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05000410/FIN-2018003-022018Q3Nine Mile PointMinor ViolationDuring the review of Licensee Event Report (LER) 05000220/2017-002-01, Manual Reactor Scram Due to Presesure Oscillations, the inspectors identified a minor violation of 10 CFR 50.9, Completeness and accuracy of information. The LER was found to be inaccurate. Specifically, the LER timeline contained inaccurancies regarding the time operators entered a special operating procedure and did not include an actuation of high-pressure coolant injection (HPCI). The timeline stated at 2:10 AM operators entered the special operating procedure for Pressure Regulator Malfunction, due to reactor pressure oscillations of 2-3 psig. At 2:27 AM operators inserted a manual scram of the reactor due to pressure oscillations exceeding procedural limits. This information was confirmed by a review of the operational logs for March 20, 2017. During OI Investigation 1-2018-002, it was determined that this entry was not accurate and although an exact time could not be established is was estimated to have been at 2:20 AM vice 2:10 AM. Additionally the timeline did not include a mention that at 2:16 AM unexpected turbine trip signal was received and HPCI was initiated due to a tagging error. Operators reset HPCI at 2:18 AM and restored main feedwater flow to restore Reactor Vessel water level. A sixty day telephone notification instead of a written licensee event report was conducted for this invalid initiation of HPCI was completed on May, 11, 2017, as EN 52747 as allowed by 10 CFR 50.73(a)(2)(iv). Screening: Violations involving the submittal of inaccrurate or incomplete information are evaluated under Traditional Enforecement because they impact the NRCs regulatory process. Accordingly, the inspectors evlauted this issue against the example violations in Section 6.9 of the NRC Enforcement Policy. Inspectors concluded that the violation is of minor safety significance because the inaccurate information did not change the NRCs review of the licensee event report. Enforcement: 10 CFR 50.9 requires that information provided to the Commission by a licensee shall be complete and accurate in all material respects. Contrary to the above, on June 22, 2015, Entergy provided information to the Commission that was not complete and accurate in all material respects. In the licensee event report, Exelon documented incorrect information that resulted in the NRC launching a substation further inquiry (OI investigation), but did not substantiate that licensed operators deliberately failed to follow a Technical Specifications required procedure. Exelon identified the inaccuracy and entered the issue into the corrective action program (IR 04091110) on January 7, 2018, and submitted LER 05000220/2017-002-01 on August 18, 2018, revising the timeline to show operators entering N1-SOP-31.2 at 2:20 AM vice 2:10 AM. The disposition of this violation closed Licensee Event Report 05000220/2017-002-01
05000410/FIN-2018003-012018Q3Nine Mile PointFailure to Ensure that Thermal Power is Less Than or Equal to the Licensed Power LimitThe inspectors identified a Green finding and associated non-cited violation (NCV) of the NMPNS Unit 2 Operating License (NPF-69), Condition 2.C(1), Maximum Power Level, when Exelon did not ensure that thermal power was less than or equal to the licensed power limit of 3988 megawatts-thermal (MWth). Specifically, on multiple occurrences between May 22, 2018 and October 19, 2018, licensed operators in the main control room did not appropriately monitor and control 2-hour average thermal power at or below the licensed power limit. The inspectors determined the 2-hour average thermal power exceeded the licensed power limit outside of normal steady-state fluctuations, and did not take timely, effective corrective action to reduce thermal power below the licensed power limit when the 2-hour average was found to exceed the licensed power limit
05000336/FIN-2018011-012018Q3MillstoneReviews of Incoming Industry Operation Experience Not CompletedThe inspectors identified that Millstone could not demonstrate that incoming industry operational experience reports (ICES) since 2015 had been properly reviewed for applicability to Millstone and for those items that were applicable, were evaluated and corrective actions developed as necessary as required by program guidance. A population of over 1600 ICES reports were identified where it could not be determined if required reviews were complete. Because there are parallel processes which may have reviewed these items, additional review is necessary to determine whether this issue represents a performance deficiency that is of more than minor significance. Therefore, this item is characterized as an unresolved item (URI). The purpose of the operational experience program is to identify conditions adverse to quality (CAQs) found at other plants, evaluate whether the concern is applicable to either Millstone unit, and evaluate and develop corrective actions for those CAQs when necessary. The inspectors noted that a performance improvement report (PIR) is automatically created for the Dominion fleet whenever an OPEX report is received (regardless of its source). Once the corporate PIR is generated, each site is required to check a box that it was received and also disposition it. The PIR remains opened until each site has completed this action. Prior to 2015, the corporate Operating Experience Coordinator would perform an applicability review and assign the remaining items to the site for further evaluation. When the corporate organization was reorganized, the headquarters review of OPEX became mostly administrative and the individual sites were expected to fully disposition the report. Since 2015, more than 1600 OPEX records were discovered that required disposition for Millstone. These records were still open and no records exist to show whether reviews were completed. Therefore it is uncertain if all applicable ICES reports were reviewed. Planned Closure Actions: The NRC will conduct a problem identification and resolution annual sample using NRC IP 71152 once Dominion has notified the NRC that they have completed their review of the 1600 ICES reports. Licensee Actions: Dominion wrote Condition Report (CR) 1105042 to capture the issue, conducted an investigation, and developed a plan to review the 1600 ICES reports which have no documented reviews. Dominion anticipates this review will be completed by the end of the first quarter of 2019.Corrective Action Reference: CR 1105042NRC Tracking Number: 05000336 & 05000423/2018-011-01
05000220/FIN-2018002-022018Q2Nine Mile PointInadequate Procedure Causes Water Hammer Condition Resulting in Isolation and Inoperability of the 12 Train of the Emergency Condenser SystemThe inspectors identified a Green finding and associated NCVof 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, when Exelon did not provide appropriate quantitative or qualitative criteria and guidance to operators in procedure N1- OP- 13 Emergency Cooling System to return an emergency condenser loop to service without inducing a water hammer condition which caused operators to re-isolate the emergency condenser loop and declare it inoperable
05000410/FIN-2018002-012018Q2Nine Mile PointFailure to Ensure Proper Control of the Standby Gas Treatment System Damper Valve, 2GTS*V2000B, Within Procedures, Materials, and Design Control MeasuresThe inspectors identified a Green finding and associated NCVof 10 CFRPart 50, Appendix B, Criterion III, Design Control, when Exelon failed to ensure proper control of the SGTS damper valve 2GTS*V2000B within procedures, materials, and design control measures. Specifically, on April 15, 2018 operators attempted to run B SGTS for containment purge; however, no flow was observed and the system was secured. Operators discovered the 2GTS*V2000B closed due to the failure of the operating mechanism to maintain control of the valve position.
05000244/FIN-2018011-022018Q1GinnaFailure to Procedurally Verify Fuel Transfer Cart Results in Fuel Interference EventA self-revealing Green non-cited violation (NCV)of Technical Specification 5.4.1.a, Procedures, was identified for the failure of Exelon to operate refueling equipment in accordance with technical procedures in April and May of 2017, which resulted in a fuel interference event, damage to the rod cluster control assembly, and the need for a detailed inspection of a fuel assembly
05000244/FIN-2018011-012018Q1GinnaPotential Preconditioning of Turbine Driven Auxiliary Feedwater Surveillance TestingThe NRC identified a Green non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion XI, Test Control, because Exelon established unevaluated preconditioning, with a reasonable doubt of whether the preconditioning was acceptable, prior to testing of the turbine driven auxiliary feedwater pump. This results in the loss of as-found conditions which challenge the capability of the test to assure that the turbine driven auxiliary feedwater pump will perform satisfactorily in service.
05000317/FIN-2018001-012018Q1Calvert CliffsFailure to Conduct Adequate Radiation Surveys and Evaluate Potential Radiological HazardsAself-revealed Green non-cited violation(NCV)of Title 10 Code of Federal Regulations(10 CFR) 20.1501, Surveys and Monitoring: General, was identified when Exelon failed to perform adequate surveys of the 11 reactor coolant pump bay area following the aggregation of 25 high dose-rate in-core detectors in one area of the flooded refueling cavity, which is adjacent to the pump bay. Surveys were not performed as required after radiological conditions changed and radiological hazard mitigation measures, such as locking and controlling access in accordance with Exelon procedures, were not implemented, resulting in accessible dose-rates of up to 2,000 millirem per hour(mrem/hr)in the pump bay
05000247/FIN-2017010-012017Q4Indian PointInadequate Diesel Fuel Oil Temperature ProtectionThe NRC identified a finding for the failure to assure that diesel powered Diverse and Flexible Coping Strategies (FLEX) equipment would be reliable to mitigate postulated beyond-design basis external events during very low temperature conditions. Specifically, at temperatures below 21F, portable FLEX equipment, such as emergency diesels, steam generator and reactor makeup pumps, and transfer pumps, were susceptible to conditions in which they would not have been capable of starting and operating due to fuel crystalizing or gelling. (CR-IP2-2017-04902/IP3-2017-05574)The failure to ensure that the portable diesel equipment could function within the required temperature range was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external factors attribute of the Mitigating Systems cornerstone and adversely affected the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The significance of the finding was evaluated using NRC Inspection Manual Chapter 0609, Appendix O, Significance Determination Process for Mitigating Strategies and Spent Fuel Pool Instrumentation (Orders EA-12-049 and EA-12-051), dated October 7, 2016, and Appendix M, Significance Determination Process Using Qualitative Criteria, dated April 12, 2012. The event of concern was determined to be a seismic event greater than 0.3g resulting in a loss of offsite power during extreme cold weather events. A bounding evaluation was performed in accordance with Step 4.1.1 of Appendix M. Indian Point declared full compliance with the order on August 12, 2016. The preliminary review of available weather conditions for the site, from the time of full compliance, shows that the temperature was below the cloud point of the fuel for over 200 hours. The Indian Point Unit 3 External Initiator Risk Informed Notebook was utilized to estimate the risk and was determined to adequately model the risk of both units. Utilizing Table 5.3.2, sequences that included emergency power, auxiliary feedwater, and high pressure makeup were evaluated. Assuming a 200 hour exposure and the unavailability of all diesel driven FLEX equipment the risk was determined to less than 1E-7/yr. Therefore, the finding was determined to have a very low risk significance. The finding had a cross-cutting aspect in the Avoiding Complacency of the Human Performance area because the licensee failed to ensure that all susceptible elements of the mitigation strategies were designed, maintained, or operated in such a manner that they could reliably function over then entire temperature spectrum for beyond-design basis external events. (H.12)
05000220/FIN-2017001-012017Q1Nine Mile PointDeficient Design Control of Outboard MSIV Pilot Valve Instrument Air SupplyGreen. 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-2017001-022017Q1Nine Mile PointFailure to Identify and Correct a Non- Conforming Condition in Safety-Related UPSsGreen. 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.
05000388/FIN-2016008-032016Q3SusquehannaFailure to Implement or Develop Timely Interim or Final Corrective Actions for a Degraded ConditionThe inspectors documented a self-revealing finding of very low safety significance (Green) against Susquehanna procedures LS-125 Revision 4, Corrective Action Program (CAP), and OI-AD-096 Revision 18, Operator Challenges, for the failure to correct and establish appropriate corrective actions for a known degraded condition for an uninterruptable power supply (UPS) for vital 120 VAC load centers. Specifically, Susquehanna did not correct nor establish compensatory actions for the transfer switch for a UPS which was failed for over one year. The degraded condition subsequently complicated operator response to the loss of a vital 480 VAC switchboard and resulted in an unplanned manual reactor scram and valid emergency core cooling system (ECCS) actuation on May 13, 2016. Susquehanna entered this issue into their CAP, conducted an apparent cause evaluation, and repaired the UPS transfer switch. The finding was more than minor because it was associated with the Equipment Performance attribute of the Initiating Events cornerstone and adversely affected the associated cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the long standing degraded condition of UPS 2D14212/2B246082 was not corrected or compensated for and did not function as designed, as a result operators had to manually scram the reactor following the loss of a vital bus on May 13, 2016. In accordance with IMC 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Exhibit 1 of IMC 0609, Appendix A, The SDP for Findings At-Power, dated June 19, 2012, the inspectors determined that this finding is of very low safety significance (Green) because the performance deficiency did not cause both a reactor trip and loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. Specifically, while this performance deficiency resulted in a reactor scram, it was not the cause of the loss of mitigation equipment credited in the Susquehanna safety analysis. This finding had a cross-cutting aspect in the area of Problem Identification and Resolution Resolution because the organization did not take effective corrective actions to address issues in a timely manner commensurate with its safety significance. Specifically, failing to establish appropriate compensatory actions for this known degraded condition, prevented the operators from responding appropriately to a loss of a vital 480 VAC switchboard initiating event. (P.3). (Section 4OA2.1.c(3))
05000387/FIN-2016008-012016Q3SusquehannaFailure to Write a Condition Report for Degraded Conditions Which Challenged Operability of Safety Related EquipmentThe inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for Susquehanna failing to identify and correct conditions adverse to quality in a timely manner. Specifically, between April 16, 2016 and April 22, 2016, condition reports for potential or suspected degraded or non-conforming conditions related to the High Pressure Coolant Injection System (HPCI) and Reactor Core Isolation Cooling System (RCIC) were not written and operability determinations performed. In both cases, the equipment was subsequently declared inoperable due to the conditions. The issues were entered into the CAP and the equipment was taken out of service, repaired, and retested satisfactorily. The inspectors determined that there were two examples of the same performance deficiency and violation. In accordance with NRC Enforcement Manual Section 1.3.4, Documenting Multiple Examples of a Violation, multiple examples of a single violation are allowed to be documented as a single violation bounded by the characterization of the most significant example. The RCIC example is considered the most significant due to the longer exposure time in a required mode and number of mode changes that occurred during the exposure period. The finding was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the associated cornerstone objective to ensuring the capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the failure to identify and correct degraded conditions associated with a RCIC system lube oil leak which rendered that system inoperable. In accordance with IMC 0609.04, Initial Characterization of Findings, dated June 19, 2012, the inspectors determined that this finding screened to Green because the safety function was not lost, and the finding did not represent an actual loss of function of at least a single train for greater than its Tech Spec Allowed Outage Time or two separate safety systems out-of-service for greater than its Tech Spec Allowed Outage Time. This finding had a cross-cutting aspect in the area of Human Performance, Teamwork, because individuals and work groups did not communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety is maintained. Specifically, in both examples, individuals were aware of potential degraded conditions but actions were not taken to communicate the activity to other groups, such as the control room operators, to allow for the issues to be evaluated for operability and determine if proposed actions were timely and/or appropriate. (H.4) (Section 4OA2.1.c(1))
05000387/FIN-2016008-022016Q3SusquehannaFailure to Implement and Maintain Quality Procedure Results in Control Room Chiller InoperabilityThe inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to implement and maintain a quality procedure, MT-GE-021, Chiller Maintenance and Inspection. This resulted in the safety related 0K112A chiller being operated outside of its design specifications and being declared inoperable. Specifically, on January 9, 2014, a system engineer directed the maintenance personnel to overcharge 0K112A with R-114a refrigerant, which led to higher power consumption by the chillers compressor motor, and the failure of the next biennial surveillance test on December 10, 2015 due to excessive compressor motor current. Susquehanna entered the issue into the CAP, conducted testing to establish the proper refrigerant charge, removed the excess refrigerant, and revised the procedure. The finding was determined to be more than minor because it was associated with the Mitigating System cornerstone attribute of Equipment Performance and adversely affected the associated cornerstone objective to ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). The refrigerant overcharge condition resulted in the 0K112A chiller being inoperable and unable to fulfil its safety function to cool safety related switchgear and equipment during accident conditions for a period of 23 months. 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 a detailed risk evaluation would be required because the finding involved an actual loss of function of at least a single Train for greater than its Technical Specification allowed outage time of 30 days. A detailed risk assessment was performed by a Region 1 Senior Reactor Analyst (SRA). The SRA determined the finding to be of very low safety significance (Green.) This finding had a cross-cutting aspect in the area of Human Performance, Procedure Adherence because individuals did not follow processes, procedures, and work instructions. Specifically, for many years maintenance and engineering personnel relied upon informal work practices vice referring to the procedure when charging the chillers with refrigerant. (H.8) (Section 4OA2.1.c(2))
05000387/FIN-2016008-042016Q3SusquehannaFailure to Promptly Identify and Correct a Condition Adverse to Quality on Vital 480 VAC MCCsThe inspectors documented a self-revealing Green NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for failure to identify and correct a condition adverse to quality. Specifically, in October and December 2006 and July 2009, Susquehanna did not identify a non-conforming condition with the design and performance requirements of several 480 volt motor control center (MCC) breaker assemblies during receipt inspections. These non-conforming breaker assemblies were installed in vital 480 VAC applications and subsequently led to a phase to ground short and loss of a 480 volt safety-related motor control center on May 12, 2016. Susquehanna entered this issue into their CAP, conducted an apparent cause evaluation, replaced the damaged breaker assembly, and is conducting an extent of cause review for other susceptible breaker assemblies. The finding was more than minor because it was associated with the Design Control attribute of the Initiating Events cornerstone and adversely affected the associated cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, on May 12, 2016, an electrical transient on vital AC bus 2B246 occurred as a result of a phase to ground fault in breaker cubicle 2B24609, which resulted in a loss of bus 2B246 and associated safety related loads. In accordance with IMC 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Exhibit 1 of IMC 0609, Appendix A, The SDP for Findings At-Power, dated June 19, 2012, the inspectors determined that this finding is of very low safety significance (Green) because the performance deficiency did not cause both a reactor trip and loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. This finding did not have a crosscutting aspect because the performance deficiency was a historical issue with the actions taken in 2005, 2006, and 2009, and is not indicative of current licensee performance. (Section 4OA2.1.c(4))
05000317/FIN-2016003-012016Q3Calvert CliffsDeficient Design Control of Air Pressure Available for Unit 1 Component Cooling Water Air Operated ValvesThe inspectors identified a Green non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion III, Design Control, for Exelons failure to establish measures to assure that the design basis was correctly translated into specifications affecting safety related functions of air operated valves (AOV). Specifically, when implementing a design change, Exelon failed to verify the air pressure supplied to AOVs in the component cooling (CC) water system was adequate to ensure that the valves would have performed their safety function to close during certain specific accident conditions. The inspectors determined that Exelons failure to verify ECP-15-000213 ensured that air pressure supplied to safety related Unit 1 CC heat exchanger (HX) outlet AOVs was sufficient to support their safety function of closing during a design basis accident (DBA) was a performance deficiency that was reasonably within its ability to foresee and correct and should have been prevented. Exelons immediate corrective actions included conducting an engineering evaluation that demonstrated the operability of the CC system in the degraded condition and increasing the air pressure supplied to the CC HX outlet valves to ensure the valves are capable of fully closing during a DBA. Exelon entered this issue into its corrective action program (CAP) as action request (AR) 02680281. The inspectors reviewed IMC 0612, Appendix B, Issue Screening, and determined the issue is more than minor because it adversely affected 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. The inspectors also reviewed IMC 0612, Appendix E, Examples of Minor Issues, and found it was sufficiently similar to Example 3.j, in that the design analysis deficiency resulted in a condition where reasonable doubt existed regarding the operability of the Unit 1 CC HX outlet valves. In accordance with IMC 0609, Attachment 4, Initial Characterization of Findings, issued on June 19, 2012, and IMC 0609, Appendix A, The Significance Determination Process for Findings at Power, issued on June 19, 2012, the inspectors determined that this finding is of very low safety significance (Green) since, the finding did not involve an actual open pathway in the physical integrity of reactor containment. The inspectors determined that the cause of the finding has a cross-cutting aspect in the area of Human Performance, Documentation, because Exelons AOV program, as implemented by ER-AA-410, Air Operated Valve Implementing Program, Revision 2, did not require that complete, accurate, and up-to-date documentation on the CC HX outlet valves design be maintained. (H.7)
05000244/FIN-2016002-012016Q2GinnaIncorrect Emergency Action Level TableExelon identified that they had inadvertently made a change to the Ginna Emergency Plan. The NRC determined that this error is a preliminary White finding under the Reactor Oversight Process and a violation of Title 10 of the Code of Federal Regulations (10 CFR) 50.54 (q)(2), Emergency Plans, because Exelon did not maintain the effectiveness of Ginnas Emergency Plan such that it met the requirements of Appendix E, Emergency Planning and Preparedness for Production and Utilization Facilities, and the planning standards of 10 CFR 50.47(b). Specifically, Exelon implemented a revision to the emergency action level (EAL) table for the fission product barrier matrix that was incorrect with respect to the EAL threshold associated with potential loss of containment barrier. This could have resulted in an untimely declaration of a General Emergency or a failure to declare a Site Area Emergency during an actual event. Using IMC 0612, Appendix B, Issue Screening, the performance deficiency was determined to be more than minor because it impacted the procedure quality attribute of the Emergency Preparedness cornerstone and adversely affected the associated cornerstone objective to ensure that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Specifically, Exelons EAL table was revised without adequate technical reviews resulting in a discrepancy between the EAL table and the EAL technical basis. The EAL wording of Table F-1 containment barrier potential loss, block C.6 did not meet the minimum required operable equipment in all situations and could have resulted in a delayed General Emergency declaration or a failure to declare a Site Area Emergency. The inspectors utilized IMC 0609, Appendix B, Emergency Preparedness Significance Determination Process (SDP), to determine the significance of the performance deficiency. The performance deficiency is associated with the emergency classification system planning standard and is considered a risk-significant planning standard function. The inspectors were directed by the SDP to compare the performance deficiency with the examples in Section 5.4, 10 CFR 50.47 (b)(4), Emergency Classification System, to evaluate the significance of this performance deficiency. In accordance with Section 5.4, when an EAL has been rendered ineffective such that any General Emergency declaration would not be declared, but due to other EALs, an appropriate declaration would be made in a degraded manner or any Site Area Emergency would not be declared for a particular off-normal event, a degradation of risk-significant planning standard function (b)(4) is determined; and the finding is White. The finding has a cross-cutting aspect in the area of Human Performance, Change Management, because Exelon did not use a systematic process for evaluating and implementing change so that nuclear safety remains the overriding priority. Specifically, Exelon did not maintain a clear focus on nuclear safety when implementing changes to the EALs resulting in a significant unintended consequence, the potential to make an untimely emergency declaration.
05000317/FIN-2016002-022016Q2Calvert CliffsFailure to Report Conditions as Required by 10 CFR 50.73The inspectors identified a Severity Level IV, NCV of 10 CFR 50.73(a)(2) for Exelons failure to report within 60 days of discovery, a condition that could have prevented the fulfillment of the safety function of the service water (SRW) system needed to mitigate the consequences of an accident. Additionally, Exelon failed to report within 60 days of discovery, a single condition that caused two trains of the SRW system, a system designed to mitigate the consequences of an accident, to become inoperable. Exelon entered the issue into their CAP as IR 02688409 and on July 20, 2016, submitted LER 05000317/2016-004-00, High Energy Line Break Barrier Breached Due to Human Performance Error Causing Both Service Water Trains to be Inoperable. The inspectors determined that Exelons failure to report a single condition that caused the inoperability of two trains of SRW and may have prevented SRW from fulfilling its design functions to mitigate the consequences of an accident within 60 days of discovering the condition was a violation of 10 CFR 50.73(a)(2), and could have impacted the regulatory process. The inspectors reviewed IMC 0612, Appendix B, Issue Screening, and the NRC Enforcement Policy, revised February 4, 2015, and determined the violation is of SL-IV because it is most similar to example 6.9.d.9 of the NRC Enforcement Policy, A licensee fails to make a report required by 10 CFR 50.72 or 10 CFR 50.73, which is a SL-IV violation. The inspectors determined that the violation did not have a cross-cutting aspect because it involved the traditional enforcement process only.
05000318/FIN-2016002-032016Q2Calvert CliffsFailure to Implement Engineering Change Procedures Results in Plant TripThe inspectors documented a self-revealing, Green finding for Exelons failure to implement procedures for engineering changes. Specifically, Exelon failed to address the full scope and critical parameters associated with a modification to a steam generator feed pump (SGFP). As a result, the 22 SGFP turbine pedestal studs were improperly torqued, resulting in the SGFP shifting, becoming misaligned, and eventually resulting in the failure of the turbine to pump coupling. This resulted in the unexpected tripping of the 22 SGFP on December 1, 2015, and operators inserting a manual reactor trip as required by procedure. The inspectors determined that Exelons failure to properly implement procedures CNG-CM-1.01-1003, Design Inputs and Change Impact Screen, Revision 00601, Attachment 12; CNG-CM-1.01-2000, Scoping and Identification of Critical Components, Revision 00201; and CNG-FES-007, Preparation of Design Inputs and Change Impact Screen, Revision 00010 was a performance deficiency that was a performance deficiency that was within Exelons ability to foresee and prevent. Exelons corrective actions included, replacing the failed coupling, verifying the torque on the 21 SGFP using a HYTORCTM, and developing an adverse condition monitoring plan for Unit 1s SGFPs. Exelon conducted a root cause evaluation (RCE) and developed corrective actions to preclude repetition (CAPR) including implementation of Exelon procedure HU-AA-1212, Technical Task Risk/Rigor Assessment, Pre-Job Brief, Independent Third Party Review, and Post-Job Review, Revision 007 and conducting critical parameters and rigor training for engineering personnel including the expectations for three pass reviews and verification of assumptions. The inspectors reviewed IMC 0612, Appendix B, Issue Screening, and IMC 0612, Appendix E, Examples of Minor Issues and determined the issue is more than minor because it was associated with the Design Control Attribute of the Initiating Events Cornerstone and adversely impacted the associated cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the performance deficiency resulted in a reactor trip from full power on December 1, 2015. The inspectors evaluated the finding using IMC 0609, Attachment 4, Initial Characterization of Findings, issued on June 19, 2012, and IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 1, Initiating Events Screening Questions, issued on June 19, 2012 and determined the finding to be of very low safety significance (Green) because the finding did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The inspectors determined that the finding had a cross-cutting aspect in the area of Human Performance, Documentation, because Exelon failed to develop and maintain complete and accurate engineering change packages (ECP), work orders (WO), and maintenance procedures.(H.7)
05000410/FIN-2016002-012016Q2Nine Mile PointIneffective Corrective Action Results in Water Intrusion to Battery Switchgear RoomThe 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-2016002-022016Q2Nine Mile PointFailure to Identify Wide Range Level Indication Impacts Operability of HPCS and RCICThe 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)
05000410/FIN-2016002-032016Q2Nine Mile PointFailure to Understand Radiological Conditions Results in Unintended ExposureA 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.
05000317/FIN-2016002-012016Q2Calvert CliffsScaffolding Impairs Fire Sprinkler Systems in Safety Related Fire AreasThe inspectors identified a Green, NCV of CCNPP Renewed Facility Operating License for Units One and Two, paragraph 2.E for Exelons failure to maintain in effect all provisions of the approved fire protection program as described in the Updated Final Safety Analysis Report (UFSAR). Specifically, Exelon installed scaffolding in safety related areas not in accordance with approved procedures and, therefore, impaired fire sprinkler systems that were required by the approved fire protection program without establishing approved contingency measures. The inspectors determined that Exelons impairment of fire sprinkler systems by installing scaffolding with dimensions exceeding those approved in Exelon procedure MA-AA-716-025 was a performance deficiency that was within Exelons ability to foresee and prevent. The performance deficiency led to the violation of CCNPP Renewed Facility Operating License, paragraph 2.E, because Exelon failed to maintain in effect all provisions of the approved fire protection program. Exelons immediate corrective actions included stationing continuous fire watches and removal of the scaffolding deck boards which were impairing the fire sprinkler systems. Exelon entered these issues in to their corrective action program (CAP) as issue reports (IR): 02642463, 02642549, 02642844, 02644495, 02647104, 02647454, and 02647455. The inspectors reviewed IMC 0612, Appendix B, Issue Screening, and determined the issue is more than minor because it adversely affected the protection against external factors attribute of the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, Exelon installed scaffolding that exceeded the allowed dimensions in MA-AA-716-025 and impaired the function of fire sprinkler systems in areas containing safety related equipment. The inspectors evaluated the finding using IMC 0609, Attachment 4, Initial Characterization of Findings, issued on June 19, 2012, and IMC 0609, Appendix F, The Fire Protection SDP Worksheet issued on September 20, 2013 and determined the finding to be of very low safety significance (Green) because, in all cases of impairment, the fire sprinkler systems were still capable of protecting their intended targets or were still capable to suppress fires such that no additional equipment important to safety would have been affected. The inspectors determined that the finding had a cross-cutting aspect in the area of Human Performance, Procedure Adherence, because Exelon failed to properly implement procedure MA-AA-716-025, Scaffold Installation, Modification, and Removal Request Process, Revision 11, which limits scaffolding dimensions and locations when installing scaffolding in safety related areas. (H.8)
05000410/FIN-2016001-012016Q1Nine Mile PointInadequate Procedure Leading to Failure to Manage Elevated Risk during Preventive MaintenanceThe 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-2016001-022016Q1Nine Mile Point50.65(a)(4) Risk Evaluation Not Properly Performed Prior to Residual Heat Removal Heat Exchanger TestingThe 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.
05000220/FIN-2016001-032016Q1Nine Mile PointInadequate Tagout Resulting in Reactor Building Closed-Loop Cooling Drain Down EventA 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.
05000410/FIN-2016001-042016Q1Nine Mile PointLicensee-Identified ViolationEight-hour reports. If not reported under paragraphs (a), (b)(1), or (b)(2) of this section, the licensee shall notify the NRC as soon as practical and in all cases within eight hours of the occurrence of any of the following: (v) Any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to: (C) Control the release of radioactive material. Contrary to the above, from April 2, 2014, until October 5, 2015, Exelon failed to submit an EN to the NRC within 8 hours upon discovery on a condition which could have prevented the safety function of a SSC needed to control the release of radioactivity on April 2, 2014, at 11:20 a.m. Specifically, secondary containment being declared inoperable due to both airlock doors being open at the same time in Mode 5 with an OPDRV in progress. The inspectors reviewed the violation using IMC 0612 Appendix B, Issue Screening, and the NRC Enforcement Policy. This violation impacted the regulatory process so traditional enforcement applies. Comparing this violation to the examples in the NRC Enforcement Policy Chapter 6, the violation matches Severity Level IV Example 6.9.d.9, a licensee fails to make a report required by 10 CFR 50.72 or 10 CFR 50.73. The NRC did not rely upon the information to make any regulatory decisions and the error did not result in increased scope or effort of NRC inspections. Compliance was restored when Exelon submitted LER 05000410/2014-007-01, Secondary Containment Inoperable due to Simultaneous Opening of Airlock Doors, to correct the public record and inform the NRC. Exelon staff entered the issue into its CAP.
05000410/FIN-2016001-052016Q1Nine Mile PointLicensee-Identified ViolationThe holder of an operating license under this part shall submit a Licensee Event Report (LER) for any event of the type described in this paragraph within 60 days after the discovery of the event. (v) Any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to: (C) Control the release of radioactive material. Contrary to the above from June 2, 2014, until October 5, 2015, Exelon failed to submit an LER notification to the NRC within 60 days after discovery of a condition which could have prevented the safety function of a SSC needed to control the release of radioactivity on April 2, 2014 at 11:20 a.m. Specifically, secondary containment being declared inoperable due to both airlock doors being open at the same time in Mode 5 with an OPDRV in progress. The inspectors reviewed the violation using IMC 0612, Appendix B and the NRC Enforcement Policy. This violation impacted the regulatory process so traditional enforcement applies. Comparing this violation to the examples in the NRC Enforcement Policy Chapter 6, the violation matches Severity Level IV Example 6.9.d.9, a licensee fails to make a report required by 10 CFR 50.72 or 10 CFR 50.73. The NRC did not rely upon the information to make any regulatory decisions, and the error did not result in increased scope or effort of NRC inspections. Compliance was restored when Exelon submitted LER 05000410/2014-007-01 to correct the public record and inform the NRC. Exelon staff entered the issue into its CAP.
05000317/FIN-2016001-012016Q1Calvert CliffsIssue of concern Regarding Characterization and Acceptance of a Relevant Indication in Pressurizer to Nozzle Dissimilar Metal WeldAn unresolved item (URI) was identified by the inspectors relating to an issue of concern involving Exelons acceptance and characterization of the relevant indication in weld 4-SR-1006-1 during prior refuel outages. Additional information is required to determine whether a performance deficiency, which is more than minor, exists. Description. Based on a review of Exelon letter dated February 25, 2016, the inspectors preliminarily concluded the relevant indication in weld 4-SR-1006-1 was incorrectly accepted during prior refuel outages and was not in conformance with ASME Code Section XI, Article IWA-3000. Additional inspection, including review of Exelons root cause analysis of this issue, is warranted to determine whether a performance deficiency, which is more than minor, exists related to characterization and acceptance of a relevant indication in weld 4-SR-1006-1. (URI 05000317/2016001-01, Issue of Concern Regarding Characterization and Acceptance of a Relevant Indication in Pressurizer to Nozzle Dissimilar Metal Weld)
05000317/FIN-2015003-032015Q3Calvert CliffsLicensee-Identified ViolationTS 5.4.1.a states, in part, that written procedures shall be established and maintained covering the applicable procedures recommended in RG 1.33, Revision 2, Appendix A, February 1978, of which Section 9 specifies procedures for performing maintenance. The vendor technical manual specifies the need to conduct routine lube oil sample analysis and Exelon procedure MA-AA-716-006, Control of Lubricants Program, Revision 11, directs the performance of sampling in accordance with specific site approved procedures. Contrary to the above, following the June 17, 2015, failure of the 1A EDG surveillance test, Exelon identified that appropriate procedural guidance did not exist for the processing of 1A EDG engine lube oil samples. On June 17, 2015, during surveillance testing of the 1A EDG, Exelon secured the engine due to high lube oil filter differential pressure. The engine lube oil filters were determined to be clogged due to engine coolant contamination of the engine lube oil system caused by leakage past O-rings on one engine cylinder piston. Investigation determined that monthly engine lube oil samples were not provided to the vendor for analysis from February May 2015 due to the extended absence of the regular lubrication specialist and lack of procedural guidance for processing of lube oil samples once they were obtained. Subsequent analysis of these samples revealed that the engine lube oil had elevated potassium levels which is indicative of lube oil contamination by engine coolant. The inspectors evaluated the issue using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings AtPower, which determined that the finding was of very low safety significance (Green) because the safety function was not lost and the 1A EDG was not considered inoperable for greater than its TS limiting condition for operation allowed outage time. The inspectors determined that Exelon correctly evaluated the finding and developed appropriate corrective action as documented in Exelons CAP as IR02517365.
05000317/FIN-2015003-022015Q3Calvert CliffsLicensee-Identified Violation10 CFR 74.19 (c), Recordkeeping, states, in part that, each licensee who is authorized to possess special nuclear material (SNM), shall conduct a physical inventory of all SNM in its possession, under license, at intervals not to exceed 12 months. Contrary to this, on May 22, 2015, Exelon identified that the 2014 SNM inventory had not been completed by the end of August 2014, as was required since the 2013 SNM inventory was completed in August 2013. The 2014 SNM inventory was started on August 26, 2014, and was completed on October 6, 2014. Exelon subsequently self-identified that inventories of nine locations had exceeded 12 months although all SNM was accounted for by October 6, 2014. The inspectors determined that this finding was of very low safety significance (Green), because the finding did not represent an actual loss of SNM and the performance of an inventory in June 2015, as part of the corrective actions, was completed satisfactorily. The inspectors determined that Exelon correctly evaluated the finding and developed appropriate corrective action as documented in Exelons CAP as IR02504484.
05000317/FIN-2015003-012015Q3Calvert CliffsFailure to Establish and Maintain Procedures for the Operation of the Diesel Fuel Oil SystemThe inspectors identified a Green NCV of Technical Specification (TS) 5.4.1.a for Exelons failure to adequately establish and maintain procedures as required by Regulatory Guide (RG) 1.33, Appendix A, Section 3, Procedures for Startup, Operation, and Shutdown of Safety-Related PWR Systems. The inspectors determined that Exelons failure to adequately establish and maintain a procedure for the operation of the diesel fuel oil (DFO) supply system was a performance deficiency. Exelon entered this issue into their corrective action program (CAP) as issue report (IR) 02541107. Exelons immediate corrective actions included halting of opening of 0-DFO-108, 21 Fuel Oil Storage Tank (FOST) to Auxiliary Boilers Isolation, and initiating an evaluation to determine the seismic adequacy of the piping downstream of 0-DFO-108. The inspectors reviewed IMC 0612, Appendix B, Issue Screening, and determined the issue is more than minor because it adversely affected the protection against external factors attribute of the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to adequately establish and maintain procedure Operating Instruction (OI)-21D, Fuel Oil Storage and Supply, Revision 10, for the operation of the DFO supply system resulted in the alignment of the safety-related 21 FOST to nonsafety-related/non-seismically qualified piping thus rendering the 21 FOST inoperable. In accordance with IMC 0609, Attachment 4, Initial Characterization of Findings, issued on June 19, 2012, and IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, and Exhibit 4, External Events Screening Questions, issued on June 19, 2012, the inspectors determined that a detailed risk evaluation was necessary to disposition the significance of this finding because the loss of the 21 FOST would degrade two or more trains of a multi-train system or function. A regional Senior Reactor Analyst (SRA) performed a detailed risk evaluation and determined the finding to be of very low safety significance (Green). The inspectors determined that the finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Operating Experience, because Exelon failed to adequately evaluate relevant external operating experience. Specifically, Exelon failed to evaluate for systems where non-seismically qualified piping could be connected to safety-related tanks as was described in Information Notice (IN) 2012-01, Seismic Considerations Principally Issues Involving Tanks. (P.5).
05000482/FIN-2015002-012015Q2Wolf CreekClass 1E 4kV Feeder Breakers from Station Blackout Diesel Generators Current Transformer Wiring not Installed per Design DrawingsThe inspectors identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, for not installing the current transformer wiring in the Class 1E 4kV alternate feeder breaker cubicles from the station blackout diesel generators per the design drawings. As a result, testing performed seven months after the system was declared operational identified that the connections were unable to power the safety-related buses due to incorrect wiring of the current transformers. The licensee entered this issue into the corrective action program as Condition Report 83379. This finding was more than minor because it was associated with the Mitigating Systems Cornerstone to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, due to the incorrect wiring of the current transformers, the SBO diesel generators were unable to power safety related buses as they were designed. The inspectors performed the initial significance determination for the finding using NRC Inspection Manual 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012. The finding required a detailed evaluation because it had the potential to degrade at least one train of a system that supports a risk significant system or function. Therefore, a senior reactor analyst performed a bounding detailed risk evaluation. The finding was of very low safety significance (Green) because the risk assessment programs quantified the change in core damage frequency less than 1.0x10-6. The inspectors determined that the finding had a teamwork cross-cutting aspect in the area of human performance. The licensee individuals and work groups did not communicate and coordinate their activities within and across organizational boundaries. Specifically a drawing revision was not properly attached to the work order which resulted in the incorrect wiring of both trains, and because different groups were completing different components, parts of the wiring were incorrectly installed per a superseded revision.
05000482/FIN-2015001-042015Q1Wolf CreekFailure to Station Boundary Watch for Opening Auxiliary Building Emergency Exhaust System Boundary DoorThe inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, Drawings, associated with the licensees failure to follow the requirements of Station Procedure AP 10-104, Breach Authorization, Revision 32. Specifically, the licensees failure initiate a breach permit and station a boundary watch when the auxiliary building emergency exhaust system boundary door 41015 was opened multiple times for transporting scaffolding from the turbine building to the auxiliary building. Opening this door without compensatory measures rendered the auxiliary building emergency exhaust system inoperable. The license entered this issue into their corrective action program for resolution as Condition Reports 92315 and 92630. The licensees failure to initiate a breach permit and implement required compensatory measures for when the auxiliary building emergency exhaust system boundary door 41015 was open was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the system, structure, and component and barrier performance attribute of the Barrier Integrity Cornerstone, and affected the associated cornerstone objective to ensure the radiological barrier functionality of the auxiliary building emergency exhaust system. Specifically, without a dedicated individual in constant communication with the control room, as required by AP 10-104, opening this door required entry of Technical Specification 3.7.13 Limited Condition of Operation Condition B. The longest period door 41015 was open was approximately one hour without the required compensatory measure. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Finding At-Power, dated June 19, 2012, inspectors determined that the finding screened as having very low safety significance (Green) because the finding only involved a degradation of the radiological barrier function provided for the auxiliary building. The finding has a cross-cutting aspect in the area of human performance associated with work management. Specifically, the organization failed to implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority, including the identification and management of risk commensurate to the work (H.5).
05000482/FIN-2015001-052015Q1Wolf CreekLicensee-Identified ViolationTechnical Specification Section 5.4.1.a requires, in part, that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Appendix A to Regulatory Guide 1.33, Quality Assurance Program Requirements, Revision 2, February 1978. Section 1.c of Regulatory Guide 1.33 requires procedures for equipment control (e.g. locking and tagging). Station Procedure AP 21E-001, Clearance Orders, Revision 37, requires that the shift manager, ensure that plant conditions can support establishing the clearance order boundaries, including activities such as removing equipment from service. Contrary to the above, on January 28, 2015, the licensee failed to ensure that plant conditions could support the clearance order boundaries during preparation and implementation of clearance orders. Specifically, the preparation and implementation of clearance order EJ-A-005 unintentionally rendered both trains of the residual heat removal system inoperable and necessitated an unplanned entry into Technical Specification 3.0.3 for 2 hours. The performance deficiency was determined to be more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone, and affected the associated objective to ensure availability, reliability, and capability of systems that respond to initiating event to prevent undesirable consequences (i.e. core damage). Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Finding At-Power, dated June 19, 2012, inspectors determined a detail risk evaluation was required because this finding represented a loss of system and/or function. Therefore, a senior reactor analyst performed a bounding detailed risk evaluation. The analyst noted that the isolation of valve EJ HV8716A would only affect the reliability of hot leg injection for train B. Hot leg injection is a necessary function to ensure that there will not be unacceptably high concentrations of boric acid in the core region (resulting in precipitation of a solid phase) during the long-term cooling phase following a postulated large-break loss of coolant accident. Consequently, valve alignments affecting hot leg injection are only of concern during large-break loss of coolant accidents. Using the simplified plant analysis risk model, the analyst noted that the frequency of a large-break loss of coolant accident (LLOCA) was 2.5 x 10-6 /year. As stated above, the exposure period was two hours or 2.28 x 10-4 years. The analyst then calculated the upper bound risk impact of the performance deficiency to be 5.7 x 10-10. Therefore, this finding is of very low safety significance (Green).
05000220/FIN-2015001-012015Q1Nine Mile PointFailure to Declare Notice of Unusual Event Following Sodium Bisulfite Spill in Unit 1 ScreenhouseThe 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.
05000410/FIN-2015001-022015Q1Nine Mile PointFailure to Perform an Adequate Review of Planned Work Activities Results in a Manual Reactor ScramThe inspectors documented a self-revealing Green finding (FIN) for Exelons failure to properly review a work package associated with the replacement of a reactor vessel level recorder as required by MA-AA-716-234, FIN Team Process, Revision 8. Specifically, on February 18, 2015, control room operators manually scrammed Unit 2 when reactor vessel water level unexpectedly rose above desired limits during a planned replacement of Unit 2 reactor vessel level recorder 2ISC-LR1608. The unplanned rise in reactor water level occurred when daisy chained leads associated with the level recorder were lifted, which caused an interruption in the feedwater level control circuit. The inspectors determined that Exelons failure to ensure measures were in place to address the impact on reactor vessel level prior to level recorder replacement in accordance MA-AA-716-234 was a performance deficiency that was reasonably within Exelons ability to foresee and correct and should have been prevented. 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. Specifically, Exelon did not ensure measures were in place to prevent an adverse impact on the feedwater level control system during level recorder replacement. This resulted in a rapid rise in reactor water level and subsequent manual reactor scram. 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, issued June 19, 2012, the inspectors determined that this finding is of very low safety significance (Green) because while the performance deficiency caused a reactor scram, it did not result in the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The finding has a cross-cutting aspect in the area of Human Performance, Avoid Complacency, because Exelon failed to recognize and plan for the possibility of mistakes, latent issues, and inherent risk even while expecting successful outcomes. Specifically, Exelon did not ensure measures were in place to address the impact of the level recorder replacement on the feedwater level control system.
05000482/FIN-2015001-022015Q1Wolf CreekFailure to Assess the Operability of Emergency Diesel Generator B during Emergent Work ActivitiesThe inspectors identified a non-cited violation of Technical Specification 5.4.1.a, associated with the failure to properly preplan maintenance such that it would not affect safety-related equipment in accordance with procedure AP 22C-008, On-Line Qualitative Risk Management, Revision 3. Specifically, during planning of emergent work activities on January 29, 2015, the licensee failed to recognize that when electrical cabinet doors containing safety-related under voltage and under frequency relays were opened to accomplish troubleshooting activities, the cabinet was not in a seismically qualified configuration. Thus the maintenance had the potential to impact the reliable operation of emergency diesel generator B during a seismic event. The licensee initiated Standing Order 37, Safety Related Cabinet Operability Requirements, Revision 0, to provide the requirements for assessing operability of opening safety-related electrical cabinet and panel doors out of their seismically qualified configuration during maintenance activities and entered this issue into their corrective action program for resolution as Condition Reports 91501 and 94605. The licensees failure to properly preplan maintenance such that it would not affect safety-related equipment during emergent work activities was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone, and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating event to prevent undesirable consequences (i.e., core damage). Specifically, the licensees failure to properly preplan maintenance resulted in emergency diesel generator B being placed in a condition that did not meet its seismic design requirements. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Finding At-Power, dated June 19, 2012, inspectors determined that the finding was of very low safety significance (Green) because the finding: (1) was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and (4) did not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety-significance in accordance with the licensees maintenance rule program. The finding has a cross-cutting aspect in the area of human performance associated with work management. Specifically, the licensee did not implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority, including the identification and management of risk commensurate to the work (H.5).
05000482/FIN-2015001-032015Q1Wolf CreekFailure to Complete an Adequate Operability Evaluation for Declaring the Train A Control Room Air Conditioning Unit OperableThe inspectors identified non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, associated with the licensees failure to complete an adequate operability evaluation in accordance with procedure AP-28001,Opeability Evaluations, Revision 24 following the failure to meet a surveillance test acceptance criteria. Specifically, the licensee did not have an accurate technical basis for declaring the train A control room air condition unit operable when the minimum air flow rate was not met. The licensees operability evaluation, which declared the train A control room air condition unit operable, incorrectly applied instrument uncertainty and used a superseded minimum air flow value. When these inaccuracies were addressed, the licensee determined the train was inoperable. The licensee entered this issue into their corrective action program as Condition Report 92274. The licensees use of an inadequate technical basis for an operability evaluation of a non-conforming condition resulting in the train A control room air conditioning air condition unit being declared operable when it was actually inoperable was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone, and affected the associate cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating event to prevent undesirable consequences (i.e., core damage). Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Finding At-Power, dated June 19, 2012, inspectors determined that the finding was of very low safety significance (Green) because the finding: (1) was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and (4) did not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety-significance in accordance with the licensees maintenance rule program. The finding has a cross-cutting aspect in the area of human performance associated with conservative bias component because the licensee did not use a decision making-practice that emphasized prudent choices over those that are simply allowable. A proposed action was determined to be safe in order to proceed, rather than unsafe in order to stop (H.14).
05000220/FIN-2014005-012014Q4Nine Mile PointIncomplete and Inaccurate Medical Information Provided by Exelon Which Impacted Issuance of Initial and Renewal LicensesExelon Generation Company, LLC (Exelon) identified two AVs: (1) An AV of Title 10 of the Code of Federal Regulations (10 CFR) 50.9, Completeness and Accuracy of Information; and (2) An AV of 10 CFR 50.74, Notification of Change in Operator or Senior Operator Status. Specifically, during an internal audit in July 2014, Exelon identified that between September 2002 and February 2012, NMPNS staff submitted certified copies of an NRC reactor operator and/or senior operator license applications for seven applicants that did not specify that the applicants required a restriction in order to maintain medical qualifications. The NRC issued the reactor operator and senior operator initial and renewed licenses for the seven applicants, but without the necessary medical restrictions (AV #1). From June 2002 through August 2014, Exelon had numerous additional opportunities to identify these potentially disqualifying medical conditions and that license conditions were required during the biennial licensed operator requalification program reviews and medical examinations. On September 25, 2014, a period that exceeded 30 days from when the conditions were identified, the facility notified the NRC of these medical conditions via a letter requesting amendment to the seven operators licenses to include the appropriate restrictions (AV #2). The NRC issued the license amendment with the new restrictions. The NRC inspectors also identified an additional example of both AVs which had not been reported by Exelon to the NRC in the September 25, 2014 letter. On November 5, 2014, Exelon requested termination of the license for that operator. This issue was entered into Exelons corrective action program (CAP) The inspectors determined that Exelons failure to provide complete and accurate information to the NRC in the reactor operator and senior operator license applications and to notify the NRC of a change in a reactor operator or senior operators status for a condition which was known by Exelon were performance deficiencies that were within their ability to foresee and correct and should have been prevented. The inspectors determined that traditional enforcement applies, as the issue affected the NRCs ability to perform its regulatory function. Namely, the NRC requires Exelon to ensure all licensed operators meet the medical conditions of their licenses. If, during the term of the individual operator license, an operator develops a permanent physical or mental disability that causes the operator to fail to meet the requirements of 10 CFR 55.21, Medical Examination, the licensee shall notify the NRC within 30 days of learning of the diagnosis, in accordance with 10 CFR 50.74(c). Additionally, the NRC issued reactor operator and senior operator licenses to the applicants based on information that was not complete and accurate in all material aspects. The performance deficiencies were screened against the Reactor Oversight Process per the guidance of IMC 0612, Appendix B, Issue Screening. No associated Reactor Oversight Process finding was identified and no cross-cutting aspect was assigned. These issues constitute AVs in accordance with the NRCs Enforcement Policy, and their final significance will be dispositioned in separate future correspondence. (Section 1R11)
05000410/FIN-2014005-042014Q4Nine Mile PointAssessment of UPS3B Failure Which Resulted in a Reactor ScramIntroduction. A URI was identified pending Exelons revision and approval of their root cause report associated with the failure of UPS3B that caused a Unit 2 reactor scram on March 4, 2014 Description. Unit 2 is equipped with two 10-kVA UPSs (2VBB-UPS3A and 2VBBUPS3B) that feed RPS logic trip channel loads and main steam line isolation valves control solenoids through their associated distribution panels. 2VBB-UPS3B feeds the RPS trip system B. The loads are normally energized from 600 volts alternating current (VAC) non-safety-related power. In the case of the loss of normal supply power, an inverter allows the loads to receive power from its backup direct current source. In the case of an inverter failure, the UPS can be fed from an alternate non-safety-related 600 VAC source. Each UPS is connected to its associated distribution panel through two redundant electric protective assemblies connected in series. The electric protective assemblies provide redundant protection to the RPS system and other associated essential circuits against overvoltage, undervoltage, and under frequency conditions in the non-safety-related power sources. On March 4, 2014, 2VBB-UPS3B experienced a capacitor failure on an associated circuit card. This failure prevented the UPS from transferring to its alternate source of power causing the electrical protective assemblies to trip, a loss of cooling water to the reactor recirculation pumps, and a subsequent reactor trip. Exelon staff documented the issue in CR-2014-001725 and performed a root cause analysis. Using investigative root cause techniques outlined in procedure CNG-CA-1.01-1004, Root Cause Analysis, Revision 00801, Exelon staff determined the root cause to be a lack of vendor and industry guidance and internal/external operating experience resulting in lack of PM task to preclude backplane failure. The corrective actions to prevent recurrence involved revising the PM strategy in the IQ Review and Maximo database to include replacement of all single-point vulnerable components in 2VBA*UPS2A/2B and 2VBB-UPS3A/3B. During inspection of Unit 2 LER 2014-003-00, Uninterruptible Power Supply Failure and Subsequent Manual Scram, the inspectors reviewed the root cause report associated with this event. The inspectors discovered that, although the root cause postulated that warping/cracking of the backplane contributed to UPS3B failure, when new information regarding the backplane that contradicted this root cause was discovered, Exelo personnel did not properly enter this new information into the CAP or elevate the concern to Exelon plant management. Specifically, the engineering staff and a vendor representative had examined the UPS3B backplane during the Unit 2 refueling outage and found no indication of cracking or warping. This examination occurred following management review committee approval of the root cause. This information, along with other testing performed on the UPS3B during the refueling outage, showed that the theory for potential backplane warping/cracking likely was not the actual root cause and that the corrective actions developed for backplane replacement may not prevent recurrence of the UPS failure. Exelon documented the inspectors observation in IR 2416757 and plans to evaluate the issue further and to reopen and update the root cause report. This issue will be opened as a URI pending Exelon revision of the root cause report; and NRC review of the root cause report to determine whether the issue contains performance deficiency, whether or not that performance deficiency is more than minor, and whether a violation exists. Exelon is tracking this issue through their CAP database with a date to determine root cause revision requirements by December 19, 2014. (URI 05000410/2014005-04, Assessment of UPS3B Failure Which Resulted in a Reactor Scram)
05000244/FIN-2014005-022014Q4GinnaFailure to Report a Permanent Change in a Licensed Operator's Medical Status and Request a Condition be Placed on the Operator's LicenseExelon Generation Company, LLC (Exelon) identified two apparent violations (AVs): (1) An AV of Title 10 of the Code of Federal Regulations (10 CFR) 50.9, Completeness and Accuracy of Information; and (2) An AV of 10 CFR 50.74, Notification of Change in Operator or Senior Operator Status. Specifically, on October 8, 2008, Ginna submitted certified copies of an NRC senior operator license application that did not specify that the applicant required a restriction (to take medication as prescribed for high blood pressure) in order to maintain medical qualifications. The NRC issued the senior operators initial license on December 5, 2008, but without the necessary medical restriction (AV #1). From October 8, 2008, until July 16, 2014, Ginna had several additional opportunities to identify that the blood pressure medication was required to compensate for a disqualifying medical condition and that a license condition was required during the licensees biennial licensed operator requalification program reviews and medical examinations. On July 16, 2014, a period that exceeded 30 days from when the condition was identified, the facility notified the NRC of the medical condition via a letter requesting amendment to the operators license to include the restriction (AV #2). On August 28, 2014, the NRC issued the license amendment with the new restriction. This issue was entered into Exelons corrective action program (CAP). The inspectors determined that Exelons failure to provide complete and accurate information to the NRC in the senior operator license application and to notify the NRC of a change in a senior operators status for a condition which was known by the licensee and were a performance deficiencies that were within their ability to foresee and correct and should have been prevented. The inspectors determined that traditional enforcement applies, as the issue affected the NRCs ability to perform its regulatory function. Namely, the NRC relies upon Exelon to ensure all licensed operators meet the medical conditions of their licenses. If, during the term of the individual operator license, an operator develops a permanent physical or mental disability that causes the operator to fail to meet the requirements of 10 CFR 55.21, Medical Examination, the licensee shall notify the NRC within 30 days of learning of the diagnosis, in accordance with 10 CFR 50.74(c). Additionally, the NRC issued a senior operator license to the applicant based on information that was not complete and accurate in all material aspects. The performance deficiencies were screened against the Reactor Oversight Process per the guidance of IMC 0612, Appendix B, Issue Screening. No associated Reactor Oversight Process finding was identified and no cross-cutting aspect was assigned. These issues constitute AVs in accordance with the NRCs Enforcement Policy, and their final significance will be dispositioned in separate future correspondence. (Section 1R11)
05000244/FIN-2014005-052014Q4GinnaA' Emergency Diesel Generator Output Breaker Fails to Close during Routine Surveillance Testing Resulting in a Condition Prohibited by Technical Specifications and a Potential Inability to Fulfill a Safety FunctionOn September 10, 2014, during performance of a routine scheduled surveillance test, STP-O-12.1, Emergency Diesel Generator A, Revision 01600, the output supply breaker to safeguards bus 14 failed to close on demand. Initial troubleshooting revealed no obvious issues with the breaker, and the output supply breaker functioned as required during a second test. A spare breaker was installed and tested satisfactorily on Enclosure 22 September 11, and the A EDG was restored to operable. Exelon concluded that the A EDG had been inoperable since the last successful performance of STP-O-12.1 on August 13, 2014. This 29 day period exceeded the TS allowable outage time of 7 days. Exelons subsequent troubleshooting revealed no electrical issues with the circuit breaker, and the failure modes and effects analysis concluded that the most likely cause of the circuit breaker failing to close was the breaker did not properly reset after performance of the surveillance test on August 13, 2014. The breaker could not be verified to be reset without an internal inspection. The original equipment manufacturer was also requested by Exelon to investigate the cause of the breaker failure. The original equipment manufacturer concluded that the lack of free movement of the operating mechanism trip shaft was the cause of the breaker not resetting and closing. The trip shaft did not move freely due to lack of end-to-end play. Exelons apparent cause evaluation associated with this issue and AR 02178745 noted that these circuit breakers undergo full PM every 4 years, and all PMs on both EDG output breakers have been done in accordance with the PM frequency. The last performance of the PM for the bus 14 breaker was on November 14, 2011. The procedure for the PM has the technicians check for free movement of the trip of the trip shaft, but not end-to-end play movement or clearances to allow end-to-end play. Additionally, the vendor manual does not direct measuring clearances or verifying end-to-end play; this is called out as a vendor task. Therefore, the inspectors concluded that no performance deficiency existed since it was not reasonable for Exelon to foresee and prevent this issue. The inspectors reviewed LER 2014-003-00 and determined that traditional enforcement applies in accordance with IMC 0612, Sections 0612-09 and 0612-13, and NRC Enforcement Policy, Section 2.2.4.d, because a violation of NRC requirements existed without an associated Reactor Oversight Process performance deficiency. The inspectors determined that the maintenance completed on the bus breaker was in accordance with vendor recommendations. This issue was considered to be a Severity Level IV violation of TS 3.8.1 in accordance with Enforcement Policy Section 6.1.d. In addition, IMC 0612, Appendix B, Figures 1 and 2, Issue Screening, were referenced in documenting this Severity Level IV self-revealing violation. This issue was entered into Exelons CAP as AR 02178745. Because it was not reasonable for Exelon to have been able to foresee and prevent the breaker failure, the NRC determined no performance deficiency existed. Thus, the NRC has decided to exercise enforcement discretion in accordance with Section 3.5 of the NRC Enforcement Policy and refrain from issuing enforcement action for the violation (EA-15- 004). Further, because Exelons action and/or inaction did not contribute to this violation, it will not be considered in the assessment process or the NRCs action matrix. This LER is closed.
05000220/FIN-2014005-022014Q4Nine Mile PointFailure to Make Timely Reports of Changes in Licensed Operator Medical Status Which Impacted Issuance of Initial and Renewal LicensesExelon Generation Company, LLC (Exelon) identified two AVs: (1) An AV of Title 10 of the Code of Federal Regulations (10 CFR) 50.9, Completeness and Accuracy of Information; and (2) An AV of 10 CFR 50.74, Notification of Change in Operator or Senior Operator Status. Specifically, during an internal audit in July 2014, Exelon identified that between September 2002 and February 2012, NMPNS staff submitted certified copies of an NRC reactor operator and/or senior operator license applications for seven applicants that did not specify that the applicants required a restriction in order to maintain medical qualifications. The NRC issued the reactor operator and senior operator initial and renewed licenses for the seven applicants, but without the necessary medical restrictions (AV #1). From June 2002 through August 2014, Exelon had numerous additional opportunities to identify these potentially disqualifying medical conditions and that license conditions were required during the biennial licensed operator requalification program reviews and medical examinations. On September 25, 2014, a period that exceeded 30 days from when the conditions were identified, the facility notified the NRC of these medical conditions via a letter requesting amendment to the seven operators licenses to include the appropriate restrictions (AV #2). The NRC issued the license amendment with the new restrictions. The NRC inspectors also identified an additional example of both AVs which had not been reported by Exelon to the NRC in the September 25, 2014 letter. On November 5, 2014, Exelon requested termination of the license for that operator. This issue was entered into Exelons corrective action program (CAP) The inspectors determined that Exelons failure to provide complete and accurate information to the NRC in the reactor operator and senior operator license applications and to notify the NRC of a change in a reactor operator or senior operators status for a condition which was known by Exelon were performance deficiencies that were within their ability to foresee and correct and should have been prevented. The inspectors determined that traditional enforcement applies, as the issue affected the NRCs ability to perform its regulatory function. Namely, the NRC requires Exelon to ensure all licensed operators meet the medical conditions of their licenses. If, during the term of the individual operator license, an operator develops a permanent physical or mental disability that causes the operator to fail to meet the requirements of 10 CFR 55.21, Medical Examination, the licensee shall notify the NRC within 30 days of learning of the diagnosis, in accordance with 10 CFR 50.74(c). Additionally, the NRC issued reactor operator and senior operator licenses to the applicants based on information that was not complete and accurate in all material aspects. The performance deficiencies were screened against the Reactor Oversight Process per the guidance of IMC 0612, Appendix B, Issue Screening. No associated Reactor Oversight Process finding was identified and no cross-cutting aspect was assigned. These issues constitute AVs in accordance with the NRCs Enforcement Policy, and their final significance will be dispositioned in separate future correspondence. (Section 1R11)
05000244/FIN-2014005-012014Q4GinnaIncomplete and Inaccurate Medical Information Provided by Exelon Which Resulted in Issuance of an Initial Senior Operator License without a Required Medical RestrictionExelon Generation Company, LLC (Exelon) identified two apparent violations (AVs): (1) An AV of Title 10 of the Code of Federal Regulations (10 CFR) 50.9, Completeness and Accuracy of Information; and (2) An AV of 10 CFR 50.74, Notification of Change in Operator or Senior Operator Status. Specifically, on October 8, 2008, Ginna submitted certified copies of an NRC senior operator license application that did not specify that the applicant required a restriction (to take medication as prescribed for high blood pressure) in order to maintain medical qualifications. The NRC issued the senior operators initial license on December 5, 2008, but without the necessary medical restriction (AV #1). From October 8, 2008, until July 16, 2014, Ginna had several additional opportunities to identify that the blood pressure medication was required to compensate for a disqualifying medical condition and that a license condition was required during the licensees biennial licensed operator requalification program reviews and medical examinations. On July 16, 2014, a period that exceeded 30 days from when the condition was identified, the facility notified the NRC of the medical condition via a letter requesting amendment to the operators license to include the restriction (AV #2). On August 28, 2014, the NRC issued the license amendment with the new restriction. This issue was entered into Exelons corrective action program (CAP). The inspectors determined that Exelons failure to provide complete and accurate information to the NRC in the senior operator license application and to notify the NRC of a change in a senior operators status for a condition which was known by the licensee and were a performance deficiencies that were within their ability to foresee and correct and should have been prevented. The inspectors determined that traditional enforcement applies, as the issue affected the NRCs ability to perform its regulatory function. Namely, the NRC relies upon Exelon to ensure all licensed operators meet the medical conditions of their licenses. If, during the term of the individual operator license, an operator develops a permanent physical or mental disability that causes the operator to fail to meet the requirements of 10 CFR 55.21, Medical Examination, the licensee shall notify the NRC within 30 days of learning of the diagnosis, in accordance with 10 CFR 50.74(c). Additionally, the NRC issued a senior operator license to the applicant based on information that was not complete and accurate in all material aspects. The performance deficiencies were screened against the Reactor Oversight Process per the guidance of IMC 0612, Appendix B, Issue Screening. No associated Reactor Oversight Process finding was identified and no cross-cutting aspect was assigned. These issues constitute AVs in accordance with the NRCs Enforcement Policy, and their final significance will be dispositioned in separate future correspondence. (Section 1R11)
05000244/FIN-2014005-042014Q4GinnaLicensee-Identified ViolationAccording to 10 CFR 50.74, each licensee shall notify the NRC within 30 days of a change in an operators or senior operators status including termination of any operator or senior operator. Contrary to this requirement, in AR 02120732, Exelon identified that Ginna staff did not notify the NRC of termination of two senior operators. The facility terminated the affected operators August 9, 2013, but did not notify the NRC of the change in status until September 10, 2014. This issue meets the criteria for a Severity Level IV violation because the September 10, 2014, notification did not result in increased inspection activities or cause the NRC to reconsider a regulatory position.
05000410/FIN-2014005-032014Q4Nine Mile PointMissed Surveillance Test of Alternate Decay Heat Removal Secondary Containment Isolation ValvesThe inspectors identified a Green NCV of Unit 2 Technical Specification (TS) 5.4, Procedures, for Exelons failure to properly perform procedure N2-OSP-GTS-R001, Secondary Containment Integrity Test, Revision 01100. Specifically, Exelon staff failed to ensure spectacle flanges associated with alternate decay heat (ADH) secondary containment isolation were properly installed. As a result, surveillance testing associated with ADH check valves 2ADH*V21A/B and 2ADH*V22A/B was not performed to ensure secondary containment integrity as required by N2-OSP-GTS-R001. Exelon immediately entered this issue into their CAP as issue report (IR) 2403311. Exelon entered TS Surveillance Requirement (SR) 3.0.3, Limiting Condition for Operability Applicability, which is used when a licensee discovers that a surveillance test requirement has not been performed. As required by the TS, Exelon completed a risk evaluation of the missed surveillance and determined large early release frequency remained low without ADH secondary containment isolation. Exelon also performed extent-of-condition inspections for other systems which may not have proper alignment to ensure they are meeting TS requirements. On December 4, Exelon rotated the spectacle flanges to the no flow isolation position to ensure secondary containment integrity was maintained The finding is more than minor because it is associated with the configuration control attribute of the Barrier Integrity cornerstone objective to provide reasonable assurance tha physical design barriers protect the public from radionuclide releases caused by accident or events. Specifically, by performing N2-OSP-GTS-R001 in 2012 and 2014 without first ensuring the spectacle flanges were properly installed, Exelon did not verify the secondar containment requirements of TS SR 3.4.6.1 were maintained. Additionally, this issue wa similar to Example 3.d in IMC 0612, Appendix E, Examples of Minor Issues, in that th failure to implement the TS SR as required was not minor if the surveillance had not bee conducted. By not correctly testing the secondary containment in 2012 and 2014, the SR o TS 3.4.6.1 was not met. In accordance with IMC 0609.04, Initial Characterization o Findings, and Exhibit 3 of IMC 0609, Appendix A, The Significance Determination Proces for Findings At-Power, the inspectors determined this finding is of very low safet significance (Green) because the finding only represents a degradation of the radiologica barrier function provided for the control room, or auxiliary, spent fuel pool (SFP), or standb gas treatment system (boiling water reactor). This finding has a cross-cutting aspect in th area of Human Performance, Avoid Complacency, because Exelon staff did not implemen appropriate error reduction tools. Specifically, operators did not use error reduction tools t ensure the spectacle flanges were installed in the no flow position and as a result, the failed to leak test the ADH check valves in the secondary containment drawdown test a required by N2-OSP-GTS-R001 (H.12).
05000244/FIN-2014005-032014Q4GinnaLicensee-Identified ViolationAccording to 10 CFR 55.21 and 33, licensed operators are required to have a physical examination every 2 years to ensure that their medical condition and general health will not adversely affect the performance of assigned operator job duties or cause operational errors endangering public health and safety. As a part of licensed operator medical evaluations, olfactory testing is required as specified in ANSI/ANS 3.4 1983. Olfactory testing in the standard states, Nose. Ability to detect odor of products of combustion and of tracer and marker gases. Contrary to this requirement, in CR-2014-003860, Exelon identified that Ginna medical staff had not been testing operators for the mercaptan marker used in natural gas. This violation is subject to traditional enforcement because of the potential impact upon the regulatory process since the operators medical conditions are reviewed by the NRC when issuing or renewing operator licenses. This issue meets the criteria for a Severity Level IV violation because upon subsequent olfactory testing, all operators were found to meet the health requirements for licensing.
05000317/FIN-2014004-012014Q3Calvert CliffsMain Steam Line Drain Containment Isolation Valves not Scoped in In-Service TestingThe inspectors identified a Green NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50.55a, Codes and Standards, for Exelons failure to meet the test requirements set forth in the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) for main steam line drains (MSLDs) and containment isolation valves (CIVs) motor operated valves (MOVs) (6611, 6612, 6613, 6615, 6620, 6621). Specifically, Exelon failed to scope the MSLD MOVs in their in-service testing (IST) program. As a result, the MOVs reliability was not ensured due to valve degradation not being trended as required in the IST program. Also, the MOV operability was in question because the valves were never tested to perform their containment isolation function. Exelon entered this issue into their corrective action program (CAP) as condition report (CR)-2014-005961. Immediate corrective actions included testing the MOVs. The inspectors determined that the failure to scope and meet the testing requirements of the OM Code for MSLD MOVs in accordance with 10 CFR 50.55a was a performance deficiency. This finding is more than minor because it was associated with the barrier performance attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system (RCS), and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the failure to scope and test the MSLD MOVs in accordance with the OM Code did not ensure component reliability by monitoring valve degradation and did not provide assurance that the MSLD MOVs would perform their CIV function in order to protect the public from radionuclides releases during a steam generator tube rupture (SGTR) with a loss of offsite power event. The inspectors reviewed IMC 0609.04, Initial Characterization of Findings, issued June 19, 2012, and IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 3, Barrier Integrity Screening Questions issued June 19, 2012, and determined that the finding was of very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containment, containment isolation system, and heat removal components and the finding did not involve an actual reduction of hydrogen igniters in the reactor containment. The inspectors determined that this finding did not have a cross-cutting aspect because the most significant contributor to the performance deficiency was not reflective of current licensee performance. Specifically, the 2007 IST fourth year interval submittal was the last reasonable opportunity for Exelon to identify this issue.
05000317/FIN-2014004-022014Q3Calvert CliffsLicensee-Identified ViolationTS 3.4.10, Pressurizer Safety Valves, requires two pressurizer safety valves to be operable during Modes 1 and 2, and in Mode 3 when all RCS cold leg temperatures are greater than 365F for Unit 1 or 301F for Unit 2. With one pressurizer safety valve inoperable, TS 3.4.10, Condition A, requires the inoperable valve to be restored within 15 minutes. If this is not able to be completed or if two pressurizer safety valves are inoperable, then TS 3.4.10, Condition B, is entered which requires the unit to be in Mode 3 within 6 hours AND the unit to be cooled down to below 365F for Unit 1 or 301F for Unit 2 within 12 hours. Contrary to the above, on March 12, 2013, Unit 2 pressurizer safety valve BNO4375, which had been installed in position 2RV200 during the previous operating cycle, was measured higher than its TS allowable value during as-found lift point testing. On February 28, 2014, Unit 1 pressurizer safety valves BN04373 and BM07952, which had been installed in positions 1RV200 and 1RV201 respectively during the previous operating cycle, were measured lower than their TS allowable value during as-found lift point testing. In both cases, the valves had been replaced with tested, operable valves prior to discovery of the as-found condition. Exelon concluded that the valve had been inoperable for a period of time greater than the allowed TS outage times specified in TS 3.4.10. Exelon entered both issues into their CAP as CR-2013-002415, CR- 2014-002236, and CR-2014-002237. In accordance with IMC 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix A, The Significance Determination Process for Findings at Power, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that each example was a finding of very low safety significance (Green) because the finding did not represent an actual loss of the pressurizer safety valve systems credited safety function to relieve pressure to prevent RCS pressure from exceeding 110 percent of RCS pipings design pressure.
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