators entering abnormal operating procedures to recover the RBCLC system. As part of corrective actions, proper configuration was promptly restored and the operator involved in the event was given a remediation plan for requalification and placed on an operations excellence plan. This finding is more than minor because it is associated with the configuration control attribute of the Mitigating Systems cornerstone and adversely affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences; and if left uncorrected, the event had potential to lead to a more significant safety concern. Specifically, the failure to quickly isolate the drain down of the RBCLC system would have required a manual reactor scram, a manual trip of all five reactor recirculation pumps (RRPs), a manual isolation of the reactor water cleanup system, a loss of cooling to the spent fuel pool (SFP) cooling system, instrument air compressors, and the control room emergency ventilation system. The inspectors evaluated the finding using IMC 0609.04, Initial Characterization of Findings, and Exhibit 1 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power. The inspectors determined that this finding was of very low safety significance (Green) because the performance deficiency did not result in the loss of a support system, RBCLC, or affect mitigation equipment. This finding has a cross-cutting aspect in the area of Human Performance, Procedure Adherence, because the non-licensed operator failed to follow Exelons procedures and the instructions he received at the pre job brief stop when manipulating the drain valve. Specifically, the non-licensed operator rationalized, without being the designated performer of the tagout, that it was acceptable to perform a valve manipulation out of sequence with the tagout plan.
05000219/FIN-2016001-03 31 March 2016 23:59:59 Oyster Creek Inadequate Instructions for the Flexible Coupling Hose Preventative Maintenance Resulting in an Inoperable Emergency Diesel Generator The inspectors identified a preliminary White finding and associated apparent violation of Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because Exelon did not appropriately prescribe instructions or procedures for maintenance on the emergency diesel generator (EDG) No. 1 cooling water system to ensure the EDG cooling flexible coupling hose was maintained to support the EDG safety function. Specifically, Exelon did not have appropriate work instructions to replace the EDG cooling flexible coupling hoses every 12 years as specified by Exelons procedure and vendor information. As a result, the flexible coupling hose was in service for approximately 22 years and subjected to thermal degradation and aging that eventually led to the failure of EDG No. 1 during operation on January 4, 2016. As a consequence of this inappropriate work instruction issue, Exelon violated Technical Specification 3.7.C because EDG No. 1 was determined to be inoperable for greater than the technical specification allowed outage time of seven days. Exelons immediate corrective actions included entering the issue into their corrective action program (issue reports 2607247 and 2610027), replacing of the EDG No. 1 and No. 2 flexible coupling hoses, and initiating a failure analysis to determine the causes of the failed flexible coupling hose. This finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the ruptured flexible coupling hose caused the failure of EDG No. 1 to perform its safety function. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, this finding required a detailed risk evaluation (DRE) because EDG No. 1 was inoperable for greater than the technical specification allowed outage time. The DRE estimated the increase in core damage frequency was 7E-6, or White (low to moderate safety significance) for this finding. This finding does not have an associated cross-cutting aspect because the performance deficiency occurred in 2005 and is not reflective of present performance.
05000219/FIN-2016001-02 31 March 2016 23:59:59 Oyster Creek Failure to Use Respiratory Protection as Required in RWP/ALARA Plan for Drywell Head Reassembly A self-revealing NCV of Technical Specification 6.8.1, Procedures and Programs was identified for Exelons failure to use respiratory protection, as required in the radiation work permit (RWP)/as low as reasonably achievable (ALARA) plan 14-406 for drywell head reassembly work on October 2, 2014. The radiation protection (RP) supervisor overseeing this work removed the respiratory protection requirement for this work contrary to the RWP/ALARA requirement and without engineering approval. As a result, two workers received an unplanned intake of radioactive material that resulted in unintended internal dose. Upon identification of the intake, Exelon stopped work on this task and subsequently reinstituted the respiratory protection requirements to complete the remaining work and entered this event into their corrective action program as issue report 2390111. This finding is more than minor because it is associated with the Occupational Radiation Safety cornerstone to ensure adequate protection of the worker from radiation exposure. Specifically, without the use of respiratory protection two workers received unintended internal dose. The inspectors evaluated the finding using inspection manual chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process. The inspectors determined that this finding is of very low safety significance (Green), because it did not result in an overexposure as defined by 10 CFR 20.1201, there was no substantial potential for an overexposure, and the ability to assess dose was not compromised. This finding has a cross-cutting aspect in Human Performance, Procedural Adherence, because Exelon did not follow procedures and work instructions. Specifically, RP supervision instructed the workers that respiratory protection was not required contrary to the applicable RWP/ALARA plan.
05000219/FIN-2015004-01 31 December 2015 23:59:59 Oyster Creek Preconditioning of the Standby Liquid Control Relief Valves The inspectors identified an NCV of 10 Code of Federal Regulations (CFR) 50, Appendix B, Criterion XI, Test Control, because Exelon conducted unacceptable preconditioning of the standby liquid control (SLC) relief valves prior to American Society of Mechanical Engineers (ASME) code testing. Specifically, Exelon performed a SLC system functional test prior to performing the SLC relief valve as-found testing. Exelons immediate corrective actions included completing the as-found test prior to the functional test. Exelon entered this issue into their corrective action program (CAP) as issue report 2566036 to track the resolution of the issue. The performance deficiency is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and affects the cornerstone objective of ensuring availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Additionally, if left uncorrected, the performance deficiency could have the potential to lead to a more significant safety concern. Specifically, completion of the functional test prior to the replacement of the SLC relief valves masks the actual as-found condition by solidifying the valve internals. As a result, the as-found condition of the SLC relief valves have not been conducted and in the worst case scenario, could open below the design setpoint, which would divert flow back to the liquid poison tank instead of into the vessel to shut down the reactor during an anticipated transient without scram (ATWS) condition. The inspectors evaluated the finding using IMC 0609, Attachment 4, Initial Screening and Characterization of Findings, and determined the finding was of very low safety significance (Green) because the structure, system or component (SSC) maintained its operability. The finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation because Exelon did not thoroughly evaluate issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, Exelon did not evaluate the effect of performing the SLC system functional test prior to conducting the ASME code as-found test on the SLC relief valves.
05000219/FIN-2015004-02 31 December 2015 23:59:59 Oyster Creek Inadequate Problem Identification and Resolution Leading to Degradation of EPR Causing a Reactor Scram A self-revealing finding was identified because Exelon did not adequately identify and correct conditions, per LS-AA-120, Issue Identification and Screening Process, that led to degradation of the electric pressure regulator (EPR) wiring, which resulted in an uncontrolled rise in reactor pressure and subsequent reactor scram on average power range monitor (APRM) Hi-Hi Flux. Specifically, Exelon failed to generate issue reports to document degraded EPR wiring that was previously identified, and therefore did not take corrective action prior to a reactor scram. Planned corrective actions include reinforcing with station personnel that an issue report is required when issues are identified. This finding is more than minor because it is associated with the equipment performance attribute of the Initiating Events cornerstone and adversely impacted its objective to limit the likelihood of events that upset plant stability and challenge critical safety functions. In accordance with IMC 0609, Attachment 4 and Exhibit 1 of Appendix A, the inspectors determined that this finding is of very low safety significance (Green) because the finding did not cause both a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The inspectors determined there is no cross-cutting aspect associated with this finding since it is not representative of current Exelon performance. Specifically, in accordance IMC 0612, the causal factors associated with this finding occurred outside the nominal three-year period of consideration and considered not representative of present performance.
05000219/FIN-2015003-01 30 September 2015 23:59:59 Oyster Creek Non-Conservative Temperature Input in the Electromatic Relief Valve Voltage Drop Calculation The inspectors identified an NCV of 10 Code of Federal Regulations (CFR) 50, Appendix B, Criterion III, Design Control, in that Exelons measures for verifying the adequacy of design of the electromatic relief valve (EMRV) voltage drop calculation were inadequate. Specifically, non-conservative temperature inputs were used for the safety related EMRV direct current voltage drop calculation, which reduced the margin of available voltage to the EMRV solenoids. Exelon entered this issue into the corrective action program for resolution as issue report 2522756, and corrective actions included revising the calculation to include the correct temperature values and conduct an extent of condition of other voltage drop calculations that could have similar temperature values. The performance deficiency is more than minor because it is associated with the design control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, lower voltage to the EMRV solenoid at higher temperatures could affect the reliability and capability of the EMRV to perform its design function. In addition, the performance deficiency is determined to be more than minor because it is similar to example 3.j of NRC IMC 0612, Appendix E, Example of Minor Issues, in that as a result of the calculation errors and the magnitude of the decrease of margin, there was a reasonable doubt on the operability of the component. The inspectors evaluated the finding using 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix A, Exhibit 2, Mitigating System Screening Questions. The inspectors determined that this finding is a deficiency that affected the design or qualification of a mitigating structure, system or component (SSC), where the SSC maintained its operability or functionality. Therefore, inspectors determined the finding to be of very low safety significance (Green). The finding is not assigned a cross-cutting aspect because it is not reflective of current performance. Specifically, the last time Exelon had an opportunity to evaluate this issue was in 2010 when Exelon identified that the EMRV solenoid voltage had low margin.
05000220/FIN-2015009-02 30 September 2015 23:59:59 Nine Mile Point Inadequate Maintenance Rule Monitoring of Unit 1 600 VAC Breaker Super System The inspectors identified a Green NCV of 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, under section (a)(1) and (a)(2) for failing to properly monitor the 600 volt alternating current (VAC) system at Unit 1 in accordance with established maintenance rule reliability criteria to assure that breakers were capable of performing their intended function. Specifically, the inspectors identified four events that were not evaluated against the established (a)(2) reliability criteria. This resulted in a failure to evaluate the 600 VAC system for potential corrective actions in accordance with (a)(1) and did not ensure effective control through preventative maintenance to show the system was capable of performing its intended function in accordance with (a)(2). Exelons immediate corrective actions included evaluations of the failures and planning for a maintenance rule expert panel for consideration of placing the system into (a)(1) where corrective actions could be developed to return the system to (a)(2) monitoring. Exelon also noted that IR 2416790 documented the challenge associated with overcurrent trip device drift and subsequent pump failures. This IR was open at the time of the inspection with actions to determine if a replacement is possible and to present any potential options to Plant Health Committee in October 2015. This performance deficiency is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the overcurrent trip devices associated with Unit 1 600 VAC General Electric (GE)-AK breakers were unreliable and resulted in the trip of five safety-related pumps between April 2013 and February 2014. Only one of the five functions was evaluated by Exelon. This impacted the ability of these pumps to be able to perform their function to provide cooling to their respective systems. In accordance with IMC 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2012, the inspectors determined this finding was of very low safety significance (Green) because this finding did not represent an actual loss of system safety function, did not represent an actual loss of function of at least a single train 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 in accordance with Exelons maintenance rule program for greater than 24 hours. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, because Exelon failed to thoroughly evaluate the failures against the monitoring criteria specified for the Unit 1 600 VAC breaker super system. Specifically, between April 2013 and February 2014, four breaker failures were identified by the inspectors that were not evaluated against the Unit 1 600 VAC breaker super system, which prevented compliance with 10 CFR 50.65 (a)(1) to ensure corrective actions are established to return the system to (a)(2) monitoring.
05000410/FIN-2015003-01 30 September 2015 23:59:59 Nine Mile Point Use of Incorrect Grounding Cart Results in Loss of Electrical Bus The inspectors identified a self-revealing Green finding (FIN) for Exelon Generation Company, LLC (Exelon) personnels failure to stop when met with unexpected conditions as required by procedure HU-AA-101, Human Performance Tools and Verification Practices. On August 21, 2015, a Unit 2 division of normal switchgear was unintentionally deenergized which required an unplanned down power to 90 percent and special operating procedure entry. The loss of the switchgear was the result of installation of an incorrect sized grounding cart in the electric fire pump breaker cubicle during breaker maintenance. Use of the correct sized grounding cart was discussed during the pre-job brief. This resulted in the loss of the electric fire pump, half of the drywell coolers, a heater drain pump, and unplanned reactivity change. Exelon entered this issue into their corrective action program (CAP) for resolution and developed corrective actions which included developing procedures for the use of grounding carts and evaluating where other skill-of-the-craft work may pose the same risk. This finding is more than minor because it is associated with the human performance attribute of the Initiating Events cornerstone and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 1 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, the inspectors determined that this finding is of very low safety significance (Green). The finding has a cross-cutting aspect in the area of Human Performance, challenge the unknown, because Exelon personnel failed to stop when faced with uncertain conditions. Specifically, after having been briefed on the different stab sizes for 1200 amp and 2000 amp grounding carts, Exelon personnel failed to stop and notify supervision when faced with unlabeled grounding carts stored in the same location, Exelon personnel failed to notify supervision or compare stab sizes to ensure the correct grounding cart was used.
05000220/FIN-2015009-01 30 September 2015 23:59:59 Nine Mile Point Failure to Identify and Correct a Condition Adverse to Quality Associated with Secondary Containment Leakage The inspectors identified a Green NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50, Appendix B, Criterion XVI, Corrective Actions, because between 2007 and 2015, Exelon staff did not promptly identify and correct a deficiency associated with Unit 2 reactor building service water pipe penetration W-3177-C. Specifically, on August 20, 2015, during Exelon staffs investigation of an inspector concern associated with the service water pipe penetration into secondary containment, a leakage path into secondary containment was discovered and was not previously identified and evaluated for impact on operability of Unit 2 secondary containment. Exelon generated issue report (IR) 2544831 to document the newly identified condition. The assessment included a review of previously identified leakage paths that were being tracked in accordance with procedure, previously performed secondary containment drawdown leakage testing, and a comparison to the maximum allowable flow rate leakage area. The assessment concluded that based on the gap that was identified at secondary containment penetration W-3177-C, there was a new total of 1.783 square inches of surface area allowing leakage into the Unit 2 secondary containment. Exelon determined this to be acceptable because calculations for secondary containment drawdown testing allows for up to 33.6 square inches of surface area causing in-leakage into secondary containment. Given 1.783 square inches of total identified leakage being less than the allowable 33.6 square inches, there was reasonable assurance that standby gas treatment system will be able to perform its drawdown function and maintain secondary containment vacuum 0.25 inches of vacuum water gauge in accordance with Technical Specification (TS) 3.6.4.1, Secondary Containment. This performance deficiency was more than minor because it impacted the design control attribute of the Barrier Integrity cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, Exelons staff failed to identify the degraded penetration seal that impacted the reasonable assurance of Unit 2 secondary containment operability. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 3 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2012, the inspectors determined this finding was of very low safety significance (Green) because the finding only represented a degradation of the radiological barrier function provided for the control room, or auxiliary, spent fuel pool, or standby gas treatment system (i.e., secondary containment). This finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, because Exelon staff failed to properly evaluate the condition identified in multiple IRs to determine the extent of condition associated with secondary containment water in-leakage. Specifically, between 2007 and 2015, three IRs were generated and a 2012 structural monitoring review documented the service water penetration water in-leakage and the issue was not appropriately evaluated for the potential for a service water pipe through-wall leak or the potential impact on secondary containment.
05000220/FIN-2015002-01 30 June 2015 23:59:59 Nine Mile Point Failure to Notify of Changes to Work Scope The inspectors identified a self-revealing NCV of Unit 1 Technical Specification (TS) 6.4.1, Procedures, for failure to follow the planned scaffold erection work scope that resulted in two personnel receiving unplanned internal exposures. Specifically, on January 6, 2015, workers erecting scaffolding changed the work scope that specified the use of new equipment and used unsurveyed highly contaminated scaffold parts instead, without notifying radiation protection staff of the change in work scope that resulted in two workers receiving unplanned, unintended internal radiation exposures. The failure to follow the planned work scope is a performance deficiency. The performance deficiency constitutes a finding that is more than minor because the performance deficiency was associated with the Occupational Radiation Safety attribute of program and process and adversely affected the cornerstone objective in the protection of workers from exposure to radioactive material. Specifically, failure to follow the planned work scope resulted in two personnel receiving unplanned internal exposures. The finding is not subject to traditional enforcement because it did not affect the regulatory process or result in actual safety consequences. Using IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, the finding was of very low safety significance (Green) because it did not involve: (1) as low as 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. The cause of the finding is related to the cross-cutting area of Human Performance, Challenge the Unknown, because when workers discovered potentially contaminated scaffold materials in the work area, they did not question whether or not it was appropriate to use the material in their job and did not raise the question to their supervisors or Exelon Generation Company, LLC (Exelon) radiation protection technicians prior to deviating from the planned and briefed work scope. As a result, the radiological risks were not evaluated before proceeding to utilize the unsurveyed highly contaminated components, which resulted in unintended internal radiation exposures to the workers.
05000219/FIN-2015002-04 30 June 2015 23:59:59 Oyster Creek Reset of the Automatic Voltage Regulator Controller Led to an Automatic Reactor Scram A self-revealing finding was identified because Exelon did not properly screen work in accordance with MA-AA-716-010, Maintenance Planning. Specifically, on September 12, 2014, Exelon did not screen the automatic voltage regulators (AVR) human machine interface (HMI) post-maintenance test per the maintenance planning procedure. As a result, on October 12, 2014, Exelon personnel performing the post-maintenance test did not have a work order, which would have included plant configurations and limitations associated with the test. This led to an automatic reactor scram. Exelon entered this issue into the corrective action program. Planned corrective actions include reinforcing with work planners that a work order is required for similar work activities. This finding was determined to be more than minor because it is associated with the human performance attribute of the Initiating Events cornerstone and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during plant operation. Specifically, resetting the three AVR controllers caused an automatic plant scram. This finding is determined to be of very low safety significance (Green), because it did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. This finding has a cross-cutting aspect in the area of Human Performance, Challenge the Unknown, because Exelon did not recognize and plan for the possibility of mistakes, or implement appropriate error reduction tools. Specifically, on October 12, 2014, Exelon personnel did not stop when faced with the uncertain situation of the HMI screen that did not respond as expected.
05000219/FIN-2015008-02 30 June 2015 23:59:59 Oyster Creek Untimely Corrective Actions to Restore Design Conformance of Two SDV Vent & Drain Valves Pressure Regulator Valves The NRC identified an NCV of Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion XVI, Corrective Action, for failure to promptly correct a condition adverse to quality. Specifically, corrective actions to restore design conformance of scram discharge volume (SDV) vent and drain valve pressure regulator valves V-6-961 and V-6-962 were not taken at the first opportunity of sufficient duration which was refueling outage 25 (1R25). Additionally, justification of the basis for deferral of corrective actions beyond the restart from 1R25 on October 2014, was not documented, reviewed, or approved by site management and/or oversight organizations as required by station procedure OP-AA-108-115, Section 4.5.5. Consequently, two non-conforming pressure regulator valves which perform a safety-related function remained installed following plant startup from 1R25, without appropriate evaluation and approval. Immediate corrective action included licensee determination that V-6-961 and 962 and the associated SDV vent and drain valves (V-15-119 and 121) remained operable, but non-conforming. Exelon entered the issue into their corrective action program as IR 2482851. The finding was more than minor because it was associated with the design control and barrier performance attributes of the Barrier Integrity cornerstone and adversely affected the cornerstone objective of ensuring the operational capability of the containment barrier to protect the public from radionuclide releases caused by accidents or events. Additionally, the finding was similar to example 5.c in Appendix E of Inspection Manual Chapter (IMC) 0612, because the control rod drive system was returned to service following 1R25 with two non-conforming (non-safety-related) pressure regulator valves installed in a safety-related application. The team determined the finding was of very low safety significance because it did not affect the reactor coolant system (RCS) boundary; did not affect the radiological barrier function of the control room, auxiliary building, or spent fuel pool systems or boundaries; and did not represent an actual open pathway in containment or involve a reduction in the function of hydrogen igniters. The team assigned a cross-cutting aspect in the area of Human Performance, Consistent Process (aspect H.13) because the organization did not use a consistent systematic approach to evaluate component operability after Exelon upgraded the classification of three pressure regulator valves from a non-safety to a safety-related status.
05000219/FIN-2015008-01 30 June 2015 23:59:59 Oyster Creek Use of an Analytical Method to Determine the Core Operating Limits Without Prior NRC Approval The NRC identified a Severity Level lV non-cited violation (NCV) of Technical Specification (TS) 6.9.1.f.2 in that Exelon did not obtain NRC approval prior to using a specific analytical method to determine the core operating limits. Specifically, Exelon used an analytical method (TRACG04P) to determine the core operating limits (the average power range monitor protection settings which were identified in the Core Operating Limits Report (COLR)); however, that particular analytical method was not previously reviewed and approved by the NRC prior to Exelons use. Exelon submitted a corrective action issue report (IR) to evaluate the condition (IR2482042). The team determined that Exelon did not comply with TS 6.9.1.f.2 requirements in that Exelon used an analytical method to determine the core operating limits without prior NRC approval. The team determined that this was a performance deficiency that was within Exelons ability to foresee and correct. Because the issue had the potential to affect the NRCs ability to perform its regulatory function, the team evaluated this performance deficiency in accordance with the traditional enforcement process. Using the Enforcement Manual, the team characterized the violation as Severity Level IV because the underlying analytical method required NRC approval prior to use. Because this violation involves the traditional enforcement process and does not have an underlying technical violation that would be considered more-than-minor within the Reactor Oversight Process (ROP), the team did not assign a cross-cutting aspect to this violation in accordance with IMC 0612, Power Reactor Inspection Reports, Section 07.03.c.
05000219/FIN-2015002-01 30 June 2015 23:59:59 Oyster Creek Inadequate Assessment of 4k Emergency Switchgear Roll-Up Door Degraded Floor Gasket The inspectors identified a finding associated with Exelon procedure, OP-AA-108-115, Operability Determinations, because Exelon did not adequately assess a degraded floor gasket for the D emergency 4 kilovolt (kV) switchgear roll-up door. Specifically, Exelon did not adequately assess the flood and fire functionality of the degraded gasket, which is credited to provide protection to safety-related D emergency 4kV switchgear during a postulated internal flood event and to contain the carbon dioxide (CO2) gaseous suppression system during a postulated fire within the D switchgear room. Exelon entered this issue into the corrective action program. Planned corrective actions include reinforcing the operability determination procedure and enhancing operator training in fire and flood functionality of gaskets. Additional corrective actions included repairing the gasket and performing a detailed analysis of the ability of degraded gasket to meet its flooding and fire function. This finding is more than minor because it is associated with the protection against external factors attribute of the Mitigating Systems cornerstone, and affected the cornerstone objective of ensuring the reliability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the degraded floor gasket could have resulted in increased water level in the D emergency 4kV switchgear room during a postulated internal flood due to a fire water pipe rupture, therefore affecting the reliability of the D emergency 4k switchgear to perform its safety function. In addition, the degraded floor gasket could have resulted in CO2 leakage out of the D emergency 4k switchgear room during a postulated fire in that room, therefore affecting the reliability of the D emergency 4k switchgear gaseous suppression system to perform its safety function. The inspectors determined that this finding is of very low safety significance (Green) because it is a deficiency that affected the design or qualification of a mitigating structure, system, or component (SSC), where the SSC maintained its operability or functionality. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, because Exelon did not thoroughly evaluate issues to ensure that resolutions address the causes and extent of conditions commensurate with their safety significance. Specifically, Exelon staff did not thoroughly evaluate the issue associated with the degraded floor gasket for fire and flood functionality.
05000219/FIN-2015002-02 30 June 2015 23:59:59 Oyster Creek Failure Rates Exceed Twenty Percent for Annual Requalification Exam A self-revealing finding was identified associated with inadequate licensed operator performance during licensed operator requalification exams in accordance with TQ-AA-150, Operator Training Program. Specifically, two of seven crews failed the simulator scenario portion of the requalification examinations. As an immediate corrective action, the crews that failed were restricted from licensed duties. Exelon entered this issue into the corrective action program, and facility training staff remediated the crews (the crews were retrained and successfully retested), and those crews were returned to licensed duties. This finding is more than minor because it is associated with the human performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, two of seven crews failed to demonstrate a satisfactory understanding of the knowledge and abilities required to safely operate the facility under normal, abnormal, and emergency conditions. The inspectors determined the finding to be of very low safety significance (Green) because it is related to requalification exam results, did not result in a failure rate of greater than forty percent, and the two crews were remediated (i.e., the crews were retrained and successfully retested) prior to returning to shift. This finding has a cross-cutting aspect in the area of Human Performance, Training, because Exelon staff did not provide adequate operator requalification training to maintain a knowledgeable, technically competent workforce.
05000219/FIN-2015002-03 30 June 2015 23:59:59 Oyster Creek Reactor Water Cleanup Procedure Not Followed Resulting in a Level Transient A self-revealing NCV of Technical Specification 6.8.1(a), Procedures and Programs, was identified because Exelon did not follow procedure 303, Reactor Cleanup Demineralizer System, during the system restoration on March 26, 2015. Specifically, during startup from a forced outage (1F36), Exelon did not follow procedure 303, which required correct valve lineups for system restoration of reactor water cleanup (RWCU) after system isolation. This resulted in decreasing reactor water level, which was automatically terminated by a second RWCU isolation. Exelon entered this issue into the corrective action program. Planned corrective actions include enhancing operator training in system knowledge and procedure compliance and revising startup procedures. This finding is determined to be more than minor because it is associated with the human performance attribute of the Initiating Events cornerstone and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, Exelon did not properly lineup the RWCU system after isolation, which resulted in a water level transient and challenging the critical safety function of inventory control. This finding is determined to be of very low safety significance (Green), because it did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. This finding has a cross-cutting aspect in the area of Human Performance, Challenge the Unknown, because Exelon did not recognize and plan for the possibility of mistakes, or implement appropriate error reduction tools. Specifically, the operators did not stop and fully communicate plant condition after the initial RWCU isolation. Consequently, operators opened the RWCU system inlet valve due to the increasing water level without following procedure guidance.
05000219/FIN-2015001-03 31 March 2015 23:59:59 Oyster Creek Incomplete 50.72 and 50.73 Reports Associated with Secondary Containment Integrity The inspectors identified a Severity Level IV NCV of 10 CFR 50.9(a) in that Exelon did not provide complete information in reports submitted per 10 CFR 50.72 and 10 CFR 50.73. Specifically, a licensee event report (LER) submitted on November 18, 2014, did not discuss a separate, partially opened secondary containment door that was discovered during the same time frame, which could have prevented the fulfillment of the safety function of secondary containment, and therefore was required to be discussed in the original LER. Exelon entered this issue into their corrective action program as IR 2440641. Planned corrective actions include revising the original LER to add a discussion of the partially opened secondary containment door. The inspectors determined that not providing a complete report in accordance with 10 CFR 50.9(a) is a performance deficiency that was reasonably within Exelons ability to foresee and correct and should have been prevented. Because the issue had the potential to affect the NRCs ability to perform its regulatory oversight function, the inspectors evaluated this performance deficiency in accordance with the traditional enforcement process. In accordance with Section 2.2.2.d of the NRC Enforcement Policy, the inspectors determined that the performance deficiency identified with the reporting aspect of the event is a Severity Level IV violation because it is of more than minor concern with relatively inappreciable potential safety significance and is related to findings that were determined to be more than minor issues. In accordance with IMC 0612, Appendix B, this issue was not assigned a cross-cutting aspect.
05000220/FIN-2015001-01 31 March 2015 23:59:59 Nine Mile Point Failure to Declare Notice of Unusual Event Following Sodium Bisulfite Spill in Unit 1 Screenhouse The inspectors documented a Green NRC-identified NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50.54(q)(2) when Exelon failed to declare a Notice of Unusual Event Emergency Action Level (EAL) (HU3.1) when entry conditions were met. Specifically, on February 4, 2015, between 9:55 a.m. and 11:15 a.m., access to the Screenhouse was prohibited due to the release of a toxic gas that adversely affected normal plant operations following a spill of sodium bisulfite. Immediate corrective actions included Exelon entering the issue into their corrective action program (CAP) as issue report (IR) 02474142, formally evaluating the decision-making process used during the incident, and clarifying responsibilities for air sampling and the reporting of samples during incidents in the future. This finding is more than minor because it was associated with the Emergency Preparedness cornerstone attribute of Emergency Response Organization Performance, and affected the cornerstone objective of ensuring that a licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Specifically, between 9:55 a.m. and 11:15 a.m., access to the Unit 1 Screenhouse was prohibited due to the release of sodium bisulfite to the Screenhouse, affecting normal plant operations of the station. This finding was evaluated using IMC 0609, Appendix B, Emergency Preparedness SDP, Section 4, Failure to Implement. The performance deficiency is associated with the emergency classification planning standard and is considered a Risk-Significant Planning Standard (RSPS). The failure to declare a Notice of Unusual Event when directed by the EAL Matrix is considered a lost or degraded RSPS in accordance with Section 4 of IMC 0609. Section 4.3.c and Attachment 1 of IMC 0609, Appendix B, provide the significance determination for a Failure to Implement, and the performance deficiency was determined to be of very low safety significance (Green). The inspectors determined that the cross-cutting aspect that contributed most to the root cause is Human Performance, Challenge the Unknown: Individuals stop when faced with uncertain conditions. Risks are evaluated and managed before proceeding. Specifically, during the event, an unknown substance was released and at no point was atmospheric analysis used in the EAL declaration decision-making process. Furthermore, although spill response personnel were experiencing symptoms that were not consistent with exposure to a spill of sodium bisulfite, this unexpected condition was not fully assessed by NMPNS for significance and reportability in accordance with procedures.
05000219/FIN-2015001-01 31 March 2015 23:59:59 Oyster Creek Post Maintenance Test Results Were Not Evaluated to Assure that Technical Specifications Requirements Were Satisfied The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, when Exelon did not document and adequately evaluate test results to assure that test requirements had been satisfied. Specifically, Exelon did not perform the proper post maintenance test procedure to assure that the requirements of Technical Specification 4.5.G.3 were satisfied following installation of a temporary modification to secondary containment. Exelon entered this issue into the corrective action program for resolution as issue report (IR) 2440643. Corrective actions include revising the process to perform the correct post maintenance test to ensure Technical Specification 4.5.G.3 is met. This finding is more than minor because it is associated with the configuration control (Standby Gas Trains) attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The inspectors evaluated the finding using IMC 0609.04, Initial Characterization of Findings, issued June 19, 2012, and IMC 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process: Phase 1 Initial Screening and Characterization of Findings, issued May 9, 2014. Because the finding degraded the ability to close or isolate secondary containment, the inspectors were required to further assess the finding using IMC 0609, Appendix H, Containment Integrity Significance Determination Process, issued May 6, 2004. The inspectors determined that this finding is of very low safety significance (Green) because the decay heat values were low, given that the unit had been shut down for approximately three days, and reactor water level was greater than that required for movement of irradiated fuel assemblies within the reactor pressure vessel. This finding has a cross-cutting aspect in the area of Human Performance, Procedure Adherence, because Exelon personnel did not perform the post maintenance test specified by the work order.
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