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05000352/FIN-2016003-022016Q3LimerickLicensee-Identified ViolationThe following violation of very low safety significance (Green) was identified by Exelon and is a violation of NRC requirements which meets the criteria of the NRC Enforcement Policy for being dispositioned as a non-cited violation. 10 CFR 50.54(q)(2), Emergency Plans, requires, in part, that a holder of a licensee under this part shall follow and maintain the effectiveness of an emergency plan that meets the requirements in Appendix E to this part, and for nuclear power reactor licensees, the planning standards of 50.47(b). 10 CFR 50.47(b)(4) requires that a standard emergency classification and action level scheme, the bases of which include facility system and effluent parameters, is in use by the nuclear facility licensee. Contrary to the above, from April 25, 2016, until August 3, 2016, the spent fuel pool level emergency action level (EAL) RG2/RS2 threshold of Limericks Emergency Plan for a General Emergency and Site Area Emergency did not meet the requirements of Appendix E and the planning standards of 10 CFR 50.47(b). Specifically, Exelon identified that the spent fuel pool level for RG2/RS2 threshold was 0.08 feet, and the correct threshold value was 0.8 feet. The spent fuel pool EAL threshold values for a lowering water level for an Alert and Unusual Event were correct at 10.20 feet and less than 22 feet, respectively. The normal spent fuel pool water level is over 23 feet. The inspectors evaluated this finding using IMC 0609, Appendix B, Emergency Preparedness Significance Determination Process, Table 5.4-1. This Table indicates, in part, that the following should be assessed as low safety significance (White): an EAL has been rendered ineffective such that any General Emergency would not be declared for a particular off-normal event, but because of other EALs, an appropriate declaration could be made in a degraded manner (e.g. delayed), and, an EAL that has been rendered ineffective such that any Site Area Emergency would not be declared for a particular off-normal event. However, the inspectors confirmed that the spent fuel pool level instrumentation at LGS goes off scale at approximately 0.635 feet, and the Limerick Emergency Plan, in Addendum 3, directs any Emergency Director to assume the EAL threshold has been exceeded if the associated parameter goes off scale. In addition, the NEI recommended and NRC endorsed value for this EAL threshold would have been at nominally 0.0 feet, the level at which the fuel remains covered and actions to implement make-up water addition should no longer be deferred. Although the LGS threshold for declaration at 0.8 feet would have been exceeded, the inspectors concluded that the event would have been classified when the SFP level dropped below 0.635 feet, sufficiently above the NEI recommended level. Because the event would have been declared with margin to the actual water level needed for protection of the public, i.e. the spent fuel would still be fully covered by water at the time of the EAL declaration(s), the inspectors concluded that this performance deficiency was most similar to the Table 5.4-1 branches representing very low safety significance (Green). Exelons corrective actions included revising EP-AA-1008, Addendum 3, with the correct spent fuel pool level EAL RG2/RS2 threshold of 0.8 feet. Because this issue was of very low safety significance (Green) and Exelon entered the issue into the corrective action program (IR 2700440), this finding is being treated as a non-cited violation, consistent with Section 2.3.2.a of the Enforcement Policy.
05000353/FIN-2016003-012016Q3LimerickInadequate Design Control of Plant Processing Computer ModificationA self-revealing finding of very low safety significance (Green) was identified when Exelon did not implement their engineering design control procedures during the plant processing computer (PPC) modification. Specifically, Exelon did not fully address effects of the modification on other plant systems and did not establish a testing boundary that encompassed all components whose operation was altered by the modification. As a result, the PPC modification had a wiring design error that resulted in the trip of both reactor recirculation pumps (RRPs) which required a manual reactor trip of Unit 2. In response to this issue, Exelon initiated IR 2676712, investigated the cause of the trip, fixed the wiring design error, performed a root cause evaluation, and performed an extent of condition review. This issue is more than minor because it adversely affected the design control attribute of the initiating events cornerstone to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the PPC modification process had a wiring design error that resulted in the trip of both RRPs which required a manual reactor trip of Unit 2. The issue was evaluated in accordance with IMC 0609, Appendix A, "Significance Determination Process for Findings At-Power, using Exhibit 1, "Initiating Events Screening Questions, Section B, Transient initiators. The finding was determined 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 this finding has a cross-cutting aspect in the area of Human Performance, Challenge the Unknown, because LGS staff did not stop when faced with uncertain conditions, and risks were not evaluated and managed before proceeding. Specifically, Exelon did not stop and reevaluate the risks and effects on plant systems when changes were made to the PPC design modification package. (H.11)
05000353/FIN-2015003-012015Q3LimerickInadequate Procedure for RWCU Backwashing OperationsA self-revealing Green NCV of Technical Specification (TS) 6.8.1.a, Procedures and Programs, occurred because Exelon failed to establish, implement, and maintain an adequate procedure for the control of radioactivity and limiting personnel exposure during operation of a solid radioactive waste system. Specifically, the procedure for the conduct of reactor water cleanup (RWCU) filter media backwashing and collection was inadequate to ensure a sufficient receiving tank volume prior to transferring waste media. On June 28, 2015, this resulted in the overflow of a Unit 2 RWCU collection tank and back up of the reactor building floor drain system, causing high levels of radioactive contamination in accessible portions of the Unit 2 reactor building, and resulting in radioactive contamination of personnel. Exelon controlled access, decontaminated affected areas and personnel, conducted bounding dose assessments, performed extent of condition reviews, and revised affected procedures to address the issue. Exelon placed this issue into the corrective action program as issue report (IR) 2520732. This issue is more-than-minor because if left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, the failure to effectively control and manage radioactive material could result in significant unplanned, unintended occupational radiation exposure of workers. Using IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, the inspectors determined that this finding was of very low safety significance (Green) because the finding did not involve an as low as is reasonable achievable (ALARA) issue, was not an overexposure, did not result in a substantial potential for an overexposure, and did not compromise the ability to assess dose. The inspectors determined this finding has a cross-cutting aspect in the area of Human Performance, Avoiding Complacency, because Exelon did not recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes, and therefore did not implement appropriate error reduction tools. Specifically, Exelon operated the backwash receiving tank (BWRT) to routinely accept high level alarms with associated potential for system overflow. Consequently, although this mode of operation of the system was longstanding, the issue reflects present performance.
05000410/FIN-2015003-012015Q3Nine Mile PointUse of Incorrect Grounding Cart Results in Loss of Electrical BusThe inspectors identified a self-revealing Green finding (FIN) for Exelon Generation Company, LLC (Exelon) personnels failure to stop when met with unexpected conditions as required by procedure HU-AA-101, Human Performance Tools and Verification Practices. On August 21, 2015, a Unit 2 division of normal switchgear was unintentionally deenergized which required an unplanned down power to 90 percent and special operating procedure entry. The loss of the switchgear was the result of installation of an incorrect sized grounding cart in the electric fire pump breaker cubicle during breaker maintenance. Use of the correct sized grounding cart was discussed during the pre-job brief. This resulted in the loss of the electric fire pump, half of the drywell coolers, a heater drain pump, and unplanned reactivity change. Exelon entered this issue into their corrective action program (CAP) for resolution and developed corrective actions which included developing procedures for the use of grounding carts and evaluating where other skill-of-the-craft work may pose the same risk. This finding is more than minor because it is associated with the human performance attribute of the Initiating Events cornerstone and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 1 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, the inspectors determined that this finding is of very low safety significance (Green). The finding has a cross-cutting aspect in the area of Human Performance, challenge the unknown, because Exelon personnel failed to stop when faced with uncertain conditions. Specifically, after having been briefed on the different stab sizes for 1200 amp and 2000 amp grounding carts, Exelon personnel failed to stop and notify supervision when faced with unlabeled grounding carts stored in the same location, Exelon personnel failed to notify supervision or compare stab sizes to ensure the correct grounding cart was used.
05000277/FIN-2015003-022015Q3Peach BottomLicensee-Identified ViolationThe following violation of very low safety significance (Green) was identified by Exelon and is a violation of NRC requirements which meets the criteria of the NRC Enforcement Policy for being dispositioned as a NCV. From 2010 to 2014, PBAPS made a total of 18 shipments of radioactive waste for disposal to the Energy Solutions Clive, UT facility, which contained category 2 levels of radioactive material quantity of concern (RAM-QC), but did not implement transportation security plan for these shipments, which is contrary to the requirements of 10 CFR 71.5 and 49 CFR 172, Subpart I, Safety and Security Plans. This PD adversely affected the Public Radiation Safety cornerstone attribute of Program and Process based on inadequate procedures associated with the transportation of radioactive materials. This issue was documented in Exelons CAP as assignment reports 02484424, 02487034, and 02490534.
05000277/FIN-2015003-012015Q3Peach BottomIncomplete Testing of Components from the Remote Shutdown PanelsThe inspectors identified a Green NCV of Technical Specification (TS) 5.4.1.a after Exelon did not establish and implement procedures to adequately test the Unit 2 and Unit 3 remote shutdown panels (RSPs). Specifically, Exelons surveillance procedure did not test all the control circuits, as required by Surveillance Requirement (SR) 3.3.3.2.1, for the Unit 2 and Unit 3 RSPs. Exelons corrective actions included entering this issue into their CAP, the development of RSP testing procedures for the reactor core isolation cooling (RCIC), control rod drive (CRD), and emergency service water (ESW) system components, and a revision to the bases for TS 3.3.3.2 The performance deficiency (PD) was determined to be more-than-minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Additionally, examples 1.c, 4.l, and 4.m from IMC 0612, Appendix E, detail that a PD was more than minor if required TS surveillance testing is not performed and subsequent testing reveals that the equipment is out of specification or otherwise unable to perform a safety-related function. A detailed risk evaluation concluded that the issue was of very low safety significance (Green). This finding had a cross-cutting aspect in Human Performance, Avoid Complacency, because Exelon failed to recognize and plan for the possibility of latent problems.
05000353/FIN-2015003-022015Q3LimerickInadequate Preventive Maintenance of the HPCI System Motor Control CenterA self-revealing Green NCV of TS 6.8.1.a, Procedures and Programs, occurred when Exelon inadequately maintained and implemented a preventive maintenance (PM) task for the 2DB-1-14 high pressure coolant injection (HPCI) direct current (DC) motor control center (MCC) cubicle. Specifically, PM procedure M-095-002, 250 VDC Westinghouse MCU Maintenance, Revision 6, was performed on the main compartment but was not performed on the auxiliary compartment of the 2DB-1-14 MCC cubicle. Subsequently, the 1A timetactor failed due to lack of cleaning and inspection, which led to a fire in the HPCI DC MCC. Exelons corrective actions included initiating issue report (IR) 2480166, replacing the affected components, and revising the PM task to perform future preventive maintenance on both the main and auxiliary compartments of the 2DB-1-14 cubicle. Exelon also conducted immediate extent of condition reviews and scheduled further reviews to ensure no similar conditions exist. This issue is more than minor because it was associated with the procedures quality attribute of the Mitigating Systems cornerstone, and adversely affected the cornerstone objective to ensure the reliability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, PM procedure M-095-002, 250 VDC Westinghouse MCU Maintenance, Revision 6, was not performed on both compartments of the 2DB-1-14 cubicle and caused the fire in the HPCI DC MCC that had the potential to affect HPCI system operation. Using IMC 0609, Significance Determination Process, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that this finding was of very low safety significance (Green) because the finding was not a deficiency affecting the design or qualification of the HPCI system and the system maintained operability and functionality. Specifically, the affected portions of the HPCI system were a part of the HPCI vacuum tank condensate pump that is not required to ensure operability or functionality. The inspectors determined that the finding did not have a cross-cutting aspect because the PM task change did not occur within the last three years, and the inspectors did not conclude that the causal factors represented present Exelon performance.
05000334/FIN-2014005-022014Q4Beaver ValleyFailure to Properly Ship Category 2 Radioactive MaterialThe inspectors identified an NCV of 10 CFR 71.5, Transportation of licensed material, and 49 CFR 172, Subpart I, Safety and Security Plans. Specifically, FENOC personnel shipped a category 2 radioactive material of concern (RAM-QC) on public highways to a waste processor without adhering to a transportation security plan. FENOCs corrective actions included revising procedure NOP-OP-5201, Shipment of Radioactive Material Waste, to reflect the appropriate Department of Transportation requirements for shipment of Category 2 radioactive material. FENOC entered the issue into their corrective action program as CR 2014-17260. The issue is more than minor because it is associated with the Program and Process attribute of the Public Radiation Safety cornerstone and adversely affected its objective to ensure the safe transport of radioactive material on public highways in accordance with regulations. The finding was determined to be of very low safety significance (Green) because FENOC had an issue involving transportation of radioactive material, but it did not involve: (1) a radiation limit that was exceeded; (2) a breach of package during transport; (3) a certificate of compliance issue; (4) a low level burial ground nonconformance; or (5) a failure to make notifications or provide emergency information. The inspectors determined that the finding did not have a cross-cutting aspect because the issue was not reflective of current plant performance. Specifically, FENOC implemented changes to the radioactive waste shipment procedure that addressed applicable requirements and implemented a formal process for reviewing pending regulatory changes for impacts to FENOC operations and support activities.
05000334/FIN-2014005-012014Q4Beaver ValleyFailure to Adequately Implement Risk Management ActionsThe inspectors identified an NCV of 10 CFR 50.65(a)(4), Requirements for monitoring the effectiveness of maintenance at nuclear power plants, for FENOCs failure to implement adequate risk management actions (RMAs) associated with maintenance on the alternate intake structure A bay. Specifically, FENOC did not establish a contingency plan for the maintenance activity as required by FENOCs risk management procedure. FENOC entered the issue into their corrective action program as CR 2015-00267. The performance deficiency is more than minor because it is associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected its objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, FENOCs failure to implement a contingency plan resulted in an increase in the duration of an elevated risk condition and unavailability of equipment relied upon to mitigate the consequences of a loss of the main intake structure. The finding was determined to be of very low safety significance (Green) because the incremental core damage probability (ICDP) for the event was less than 1.0 E-6. The inspectors determined that this finding had a cross-cutting aspect in the Human Performance, Work Management, because the FENOC work process failed to adequately manage the risk commensurate to the work (H.5).
05000247/FIN-2014004-022014Q3Indian PointFailure to Identify and Evaluate Degraded Condition of the 22 Station Battery CapaciThe inspectors identified a Green NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because Entergy personnel did not adequately implement procedure EN-OP- 104, Operability Determination Process, Step 5.5, to assess the operability and degraded condition of the 22 station battery capacity. Specifically, Entergy personnel did not identify the degraded/non-conforming condition or evaluate the condition relative to support functions for Technical Specification (TS) Surveillance Requirement (SR) 3.8.6.6. The finding was more than minor because it is associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, after inspectors questioned the operability determination, the degraded condition was identified and resulted in the 22 station battery being declared OPERABLE but DEGRADED. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2012, the inspectors determined that the finding was of very low safety significance (Green), because the finding was not a design or qualification deficiency, did not represent a loss of system safety function, and did not screen as potentially risk significant due to seismic, flooding, or severe weather initiating event. Entergy placed this issue into the corrective action program (CAP) as condition report (CR)-IP2-2014-04825 and performed an immediate operability determination followed by a request for an exigent change in TS requirements. The inspectors assigned a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, because Entergy did not thoroughly evaluate the condition of the 22 station battery capacity. Specifically, Entergy did not identify the degraded/non-conforming condition or evaluate the condition relative to support functions for TS SR 3.8.6.6.
05000247/FIN-2014004-012014Q3Indian PointFailure to Maintain Radiation Exposure ALARA During Refueling ActivitiesA self-revealing finding (FIN) of very low safety significance (Green) was identified due to Entergy having excessive unintended occupational collective exposure. This resulted from performance deficiencies in planning and work control while performing reactor coolant pump (RCP) work activities during the Unit 2 refueling outage. Inadequate work planning and control resulted in unplanned, unintended collective exposure due to conditions that were reasonably within Entergys ability to control and prevent. The work activity performance deficiencies resulted in the collective exposure for these activities increasing from the planned dose of 7.269 person-rem to an actual dose of 13.742 person-rem. Entergy entered this issue into their CAP as CR-IP2-2014-02558. The finding was more than minor because it was associated with the Program and Process attribute of the Occupational Radiation Safety cornerstone and adversely affected the cornerstone objective to ensure the adequate protection of the worker health and safety from exposure to radiation. Additionally, the performance deficiency was more than minor based on a similar example (6.i) in Appendix E of IMC 0612; in that, the actual collective dose exceeded 5 person-rem and exceeded the planned, intended dose by more than 50 percent. Entergy placed this issue into the CAP as CR-IP2-2014-02558 and completed a root cause evaluation. The finding had a cross-cutting aspect in the area of Human Performance, Teamwork, in that the work groups did not coordinate activities, which involved job site activities, that adversely impacted radiological safety. Specifically, higher source term due to not delaying the start of work to reduce reactor coolant system (RCS) activity levels following the crud burst and the inability to properly sequence the installation of shielding packages with the work activities resulted in collective exposures that exceeded estimates by greater than 50 percent.
05000334/FIN-2014003-022014Q2Beaver ValleyRemoval of Missile Barrier Renders Containment InoperableThe inspectors identified a Green non-cited violation of TS limiting condition for operation (LCO) 3.6.1, Containment. Specifically, the inspectors determined that FENOC removed the missile barriers for the unit 1 and unit 2 containment equipment hatches while in a mode when containment was required to be operable. As a result FENOC did not have adequate tornado protection for containment and then did not take the actions directed by the LCO action statement when the LCO was not met. FENOC entered the issue into their corrective action program, CR 2014-11878, and placed the procedures to remove the missile barriers on administrative hold. The performance deficiency is more than minor because it adversely affected the configuration 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. In accordance with IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 3, Barrier Integrity Screening Questions, this finding screens to Green, very low safety significance. This finding has a cross-cutting aspect in the area of conservative bias where individuals use decision making-practices that emphasize prudent choices over those that are simply allowable and that a proposed action is determined to be safe in order to proceed, rather than unsafe in order to stop. Specifically, FENOC did not adequately consider the containment operability implications of removing the missile barriers for the unit 1 and unit 2 containment equipment hatches while in a mode where containment is required to be operable. (H14)
05000219/FIN-2014003-012014Q2Oyster CreekFailure to Identify and Correct High Oil Level in D Emergency Service Water Pump Upper Motor Bearing (The NRC inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, because Exelon did not promptly identify and correct a condition adverse to quality. Specifically, Exelon did not identify and correct a high oil level condition caused by water intrusion in the D emergency service water pump upper motor bearing resulting in an inoperable D emergency service water pump. Following identification of the high level by the inspections, Exelon entered this issue into their corrective action program as issue report 1645010. Exelons corrective action included sealing joints on top of the motor that are susceptible to water intrusion. The inspectors determined that inadequate identification and resolution of the condition adverse to quality into the corrective action program is a performance deficiency that was within Exelons ability to foresee and correct. This finding is more than minor because it is associated with the configuration control of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the performance deficiency affected the reliability of an emergency service water pump to perform its safety function. This issue was also similar to Example 3j of NRC IMC 0612, Appendix E, Examples of Minor Issues, because the condition resulted in reasonable doubt of the operability of emergency service water system. The inspectors determined that this finding was a deficiency affecting the design or qualification of a mitigating structure, system, or component (SSC), where the SSC maintained its operability or functionality. Therefore, inspectors determined the finding to be of very low safety significance (Green). The finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, because Exelon did not identify the issue associated with the high oil level in the emergency service water pump upper motor bearing oil in a timely manner in February and April 2014.
05000412/FIN-2014003-012014Q2Beaver ValleyFailure to Follow Procedure Results in Inoperable SI AccumulatorA self-revealing NCV of technical specification (TS) 5.4.1 was identified because the unit 2 B safety injection (SI) accumulator was made inoperable when FENOC operators did not follow procedural requirements to align nitrogen to the accumulator. Specifically, the operators did not align the nitrogen header to the accumulator prior to opening the valve to repressurize the accumulator. The inspectors noted that this resulted in the accumulator pressure falling below the TS pressure limit which required FENOC to declare the accumulator inoperable. FENOCs corrective actions included immediately realigning the system, restoring accumulator pressure and entering the issue into their corrective action program, CR 2014-09260. The performance deficiency is more than minor because it is associated with the configuration control attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, FENOC did not have reasonable assurance that the nitrogen pressure in the B SI accumulator was sufficient to ensure injection into the core during an accident due to the misalignment of the nitrogen header. This finding is of very low safety significance (Green) because the performance deficiency was not a design or qualification deficiency, did not involve an actual loss of safety function, did not represent actual loss of a safety function of a single train for greater than its technical specification allowed outage time, and did not screen as potentially risk-significant due to a seismic, flooding, or severe weather initiating event. This finding has a cross-cutting aspect in the area of Human Performance, Avoid Complacency, because FENOC operators did not recognize the possibility of mistakes and did not implement appropriate error reduction tools while attempting to re-pressurize the B SI accumulator. (H.12)
05000334/FIN-2014003-032014Q2Beaver ValleyLicensee-Identified ViolationTechnical Specification 5.7.2, High Radiation Area, requires, in part, that locked doors be provided for each high radiation area in which the intensity of radiation exceeds 1000 millirem per hour. Contrary to the above, on April 26, 2014, for approximately 2.5 hours, the door to the Regenerative Heat Exchanger room was not locked. FENOCs immediate corrective action included placing chains and padlocks on this door and all similar style entrances to locked high radiation areas, entering this issue into their corrective action program (CR-2014-07646), and performing a root cause evaluation. The finding is of very low safety significance, Green, because it did not involve ALARA, there was no overexposure, there was no substantial potential for an overexposure, and the ability to assess dose was not compromised.
05000220/FIN-2013005-032013Q4Nine Mile PointInadequate DSC Welding Procedure to Control and Monitor Hydrogen ConcentrationsA self-revealing Severity Level IV NCV of Title 10 of the Code of Federal Regulations (10 CFR) 72.150, Instructions, Procedures, and Drawings, was identified when CENG personnel did not ensure that hydrogen concentrations were being properly monitored and maintained during welding on dry shielded container (DSC) #12 on August 14, 2013. Specifically, site procedure S-MMP-ISFSI-004, DSC Sealing Operation, Revision 00201, provided inadequate direction for the control of purging and hydrogen monitoring calibration, set-up, and operation. This caused an undetected loss of DSC purge and a failure of the hydrogen monitor, ultimately resulting in a hydrogen deflagration in DSC #12. CENG staff generated CR-2013-006840 to address the hydrogen deflagration. Corrective actions included: (1) reducing water level in the DSC by 1100 gallons during welding operations to reduce the amount of hydrogen generation; (2) installed dual hydrogen monitors off the vent line to provide redundant indication; (3) required the performance of local hydrogen monitoring at the weld joint prior to commencing welding; (4) reconfigured the location of the hydrogen monitors; (5) ensured hydrogen monitors were properly configured, including the use of the low flow differential pressure switch setting in a helium environment; and (6) adjusted the alarm settings on the hydrogen monitors. The inspectors determined that CENG personnels failure to provide adequate instructions, procedures, and drawings to ensure that hydrogen concentrations were being properly monitored and maintained in accordance with 10 CFR 72.150, Instructions, Procedures, and Drawings, during welding of DSC #12 on August 14, 2013, was a performance deficiency that was reasonably within CENG staffs ability to foresee and correct, and should have been prevented. As a result, a hydrogen deflagration occurred. The failure to properly monitor and maintain hydrogen concentrations had the potential to damage the DSC and spent fuel within the DSC. Because the issue involved independent spent fuel storage installation (ISFSI) operations, consistent with the guidance in Section 2.2 of the NRC Enforcement Policy, the inspectors evaluated this performance deficiency in accordance with the traditional enforcement process. Using Example 6.3.d. from the NRC Enforcement Policy, the inspectors determined that the violation was a Severity Level IV (more than minor concern that resulted in no or relatively inappreciable potential safety or security consequence) violation. The hydrogen deflagration ultimately did not result in the damage to fuel; however, the failure to properly monitor and maintain hydrogen concentrations had the potential to damage the DSC and spent fuel within the DSC. Because the violation involved the traditional enforcement process and was not associated with ISFSI support programs conducted under a 10 CFR 50 license, the inspector did not assign a cross-cutting aspect to this violation in accordance with IMC 0612, Appendix B.
05000410/FIN-2013005-022013Q4Nine Mile PointFailure to Implement Procedural Requirements for Evaluating Control Room Deficiencies as Operator WorkaroundsThe inspectors identified a Green finding (FIN) for CENG staffs failure to properly classify operator workarounds, operator burdens, or control room deficiencies in accordance with CNG-OP-1.01-2010, Operator Workaround/Challenge Control, Revision 0. Specifically, the failure to properly classify operator workarouonds resulted in an operator error when control room operators did not recognize a meter was degraded, used that meter during the performance of a surveillance test, and overexcited the Division II emergency diesel generator (EDG) on July 30, 2013. CENG staff entered this inspector identified issue into the corrective action program (CAP) as condition report (CR)-2013-009004. Corrective actions included reviewing, classifying, and adding the inspector identified operator burdens to each of the respective Units shift turnover checklist. This finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to properly classify the Unit 2 Division II EDG degraded volt amperes reactive (VAR) meter as an operator burden resulted in an operator using the degraded meter during a surveillance test and inadvertently overexciting the diesel generator for 1.5 hours. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2012, the inspectors determined that this finding is of very low safety significance (Green) because the performance deficiency was not a design or qualification deficiency, did not involve an actual loss of safety function, did not represent actual loss of a safety function of a single train for greater than its technical specification (TS) allowed outage time, and did not screen as potentially risk-significant due to a seismic, flooding, or severe weather initiating event. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, in that CENG staff did not ensure control room deficiencies were evaluated properly in accordance with CNG-OP-1.01-2010. Specifically, CENG staff failed to classify the known degraded Unit 2 Division II EDG VARs meter as an operator burden; which resulted in the EDG being overloaded during a surveillance test.
05000220/FIN-2013005-012013Q4Nine Mile PointFailure to Perform Surveillance Test for Unit 1 Smoke Removal DampersThe inspectors identified a Green NCV of Unit 1 license condition DPR-63, Section 2.D(7), Fire Protection, because CENG staff did not perform visual inspections of fire dampers associated with Unit 1 between 2002 and 2013 in accordance with the Fire Protection Program and Updated Final Safety Analysis Report (UFSAR) Section 10A.2.4.1.10.1.A. As a result, CENG staff determined 25 dampers were non-functional due to the surveillance test not being performed. CENG staffs planned corrective actions include revising the UFSAR to state that performance-based testing requirements apply only to non-smoke removal dampers. Further, the 25 smoke removal dampers will remain nonfunctional until visual inspections can be performed as planned in work order (WO) C92482273. This issue was entered into CENGs CAP as CR-2013-009208. This finding is more than minor because it is associated with the structure, system, and component (SSC) and barrier performance attribute of the Barrier Integrity cornerstone and affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the operators in the control room from radionuclide releases caused by accidents or events. 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 for Findings At-Power, issued June 19, 2012, and the inspectors determined that this finding is of very low safety significance (Green) because the performance deficiency only represented a degradation of the smoke removal and radiological barrier function provided for the control room. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, because CENG staff failed to identify smoke removal damper visual inspections were not being performed. Specifically, UFSAR section 10A.2.4.1.10.1.A, as part of license condition DPR-63 2.D(7) and the Fire Protection Program, requires CENG staff to perform visual inspections of smoke removal dampers, which was not being performed between 2002 and 2013, resulting in the control room envelope not being operable and 25 smoke removal dampers being declared non-functional. CENG performed an evaluation to determine if the control room habitability requirements contained in TS 3.4.5.f for the control envelope were met. CENG staff subsequently determined that Unit 1 control room habitability requirements of TS 3.4.5.f were met based on previous successful surveillance testing for control room operability testing under N1-ST-C9, Control Room Emergency Ventilation System Testing, Revision 01502.
05000354/FIN-2013003-012013Q3Hope CreekInadequate Preventive Maintenance Replacement Schedule for Agastat Control RelaysA self-revealing Green NCV of Technical Specifications (TS) 6.8.1, Procedures, was identified because PSEG failed to establish an appropriate preventive maintenance (PM) schedule for Tyco/Agastat General Purpose (GP) control relays. Specifically, the evaluation PSEG performed to revise the relay replacement periodicity from 22 years to 40 years neither adequately addressed available relay references nor all applicable failure mechanisms. As a result, high pressure coolant injection (HPCI) failed to respond to logic system actuation signals during surveillance testing on April 8, 2013. PSEGs immediate corrective actions included replacing failed relays and placing the issues in the corrective action program (CAP). Additionally, PSEG plans to revise the replacement frequency and to replace other Tyco/Agastat GP control relays of high safety significance, as identified in their extent of condition review. This finding was more than minor because it was associated with the equipment performance attribute of the mitigating systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure of a control relay caused the HPCI system to fail to automatically actuate during testing, and the HPCI system was unexpectedly declared inoperable. The inspectors evaluated the finding in accordance with IMC 0609, Significance Determination Process, issued June 2, 2011, and determined the finding is of very low safety significance (Green) following a detailed risk evaluation. No cross-cutting aspect was assigned to this finding because PSEG decisions made with regard to evaluating the PM replacement periodicity were made more than 3 years ago and a PM Ownership Committee has since been created to review PM change evaluations; therefore, this performance deficiency is not reflective of current plant performance.
05000334/FIN-2013004-032013Q3Beaver ValleyLicensee-Identified Violation10 CFR 50.54 Conditions of Licenses, paragraph (q), requires, in part, that licensees maintain an emergency plan that meets the planning standards in 10 CFR 50.47(b) Emergency Plans. 10 CFR 50.47(b)(4) requires use of a standard emergency classification and action level scheme. Contrary to the above, on March 20, 2013, FENOC identified that existing instrumentation was inadequate to assess and determine if abnormal radiological conditions existed such that the Emergency Action Level (EAL) declaration process would not declare an Alert or a Site Area Emergency in an accurate and timely manner. Specifically, the maximum readable values for the containment elevated release radiation monitor low-range channel (RM-1VS-110 Channel 5) and for the cooling tower vent radiation monitor mid-range channel (RM-1GW-109 Channel 7) were less than the EAL threshold values specified for an Alert and Site Area Emergency, respectively, in FENOCs EAL scheme. FENOC entered this issue into their corrective action program as CR-2013-04092. The inspectors determined this finding to be of very low safety significance (Green) in accordance with IMC 0609, Attachment 4, Initial Characterization of Findings, and Section 5.4 of IMC 0609 Appendix B, Emergency Preparedness SDP because the finding was an example of an ineffective EAL, such that an Alert would not be declared and an example of an ineffective EAL, such that a Site Area Emergency would be declared in a degraded manner.
05000334/FIN-2013004-022013Q3Beaver ValleyLicensee-Identified ViolationTechnical Specification 5.4.1, Procedures, requires that written procedures shall be implemented covering the Fire Protection Plan. The FENOC Fire Protection Plan includes 1/2-ADM-1904, Control of Ignition Sources (Hot Work) and Fire Watches, which requires fire watch patrols be completed every seventy-five minutes. Contrary to the above, between June 4, 2013 and August 21, 2013, FENOC identified fire watch patrols were not completed in accordance with the Fire Protection Plan procedure 1/2-ADM- on fifteen patrols. FENOC entered this issue into the corrective action program as CR-2013-08322. The inspectors determined that the finding was of very low safety significance (Green) in accordance with Attachment 4, Initial Characterization of Findings and IMC 0609, Appendix F, Fire Protection Significance Determination Phase 1 Screening, because the reactors were able to reach and maintain safe shutdown conditions.
05000443/FIN-2013004-012013Q3SeabrookInadequate Operability Determination Regarding Service Water Leakage and Associated TS ViolationThe inspectors identified an NCV of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, and an associated violation of technical specification (TS) 3.7.4, because NextEra did not follow the requirements of station procedure EN-AA-203-1001, Operability Determinations/ Functionality Assessments. Specifically, NextEra did not properly evaluate and document an adequate basis for operability, when relevant information was available that would have challenged the reasonable expectation of operability threshold for a service water (SW) through-wall leak that degraded incrementally from weepage on August 7, 2013, to a significantly larger leak on August 28, 2013. NextEra completed a temporary non-code repair of the flaw with the installation of a weldolet on September 1, 2013, following NRC review and approval of a relief request. Additionally, under the corrective action process, NextEra completed apparent cause evaluations for the piping flaw, as well as engineering decision-making during the non-destructive examinations and evaluations, and are currently evaluating the fundamental issue of decision-making regarding TS operability and TS compliance. This performance deficiency is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affected its objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the prompt operability determination incorrectly concluded the B cooling tower (CT) SW header and the B SW (ocean) pumps were operable, but degraded, versus inoperable. IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, and Exhibit 4, External Events Screening Questions, were used to assess this issue and a detailed risk evaluation was completed. The inspectors assumed that functionality of the SW system, based upon the as-found wall thinning, would only be challenged when aligned to the cooling tower basin when the SW piping is subjected to a higher overall sytem pressure. This system configuration is used to mitigate a seismic event following the loss of the normal SW intake structure. Based on low probability of SW piping system failure due to a seismic event and the overall low likelihood of a seismic event of a magnitude sufficient to cause structure, system, and component (SSC) damage, 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 associated with the decision making component because NextEra failed to use conservative assumptions in decision-making and adopt a requirement to demonstrate that the proposed action is safe in order to proceed rather than a requirement to demonstrate it is unsafe in order to disapprove the action. Specifically, NextEra personnel had not considered relevant information in the form of UT data and actual leak propagation to conclude that they no longer had reasonable assurance of operability and did not declare the B header of ocean and CT SW systems inoperable.
05000334/FIN-2013004-012013Q3Beaver ValleyLicensee-Identified ViolationThe Beaver Valley Power Station Unit 2 Technical Specification limiting condition for operation (LCO) 3.5.2 requires two trains of emergency core cooling system (ECCS) to be operable in Modes 1, 2, and 3. Contrary to the above, on June 17, 2013 to June 24, 2013, FENOC failed to have two trains of ECCS operable in Mode 1 which existed for greater than the allowed restoration and shutdown completion times of the LCO due to inadequate procedures that resulted in gas voids in the 21C high head safety injection pump (HHSI) suction piping while the 21A HHSI pump was inoperable due to planned maintenance. FENOC corrective actions include increased void monitoring frequency and updating fill and vent procedures for the HHSI system (CR 2013-09725). In accordance with IMC 0609 Attachment 4, Initial Characterization of Findings, and Exhibit 2 of IMC 0609 Appendix A, The Significance Determination Process for Findings at Power, the inspectors identified that the finding screened as potentially risk-significant due to representing an actual loss of function of a single train for greater than its technical specification allowed outage time. Therefore, a detailed analysis was conducted utilizing the Beaver Valley Unit 2 SPAR model, version 8.23 run by SAPHIRE version 8.0.9. The 21C is a spare pump that can be manually aligned to either train. As a result the analysis considered cases in which it was in the standby configuration and also when it would be required to be manually realigned and started. The first case had a fault exposure time of 204 hours and was assumed to have the pump fail to start if called upon. The second condition had an exposure time of 185 hours and was assumed to fail if realigned. The increase in risk from these conditions resulted in a change in core damage frequency of less than 1E-7. The dominant sequence was a loss of containment air along with failures of reactor coolant pump seals, the ability to provide high pressure injection and the failure of secondary side heat removal. Because an increase in core damage frequency was less than 1E-7, further evaluation of external event and large early release risk was not required and the results calculated were determined to be of very low safety significance (Green).
05000352/FIN-2013003-032013Q2LimerickFailure to Follow Partial Procedure Change ProcessA self-revealing Green finding of TS 6.8, Procedures and Programs , was identified because Exelon personnel did not implement procedure use and adherence requirements when operators changed the scope of work for surveillance testing of main turbine stop and control valves. This resulted in a reactor protection system automatic scram on April 16, 2013. This issue was identified in the Exelon CAP as IRs 1503749 and 1525552. The failure of station operators to follow the partial procedure performance process during the performance of two TS required surveillances was a performance deficiency that was reasonably within Exelons ability to foresee and correct and could have been prevented. The performance deficiency was also contrary to Exelons procedure use and adherence requirements. This finding was more than minor because, if improper implementation of the partial procedure performance process is left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern such as a more severe plant transient or engineered safeguard system actuation or malfunction. Additionally, this issue is similar to example 4.b in IMC 0612, Appendix E, Examples of Minor Issues, in that the procedural error resulted in a reactor scram or other transient. The finding was determined to be self-revealing because it was revealed through the receipt of a scram signal during performance of a surveillance test which required no active and deliberate observation by Exelon personnel. The finding was determined to be of very low safety significance (Green) in accordance with Appendix G of IMC 0609, Shutdown Operations Significance Determination Process, because the finding did not require a quantitative assessment. A quantitative assessment was not required because the finding did not increase the likelihood of a loss of reactor coolant system inventory or degrade the ability to recover decay heat removal if it was lost. This finding had a cross-cutting aspect in the area of Human Performance, Decision Making, because Exelon did not ensure that personnel made safety-significant or risk-significant decisions using a systematic process to ensure that safety is maintained (H.1(a)). Specifically, the partial procedure performance process was not properly implemented which resulted in plant conditions that were improper for the next evolution. This resulted in a reactor protection system automatic scram on April 16, 2013.
05000352/FIN-2013003-022013Q2LimerickFailure to adhere to radiation protection procedures for evacuation of the Unit 2 upper drywell in preparation for irradiated component movesThe inspectors identified a self-revealing finding of very low safety significance associated with failure to comply with TS 6.8, Procedures and Programs. Specifically, the inspectors identified Exelon personnel failed to implement radiation protection procedure requirements associated with clearance of personnel from the upper levels of the Unit 2 reactor drywell in preparation for removal and movement of irradiated core component from the Unit 2 reactor vessel. Exelon personnel entered this issue into their CAP as IR 1495585. The failure to adhere to TS required radiation protection procedures for personnel exposure control related to irradiated core component movement is a performance deficiency. The performance deficiency was determined to be more than minor because it was related to the Programs and Process attribute of the Occupational Radiation Safety Cornerstone, and adversely affected the cornerstone objective to ensure adequate protection of worker health and safety from exposure to radiation from radioactive material during routine reactor operation. Further, if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern if personnel were locked in the area and irradiated hardware dropped above their work location. The finding was not subject to traditional enforcement because it was not associated with a violation that impacted the regulatory process and did not contribute to actual safety consequences. The finding was assessed using IMC 0609, Appendix C, 2 Enclosure Occupational Radiation Safety SDP, dated August 19, 2008, and was determined to be of very low safety significance (Green) because it was not related to As-Low-As-Is-Reasonably-Achievable (ALARA), did not result in an overexposure or a substantial potential for overexposure, and did not compromise the licensee\'s ability to assess dose. This finding was associated with the Work Control aspect of the Human Performance cross-cutting component. Specifically, Exelon staff did not effectively coordinate this work activity by incorporating actions to address the impact of the work on different job activities, and the need for work groups to maintain interfaces and communicate, coordinate, and cooperate with each other during activities in which interdepartmental coordination is necessary to assure plant and human performance.
05000352/FIN-2013003-012013Q2LimerickFailure to Identify and Correct a Condition Adverse to Quality Associated with Emergency Diesel Generator D24The inspectors identified a Green NCV of 10 Code of Federal Regulation (CFR) 50, Appendix B, Criterion XVI, Corrective Action , because Exelon personnel did not identify and correct a condition adverse to quality associated with emergency diesel generator (EDG) D24 lubricating oil pipe fitting supports. This resulted in EDG D24 being in a degraded condition from November 2012 until the condition was corrected in May 2013. Exelon personnel entered this issue into the corrective action program (CAP) as issue reports (IRs) 1507365, 1509125, 1511869, 1512745, 1526780, and 1528088. The failure of Exelon personnel to identify and correct the degraded instrument line pipe fitting support and insert on EDG D24s lubricating oil supply pressure sensing line following the failure of a pipe fitting on November 13, 2012 is a performance deficiency that was reasonably within Exelons ability to foresee and correct. The IR written to document the issue (IR 1439284) was inappropriately classified as not a critical component failure. This resulted in the issue receiving a lower level of investigation (work group evaluation versus an apparent cause or root cause evaluation). This NRC-identified finding was more than minor because it is associated with equipment performance and affected the Mitigating System cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating event to prevent undesirable consequences. The inspectors evaluated the finding using Appendix A, The Significance Determination Process (SDP) for Findings At-Power, to IMC 0609, Significance Determination Process. Exelon personnel conducted vibration testing which determined that the pipe fitting crack initiation and propagation occurred during engine slow start speed acceleration. This was based vibration data which showed two vibration peaks at speeds during the acceleration. Also, the crack did not propagate during normal speed operation based on the fact that the leak size did not increase during monthly testing on April 27, 2013. The inspectors determined this finding did not represent an actual loss of function of a single train for greater than it Technical Specification Allowed Outage Time. Therefore, the inspectors determined the finding to be of very low safety significance.
05000219/FIN-2012005-022012Q4Oyster CreekInadequate Application of Strippable Coating to the Refueling Cavity Liner and the Failure to Configure a Valve in the Leakage Collection System Resulting in Increased Potential for Corrosion on the Exterior of the Drywell Liner Surface in the Sand BedsA self-revealing NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified because Exelon procedures and work orders were not effective in preventing refueling cavity leakage from overflowing onto the exterior surface of the drywell liner during the refueling outage (1R24) in November 2012. The performance deficiencies that contributed to the finding were inadequate oversight of the contractors applying a strippable coating to the reactor cavity liner and a valve configuration control error on a temporarily installed leakage collection system. Upon discovery, Exelon took immediate corrective actions to open the leakage collection system filter inlet valve and restore reactor cavity liner leakage flow to the reactor building equipment drain tank. This finding is associated with the barrier integrity cornerstone and is more than minor because, if left uncorrected, this condition would have the potential to lead to a more significant safety concern. Specifically, the continued wetting of the metallic drywell liner surface could provide an environment conducive to corrosion. This finding is not more than very low safety significance because Exelon performs periodic inspections of drywell liner and exterior surface coating to ensure that liner corrosion is monitored and controlled. The inspectors completed the Phase 1 Initial Screening and Characterization of Findings, of Attachment 0609.04 of Inspection Manual Chapter (IMC) 0609, and screened the finding to Green, very low safety significance. Exelon has entered this condition into the corrective action process under IR 1440116. This finding has a crosscutting aspect in the area of Human Performance, Work Practices, for not ensuring supervisory and management oversight of work activities, including contractors and plant personnel, such that nuclear safety is supported regarding the application of the strippable coating on the reactor cavity liner.
05000219/FIN-2012005-012012Q4Oyster CreekFailure to Follow Inspection and Torquing of Bolted Connection ProcedureThe inspectors identified a Green non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because Exelon did not properly implement procedural controls to ensure adequate thread engagement for standby liquid control (SLC) squib valve spool piece flanges. Specifically, SLC squib valve flanges were installed with inadequate thread engagement (stud was not flush with the nut), as required by Exelons maintenance procedures. Exelons corrective actions included declaring the system inoperable, entering the issue into the corrective action program (IR 1444861 and 1444862) and immediately replacing the existing bolts with bolts of an appropriate length such that projection through the nut was at least flush. The performance deficiency was more than minor because if left uncorrected the inadequate thread engagement would have the potential to lead to a more significant safety concern. Specifically, Exelons evaluation stated that the SLC squib valve spool piece flanges would not have been able to perform their design function under all seismic conditions when the system was required to be operable. In consultation with the Region I senior reactor analyst, the inspectors reviewed this condition using IMC 0609, Attachment G, Shutdown Operations Significance Determination Process. As the condition occurred during the refueling outage and was identified and corrected before Exelon started up the Oyster Creek reactor, and only existed during the outage when SLC was not required to be operable (November 16 27, 2012), the issue screened to very low safety significance (Green). This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, because Exelon did not take appropriate corrective actions to address safety issues and adverse trends in a timely manner, commensurate with their safety significance and complexity. Specifically, Exelon did not take appropriate corrective actions, such as replacing bolts during the refueling outage with longer bolts, after the NRC identified a similar concern on the same SLC squib valve spool piece flanges in September 2012 (IR 1417726).
05000244/FIN-2012005-042012Q4GinnaIncorrect Oil Filter Gasket Installed in the B Main Feedwater Pump Canister CoverA self-revealing Green finding was identified for Ginna personnel not following Constellation procedure CNG-MN-4.01-GL004, Work Package Writers Guideline, Revision 00000, for planning a maintenance activity. Specifically, during the refueling outage, the work package for maintenance on the B main feedwater pump did not identify the correct gasket for the lube oil filter canister; therefore, an incorrect gasket was installed. In addition, maintenance personnel missed an opportunity to prevent the installation of the incorrect gasket when they proceeded after recognizing that the work package was not specific on the gasket required. The gasket failed after being in service for approximately 10 days resulting in a significant oil leak and causing operators to rapidly reduce plant power to 47 percent to remove the pump from service and avoid a plant trip. Immediate corrective actions included replacing the gasket with the correct one and entering this issue into the corrective action program (CAP) as CR-2012-8912. This finding is more than minor because it is associated with the human performance attribute of the Initiating Events cornerstone and adversely impacted the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Additionally, the finding is similar to IMC 0612, Power Reactor Inspection Reports, Appendix E, Examples of Minor Issues, example 4.b in that a personnel error caused a transient. Using IMC 0609, Appendix A, the inspectors determined this 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. Therefore, the inspectors determined the finding to be of very low safety significance (Green). This finding has a cross-cutting aspect in the area of Human Performance, Work Practices, because Ginna personnel proceeded in the face of uncertainty or unexpected circumstances and installed a gasket without confirming it was the correct part.
05000244/FIN-2012005-012012Q4GinnaFailure to Adequately Evaluate Changes to the Relay Room Halon Suppression System Inspection and Testing FrequencyThe inspectors identified a Green non-cited violation (NCV) of Ginna Operating License Condition 2.C.(3), Fire Protection, for failure to adequately evaluate changes to the approved fire protection program that could adversely affect the ability to achieve and maintain safe shutdown in the event of a fire. Specifically, Ginna changed the relay room halon suppression system (S08) inspection and testing frequency from semiannually to biennially and did not appropriately evaluate the change nor properly monitor conditions between testing. As a result, one of the relay room halon system storage cylinders was found below the minimum acceptable pressure. Immediate corrective actions included entering this issue into the CAP as CR-2012-7267, declaring the S08 system non-functional, and establishing a continuous fire watch within 1 hour. This finding is more than minor because if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Specifically, the S08 system was last tested on October 13, 2011, and could have degraded to the point where it could not maintain minimum required halon concentration before it would have been retested and thoroughly inspected in October 2013. Using IMC 0609 Appendix F, a low degradation rating was assigned to this finding because the S08 system was determined to be functional and was expected to display nearly the same level of effectiveness and reliability as it would have had the degradation not been present. Therefore, the inspectors determined the finding to be of very low safety significance (Green). The finding does not have a cross-cutting aspect because the performance deficiency is not reflective of present plant performance.
05000244/FIN-2012005-022012Q4GinnaFailure to Perform an Adequate Extent-of-Condition Review for Water Identified in the Technical Support Center Diesel Fuel Storage TankThe inspectors identified a Green non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (CFR) Part 50 Appendix B, Criterion XVI, Corrective Action, for Ginnas failure to establish measures to assure that conditions adverse to quality are promptly identified and corrected. Specifically, Ginna did not establish measures to promptly identify and correct accumulated water in the B emergency diesel generator (EDG) underground fuel oil storage tank. Subsequently, on November 8, 2012, Ginna identified 1.75 inches of water in the B EDG underground fuel oil storage tank and declared the EDG inoperable. Immediate corrective actions included entering this issue into the corrective action program as CR-2012-7792 and CR-2012-8407, and immediately pumping out, collecting and assessing the amount of water identified in the B EDG underground fuel storage tank. This finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and adversely impacted the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Additionally, the finding is similar to IMC 0612, Appendix E, Examples of Minor Issues, example 3.j., issued August 11, 2009, in that the water identified in the B EDG underground fuel oil storage tank created a reasonable doubt of operability of the B EDG, because the level of water exceeded the operability limit specified in the monitoring plan. Using IMC 0609, Appendix G, Attachment 1, Checklist 4, the inspectors determined this finding did not increase the likelihood of a loss of reactor coolant system (RCS) inventory, did not degrade Ginnas ability to terminate a leak path or add RCS inventory when needed, and did not degrade Ginnas ability to recover decay heat removal once it is lost. Therefore, the inspectors determined the finding to be of very low safety significance (Green). This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, because Ginna personnel did not thoroughly evaluate problems such that the resolutions addressed causes and extent of conditions.
05000244/FIN-2012005-032012Q4GinnaFailure to Meet a Conduct of Operations Standard Results in Loss of Spent Fuel Pool CoolingA self-revealing Green finding was identified for Ginna personnel not following Constellation procedure CNG-OP-1.01-1000, Conduct of Operations, Revision 00700, which requires operators to understand conditions prior to starting equipment. Specifically, Ginna operators inappropriately started the B spent fuel pool (SFP) cooling pump with the SFP low level alarm lit, SFP level decreasing, and the level very close to the pump trip set point. Consequently, 3 hours after being started, the B pump unexpectedly tripped on SFP low level resulting in a loss of SFP cooling. Immediate corrective actions included entering this issue into the CAP as CR-2012-7843, starting the A SFP cooling pump to restore SFP cooling, and adding water to the SFP. This finding is more than minor because it is associated with the human performance attribute of the Barrier Integrity cornerstone and adversely impacted the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Using IMC 0609, Appendix G, Attachment 1, Checklist 4, the inspectors determined this finding did not increase the likelihood of a loss of RCS inventory, did not degrade Ginnas ability to terminate a leak path or add RCS inventory when needed, and did not degrade Ginnas ability to recover decay heat removal once it is lost. Therefore, the inspectors determined the finding to be of very low safety significance (Green). This finding has a cross-cutting aspect in the area of Human Performance, Resources, because Ginna did not ensure that resources were available to assure nuclear safety, specifically those necessary for adequate and available facilities and equipment including physical improvements.
05000317/FIN-2012004-032012Q3Calvert CliffsInadequate Assessment of Unit 1 RCS Pressure Boundary LeakageA self-revealing NCV of Technical Specification (TS) 3.4.13, Reactor Coolant System (RCS) operational LEAKAGE, was identified because Constellation failed to completely isolate a fault in the RCS pressure boundary, which resulted in Constellation operating with RCS pressure boundary leakage for a period of time prohibited by Technical Specifications. Constellations corrective actions included entering the issue in their Corrective Action Program (CAP) (CR-2012-007012 and CR-2012-007276), performing repairs, and conducting root and apparent cause analyses for the issue. The finding is more than minor because it is associated with the equipment performance attribute of the Initiating Events cornerstone and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, after the Constellation personnel identified reactor coolant pressure boundary (RCPB) leakage at 5:15 p.m. on July 17, 2012, they did not reach Mode 3 within six hours because they did not verify complete isolation of the leak. Constellations actions did not limit the likelihood of a small loss of coolant accident (LOCA) event when they operated with RCS pressure boundary leakage from July 17 until July 21, 2012. The inspectors evaluated the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings at Power, and determined the finding is of very low safety significance (Green) because the performance deficiency, after a reasonable assessment of degradation, could not result in exceeding the RCS leak rate for a small LOCA and could not likely affect other systems used to mitigate a LOCA, resulting in a total loss of their function. The finding has a cross-cutting aspect in the area of Human Performance, Decision Making, because Constellation personnel did not use conservative assumptions in decision making and adopt a requirement to demonstrate that the proposed action was safe in order to proceed, rather than a requirement to demonstrate that it is unsafe in order to disapprove the action. Specifically, after attempting to isolate the RCS pressure boundary leakage, Constellation personnel non-conservatively assumed that the leak was isolated based on an inadequate post-isolation verification and monitoring plan.
05000387/FIN-2012004-012012Q3SusquehannaInadequate Procedure for Acts of NatureThe inspectors identified a Green NCV of TS 5.4.1, Procedures, when PPL did not maintain adequate procedures to respond proactively to acts of nature. Specifically, PPLs adverse weather procedure did not ensure timely risk management activities for imminent adverse weather were completed despite a National Weather Service (NWS) declaration of a high wind watch, high wind advisory, and a tornado watch. PPL entered this item in their Corrective Action Program (CAP) as condition report (CR) 1628452. The issue was evaluated in accordance with IMC 0612 and determined to be more than minor since it affected the procedure quality attribute of the Initiating Events cornerstone and its objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the inadequate procedure prevented PPL from taking proactive steps to limit the likelihood of high wind or tornado-related missile hazards upsetting plant electrical power systems. The finding screened to Green in accordance with IMC 0609, Attachment 4, and Appendix A, Exhibit 1, since it did not cause a reactor trip, involve the complete or partial loss of mitigation or support equipment, or impact the frequency of a fire or internal flooding event. The finding was determined to have a cross-cutting aspect in the area of Problem Identification and Resolution - CAP because PPL did not identify issues completely, accurately, and in a timely manner commensurate with their safety significance. Specifically, PPL did not identify that the Off Normal procedure was inadequate both during the 2011 periodic procedural review and during documentation of inspector observations in May 2012 as part of CR 1579977.
05000387/FIN-2012004-022012Q3SusquehannaInadequate Troubleshooting Results in Loss of Secondary Containment and Protected EquipmentA self-revealing Green finding against PPL procedure NDAP-QA-0510, Troubleshooting Plant Equipment, was identified when inadequate troubleshooting caused repeated inoperability of secondary containment, an associated unplanned Unit 2 entry into a 4-hour limiting condition for operation (LCO) action statement, and a loss of the 1C fuel pool cooling (FPC) pump during equipment restoration. The FPC pump had been designated as protected equipment as a risk management action. The failure to perform adequate troubleshooting activities to identify and correct equipment problems prior to restoration was a performance deficiency that was within PPLs ability to foresee and prevent. PPL entered this issue into their CAP as CR 1628250. The inspectors determined that the finding was more than minor because it was associated with the configuration control attribute of the Barrier Integrity cornerstone and adversely affected its objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the event resulted in the inoperability of secondary containment and loss of a FPC pump. The finding was evaluated in accordance with IMC 0609, Attachment 4, and Appendix A - Exhibit 3, and was determined to be of very low safety significance (Green) because the finding did not only represent a degradation of the radiological barrier function provided for the standby gas treatment system and it did not: a) cause the spent fuel pool to exceed a maximum temperature limit; b) cause mechanical fuel damage and detectable release of radionuclides; c) result in the loss of spent fuel pool water inventory; or d) affect spent fuel shutdown margin. This finding is related to the cross-cutting area of Human Performance Decision-Making because PPL did not make safety-significant or risk-significant decisions using a systematic process, especially when faced with uncertain or unexpected plant conditions, to ensure safety is maintained. Specifically, PPL failed to restore equipment in a systematic manner, given the intermittent nature of heater faults, to preclude a repeated loss of protected equipment and secondary containment.
05000387/FIN-2012004-032012Q3SusquehannaFailure to Implement Risk Management ActionsThe inspectors identified a Green NCV of 10 CFR 50.65(a)(4) when PPL did not implement risk management actions (RMAs) during maintenance as required by station procedures. The inspectors identified multiple examples of PPL non-compliance with 10 CFR 50.65(a)(4); PPLs implementing procedures NDAP-QA-0340, Protected Equipment Program; and NDAP-QA-1902, Integrated Risk Management. PPL entered the issue in their CAP as CRs 1611044, 1604007, 1601929, 1602495, and 1611876. The finding was more than minor because it was similar to IMC 0612, Appendix E, examples 7.e and 7.f. Specifically, elevated plant risk required RMAs or additional RMAs that were not implemented as required by plant procedures. The finding also affected the equipment performance attribute of the Mitigating Systems Cornerstone and its objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. In accordance with IMC 0609, Attachment 4, the issues were determined to involve PPLs assessment and management of risk associated with performing maintenance activities and was further assessed under IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management SDP. The issue was evaluated by a Senior Reactor Analyst utilizing flowchart 2, and the finding was determined to be of very low safety significance (Green) since it did not result in an increase to either the incremental core damage probability (ICDP) or to the incremental large early release probability (ILERP). The finding was determined to have a cross-cutting aspect in the area of Human Performance, Work Control, in that PPL did not plan work activities, consistent with nuclear safety, by incorporating risk insights. Specifically, PPL did not incorporate RMAs into its work activities despite recognition of increased risk.
05000387/FIN-2012004-042012Q3SusquehannaLicensee-Identified ViolationOn September 27, 2011, PPL declared the 10-meter wind direction instrument on the primary meteorological tower inoperable when indications showed, as confirmed by the vendor; the wind direction data was inconsistent with known weather responses. EP-TP-007, Equipment Important for Emergency Plan Implementation, states compensatory measures for an out of service Meteorological Tower include notifying the control room of potential Emergency Notification System (ENS) notifications, ensuring the availability of the backup and/or Nescopeck towers, using onsite observations by personnel and obtaining external meteorological information. The control room verified and notified the Nuclear Emergency Response Organization (NERO) Duty Planner that the compensatory measures identified in EP-TP-007 were available. However, the NERO was not notified of the meteorological tower 10-meter wind direction indication being inoperable. In addition, the wind direction indication on the plant computer system continued to display a yellow status color indicating valid data was available for use. Because the NERO was unaware the 10-meter wind direction indication on the primary meteorological tower was erroneous, the inaccurate meteorological information on the plant computer system could have been used by the NERO to make emergency classifications, perform dose projections, and make protective action recommendations (PAR). Although the data from the backup meteorological tower would have been available there were no stimuli that would have caused the NERO to use that data instead. PPLs RCA determined the cause of not notifying the NERO was due to the lack of specific procedural guidance defining the conditions for which the duty NERO personnel should be notified when equipment important to EP was out of service or inoperable. This issue was determined to be a violation of 10 CFR 50.54(q)(2), which requires licensees follow and maintain the effectiveness of an emergency plan that meets the planning standards in 50.47(b). 10 CFR 50.47(b)(9) requires the use of adequate methods, systems, and equipment for assessing and monitoring actual or potential offsite consequences of a radiological emergency condition. Contrary to the above, from September 27 through September 30, 2011, PPL did not maintain an adequate method for accurately calculating dose projections and issuing PARS to offsite agencies. In accordance with IMC 0609, Appendix B, Attachment 2, and the examples contained in Table 5.9-1, the inspectors determined the finding was Green since the meteorological tower was not functional for longer than 24 hours from the time of discovery without adequate compensatory measures. The finding was not greater than Green since the capability for immediate dose projection existed via alternate meteorological towers. The issue was entered in PPLs CAP as CR 1541932.
05000352/FIN-2012004-022012Q3LimerickFailure to Immediately Reduce Reactor Power Per Alarm Response ProcedureThe inspectors identified a cited violation of very low safety significance (Green) of TS 6.8, Procedures and Programs, because Limerick operators did not adequately follow an alarm response procedure when responding to a MCR alarm on July 11, 2012. Specifically, the operators failed to immediately reduce power per the alarm response card (ARC) procedure, ARC-MCR-107-A2, Turbine Control Valve / Stop Valve Scram Bypassed, after the MCR received the alarm condition. The operators decided to delay the immediate reduction in reactor power to validate the control room alarm indication. Overall, it took operators one hour and forty-nine minutes to commence reducing reactor power per procedure. This finding is being cited because not all of the criteria specified in Section 2.3.2.a of the NRC Enforcement Policy for a non-cited violation were satisfied in that Exelon failed to restore compliance within a reasonable amount of time after the violation was identified. Specifically, the violation was communicated to Exelon Management by the inspectors on August 22, 2012. However, this violation was not entered into the Exelon CAP, as IR 1429761, until October 22, 2012 and no interim corrective actions were identified until Standing Order 12-08 was issued on October 22, 2012 to provide operator guidance, 103 days after the initial event. The finding was determined to be more than minor because it affected the human performance 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, it resulted in operators not reducing reactor power immediately as required for reactor protection. The inspectors determined this finding did affect a single RPS trip signal but did not affect the function of other redundant trips or diverse methods of reactor shutdown, did not involve control manipulations that unintentionally added positive reactivity, and did not result in a mismanagement of reactivity by operators. Therefore, the inspectors determined the finding to be of very low safety significance (Green). This finding had a cross-cutting aspect in the area of Human Performance, Work Practices, because operators did not follow procedures
05000333/FIN-2012004-012012Q3FitzPatrickUntimely Corrective Action to Address Crescent Area Unit Cooler OperabilityThe inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, because Fitzpatrick staff did not take timely corrective action to verify that a crescent area unit cooler was operable under postulated conditions of degraded grid voltage. Specifically, Fitzpatrick staff did not schedule first time low voltage pickup testing for unit cooler 66UC-22B until after summer lake temperature had increased to the point that removing the unit cooler from service would have challenged the temperature limit for ultimate heat sink (UHS) operability. When the test was later performed, the as-found pickup voltage exceeded the maximum allowed by the procedure and required a case-specific analysis to demonstrate operability. As immediate corrective action, Fitzpatrick electricians cleaned the contact assembly and retested the unit, with satisfactory results. Fitzpatrick staff entered this issue into the corrective action program as condition report (CR)-JAF-2012-04443. The finding was more than minor because it was similar to example 3.i in Inspection Manual Chapter (IMC) 0612, Appendix E, Examples of Minor Issues, in that a case-specific engineering analysis was required to assure the accident analysis requirements were met. The finding also affected the equipment performance 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. The inspectors evaluated the finding in accordance with IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, and determined that the finding was of very low safety significance (Green) because 66UC-22B maintained its functionality. The finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, because Fitzpatrick staff did not take appropriate corrective actions to address a safety issue in a timely manner, commensurate with its safety significance
05000317/FIN-2012004-042012Q3Calvert CliffsInattentive Non-Licensed OperatorIn accordance with Inspection Procedure 92702, Followup on Traditional Enforcement Actions Including Violations, Deviations, Confirmatory Action Letters, Confirmatory Orders, and Alternative Dispute Resolution Confirmatory Orders, the inspectors conducted a follow-up inspection of a Severity Level IV NCV which was identified due to the deliberate failure of a non-licensed operator to remain attentive to their duties while performing a maintenance evolution on the 2B EDG on June 15, 2011, contrary to Technical Specification 5.4.1.a, Procedures. This issue was communicated to Constellation in a letter dated April 9, 2012, following the completion of an NRC investigation into this matter. The inspectors reviewed the scope and depth of analysis performed in addressing the identified deficiency. The inspectors also reviewed Constellations assessment of generic implications of the identified violation and evaluated the corrective actions implemented by Constellation personnel to determine whether they were adequate to address the identified deficiency and prevent recurrence. The inspectors reviewed Constellations identified causes and the actions taken to prevent recurrence of those causes.
05000317/FIN-2012004-052012Q3Calvert CliffsLicensee-Identified ViolationThe following violation of very low safety significance (Green) was identified by the licensee and is a violation of NRC requirements which meets the criteria of the NRC Enforcement Policy for being dispositioned as a non-cited violation. TS LCO 3.4.10.4, Pressurizer Safety Valves, requires that two pressurizer safety valves shall be operable, which is specifically met, in part, if the as-found setpoints are within applicable acceptance criteria during in-service testing. Contrary to this requirement, on July 7, 2011, Constellation personnel determined that the as-found lift setpoint for PSV, serial number BS03213, exceeded the TS required value by 1 psi. Therefore, Unit 2 PSV BS03213 was determined to be inoperable for an indeterminate period while it had been installed in the plant between March 2007 and March 2009. Calvert Cliffs personnel documented this issue into their corrective action program as CR-2011- 001263. The inspectors evaluated the significance of this finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings at Power. The inspectors determined that this finding was of very low safety significance (Green) because the finding was not a design or qualification deficiency; did not represent a loss of safety system function; and did not screen as potentially risk significant due to external initiating events.
05000317/FIN-2012004-022012Q3Calvert CliffsCorrective Actions Not Completed for Drains in the Intake StructureAn NRC-identified finding of very low safety significance was identified because Constellation staff did not follow Procedure CNG-CA-1.01-1000, Corrective Action Program. Specifically, Constellation staff did not complete corrective actions previously prescribed within their Corrective Action Program as a result of root and apparent cause evaluations for drain failures which impacted safety-related equipment. This resulted in a drain line within the intake structure becoming clogged and the 21 saltwater (SW) pump becoming submerged in water. Constellation personnel entered the issue into their CAP as CR-2012-008363, cleaned out the drain line, and implemented a new preventive maintenance (PM) schedule to keep the drain line clear. Planned corrective actions include overhauling the 21 SW pump bearings. The finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and adversely impacted the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, because the intake structure drain piping was clogged, the 21 saltwater pump pit filled with water and caused the pump bearing housings to be contaminated with water, which adversely impacts the long-term reliability of the pump bearings and will cause the pump to be unavailable while the issue is corrected. The inspectors evaluated the significance of this finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings at Power. The inspectors determined that this finding was of very low safety significance (Green) because the finding was not a design or qualification deficiency; did not represent a loss of safety system function; and did not screen as potentially risk significant due to external initiating events. The inspectors determined that the finding has a cross-cutting aspect in the area of Problem Identification and Resolution because Constellation personnel did not take appropriate corrective actions to address safety issues and adverse trends in a timely manner, commensurate with their safety significance and complexity. Specifically, Constellation personnel did not perform corrective actions previously prescribed to address and correct drain failures that impacted safety-related equipment.
05000317/FIN-2012004-012012Q3Calvert Cliffs2A Diesel Generator Ventilation Train 10 CFR 50.65 (a)(2) Performance Demonstration Not MetAn NRC-identified NCV of 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, paragraph (a)(2), was identified because Constellation personnel did not adequately demonstrate that the 2A diesel generator ventilation train (a)(2) performance was effectively controlled through performance of appropriate preventive maintenance. Specifically, Constellation personnel did not identify and properly account for a functional failure of the 2A emergency diesel generator (EDG) ventilation train in June 2012, and thereby did not recognize that the train exceeded its performance criteria and required a Maintenance Rule (a)(1) evaluation. The subsequent evaluation concluded that the 2A EDG ventilation train (a)(2) performance demonstration was no longer justified and therefore the train should be classified as (a)(1), corrective actions specified, and train monitoring completed. Constellation personnel entered the issue into their CAP as CR-2012- 006132. The finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and affects the cornerstone objective of ensuring the capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, following a functional failure of the 2A EDG ventilation train in June 2012, Constellation did not identify that the train should be monitored in accordance with 10 CFR 50.65(a)(1) for establishing goals and monitoring against the goals. The inspectors evaluated the significance of this finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings at Power. The inspectors determined that this finding was of very low safety significance (Green) because the finding was not a design or qualification deficiency; did not represent a loss of safety system function; and did not screen as potentially risk significant due to external initiating events. The finding has a cross-cutting aspect in the area of Problem Identification and Resolution because Constellation personnel did not thoroughly evaluate the problem such that the resolution fully addressed causes and extent of conditions, as necessary. This includes properly classifying, prioritizing, and evaluating for operability and reportabililty a condition adverse to quality. Specifically, Constellation personnel did not properly evaluate the impact of the condition of the dampers on the ability of the ventilation train to perform its safety function.
05000443/FIN-2012002-022012Q1SeabrookLicensee-Identified ViolationTechnical Specification 3.8.1.1, A.C. Sources, requires as a minimum, the following A.C. electrical power sources shall be operable: (a) two physically independent circuits between the offsite transmission network and the onsite Class 1E Distribution System, and (b) two separate and independent diesel generators. TS 3.8.1.1 Action a. requires that with one of the required offsite AC electrical power sources and one of the required independent diesel generator power sources inoperable, operators must demonstrate the operability of the remaining A.C. source by performing TS surveillance requirement (SR) 4.8.1.1.1a. within 1 hour and at least once per 8 hours thereafter. Contrary to the above on October 31, 2011, for 110 minutes, and from January 10 to 17, 2011, when both the A EDG and one of the required offsite power sources were inoperable, NextEra did not perform TS SR 4.8.1.1.1a. within 1 hour and at least once per 8 hours thereafter. In both instances, the offsite AC source via the RAT was not declared inoperable and the applicable TS action was not entered because NextEra did not recognize the impact of the EDG operation on the fast transfer feature in the TS Bases change process. Specifically, NextEra did not ensure appropriate technical evaluations were performed to review change implications against all normal plant configurations. This finding is of very low safety significance (Green) per IMC 0609 because the issue did not result in a total loss of safety function and did not contribute to both a transient initiator and the likelihood that mitigating functions would be unavailable. Specifically a fast transfer would occur following an opening of the EDG breaker if the bus and RAT were in synchronism. If the bus and RAT were not in synchronism, the RAT breaker would close when residual bus voltage relays actuated. Since the issue is of very low safety significance and was entered into the corrective action program as AR 1718306, the issue is considered a licensee-identified, non-cited violation (NCV) consistent with Section 2.3.2.a of the NRC Enforcement Policy.
05000443/FIN-2012002-012012Q1SeabrookLicensee-Identified Violation10 CFR 50, Appendix B, Criterion XVI, Corrective Action, requires, in part, that measures shall be established to assure that that conditions adverse to quality are promptly identified and corrected. Contrary to the above, NextEra did not assure that a previously identified hold down clamp and missing chaffing material on the B emergency diesel generator (EDG) fuel oil return was promptly corrected. Specifically, as part of an October 2008 extent of condition for a similar condition discovered on the A EDG, the missing clamp and chaffing material was planned to be installed but was not. In November 2011, during a surveillance test of the B EDG, a leak occurred due to vibration induced fretting of the fuel oil return line. Though NextEras engineering and maintenance walkdown procedure provided direction intended to identify the missing clamp and chaffing material, NextEra determined that the direction was ineffective in identifying the condition adverse to quality on the B EDG. This performance deficiency was identified in the corrective action program as Condition Report 1710841. The inspectors determined that this finding was of very low safety significance (Green) because it did not represent an actual loss of safety function or contribute to external event core damage sequences. Since the issue was of very low safety significance and was entered into the corrective action program it is considered a licensee-identified, noncited violation (NCV) consistent with Section 2.3.2.a of the NRC Enforcement Policy.
05000387/FIN-2011005-012011Q4SusquehannaFailure to Properly Implement Work Instructions Results in C EDG InoperabilityAn NRC-identified Green finding of TS 5.4.1, Procedures, due to PPLs failure to properly plan and implement work instructions and Quality Control (QC) hold point inspections associated with a modification to the C Emergency Diesel Generator (EDG) fuel pump assemblies was identified. The error resulted in the failure of the C EDG to continue running during surveillance testing on December 6, 2011. This resulted in PPL failing to meet the requirements of TS 3.8.1, AC Sources- Operating , when it was determined that the C EDG was inoperable from September 19, 2011, following restoration from its maintenance outage, until December 6, 2011, when the operable E EDG was substituted for the C EDG. Additionally, the failure to implement work instructions resulted in PPL failing to meet the requirements of 10 CFR Part 50, Appendix B, Criterion X, Inspection, which requires, in part, that licensees execute a program for inspection of activities affecting quality to verify conformance with the documented instructions, procedures, and drawings for accomplishing the activity. The deficiency was entered into PPLs corrective action program (CAP) as condition Report (CR) 1506105 and a root cause analysis (RCA) was performed. The performance deficiency was determined to be more than minor because the finding was associated with the Mitigating Systems cornerstone attribute of Human Performance, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). The finding was evaluated using Phase 1 and inspectors determined the finding was potentially greater than very low safety significance because the finding represented an actual loss of safety function of a single train for greater than its TS Allowed Outage Time. The Phase 2 analysis determined the finding was potentially greater than very low safety significance given an exposure time of 75 days. A Phase 3 analysis was conducted by an NRC Senior Reactor Analyst (SRA). This analysis indicated an increase in core damage frequency (ACDF) for internal initiating events in the range of 1 core damage accident in 40,000,000 years of reactor operation, in the low E-8 range per year for each unit. The dominant core damage sequences included losses of offsite power with the failure of all EDGs, due to common cause, resulting in a station blackout, followed by operator failure to extend RCIC operation with loss of DC power, failure to depressurize the reactor and failure to recover offsite power within 4 hours. The finding is related to the CCA of Human Performance, Work Practices, in that PPL personnel did not use human error prevention techniques, such as holding pre-job briefings, self and peer checking, and proper documentation of activities, commensurate with the risk of the assigned task, such that work activities are performed safely. Specifically, PPL did not perform adequate human error prevention techniques such that the incorrect assembly of delivery valve springs and stops avoided.
05000247/FIN-2011004-022011Q3Indian PointMarginally Designed Fuse Results in Fuse Failure and lnoperability of the Refueling Water Storage TankThe inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion Ill, Design Control, because Entergy personnel did not establish measures to assure that the design basis for sizing of a fuse was adequate and correctly translated into specifications, drawings, procedures, and instructions. Specifically, between November 29, 2005 and September 13, 2010, the fuse for four control room annuciator panels SASC was marginally sized which resulted in fatigue-induced fuse failure, associated loss of lighting to the annunciator panels, the loss of the refueling water storage tank (RWST) low low level alarms, and the inoperability of the RWST. Entergy personnel immediately replaced the fuse. This issue was entered into Entergy's CAP as CR-IP2-201 0-5713 and CR-IP2-2011-2967. This finding is more than minor because it is associated with the design control attribute of the Mitigating Systems cornerstone and adversely affected the objective to ensure the reliability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the loss of the RWST low low level alarms impacts an alert function relied on by operations personnel to swap the suction of the safety injection pumps from the RWST to the containment sump during accident conditions. Using IMC 0609.04, Phase 1 -Initial Screening and Characterization of Findings, the inspectors determined this finding was of very low safety significance (Green) because the finding was related to a design or qualification deficiency confirmed to result in a loss of operability of the RWST low low level alarms; however, the finding did not represent a loss of safety system function because RWST level indication was available via redundant level instruments on the control room instrument panel that operators also normally rely on and are trained to use. Also the finding did not screen as potentially risk significant due to external initiating events. The finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program attribute because Entergy personnel did not thoroughly evaluate problems associated with the fuse for control room annunciator panels SA-SC, such that the resolution address causes and extent of conditions, as necessary. This includes properly classifying, prioritizing, and evaluating for operability and reportability conditions adverse to quality.
05000247/FIN-2011004-012011Q3Indian PointWater Intrusion in the 480 Volt Room During Hurricane IreneDuring Hurricane Irene's impact at Indian Point on August 28, 2011, operations personnel identified water intrusion in the 480 volt room. Water was entering the room through the seals around SW piping that penetrated the wall between the transformer yard and the 480 volt room. Operations personnel identified that the drain nearest to the water intrusion was plugged, and used a catch basin to direct the water to another drain. Operations personnel also placed sandbags around the 480 volt switchgear. The inspectors walked down the area during the hurricane and determined no water impacted the operation of the 480 volt switchgear. The inspectors are opening an URI to review the licensee's evaluation of the causes of the water intrusion into the 480 volt room and determine if there is a performance deficiency. Entergy personnel wrote CR-IP2-2011-4324 to address this issue.
05000353/FIN-2011004-012011Q3LimerickFailure of Feedwater MOV Resulting in RCIC Inoperability for Longer than Allowed by Technical SpecificationsA self-revealing preliminary white finding and apparent violation of Technical Specification (TS) 3.7.3, Reactor Core Isolation Cooling System and TS 3.6.3, Primary Containment Isolation Valves, was identified. The inspectors determined that the failure by Exelon to ensure sufficient technical guidance was contained in operating procedures to: 1) ensure that a Main Feedwater system (FW) motor-operated valve (MOV) could close against expected system differential pressures and 2) prevent operators from attempting to close FW MOVs out of sequence resulting in differential pressures for which they are not designed; is a performance deficiency. This resulted in the Reactor Core Isolation Cooling system (RCIC) and a Primary Containment Isolation Valve (PCIV) being inoperable from April 23 to May 23, 2011, due to FW MOVs HV-041-209B and HV-041-210 failing to fully shut. As a result, both safety related systems were inoperable for greater than their Technical Specification allowed outage times. Specifically, operations procedures did not contain adequate technical guidance to ensure that operations personnel operated HV-041209 A&B and HV-041-210 in the proper sequence to remain within valve design limitations. This resulted in the HV-041-209B and HV-041-210 valves failing to fully close on April 22, 2011, although they indicated closed in the Main Control Room. Upon identification, Limerick operations staff fully closed the valves restoring RCIC and PCIV operability, entered the issue into the CAP as issue report (IR) 1219476 and conducted a cause evaluation. Subsequent corrective actions included an extent-of-condition review, revisions to the operating procedure, and revisions to maintenance and testing procedures. The inspectors determined that this finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, operating procedures, maintenance and testing were not adequately implemented to ensure that the design capability of HV-041-209B and HV-041-210 to close against expected system differential pressures was maintained. The finding was evaluated using NRC Inspection Manual Chapter 0609 Appendix A, User Guidance for Significance Determination of Reactor Inspection Findings for At-Power Situations. Phase I, II, and III evaluations were conducted. The NRC total estimated L\\\\CDF in this preliminary assessment is Low E-6/yr (WHITE) and the NRC total estimated Large Early Release Frequency (LiLERF) in this preliminary assessment is 3.6E-9/yr (GREEN). The inspectors also determined that this issue has a cross-cutting aspect in the area of Human Performance, Resources, because Exelon did not ensure long term plant safety by maintaining design margins and minimizing preventive maintenance deferrals (H.2. (a)). Specifically, design limitations of the HV-041209 A & B valves were not adequately captured in the procedural guidance, which contributed to the operators continuing on in the procedures for securing the FW long path recirculation line up when problems with the HV-041-21 0 valve were encountered. Additionally preventive maintenance activities which could potentially have prevented this issue were deferred without an appropriate evaluation.
05000352/FIN-2011004-042011Q3LimerickLicensee-Identified Violation10 CFR 50.54(q) requires, in part, that a power reactor licensee follow an emergency plan that meets the requirements of 10 CFR 50.47(b) and Appendix E to 10 CFR Part 50. Contrary to the above, Exelon did not make timely notification when the emergency action level threshold was met for HU5, Natural and Destructive Phenomena Affecting the Protected Area. Specifically, Exelon operators did not declare an Unusual Event within the required fifteen minutes of the earthquake felt onsite on August 23. The actual declaration was nine minutes late. At 1:51 PM, control room operators received a Seismic Monitor System Recording Activated alarm coincident with reports of seismic activity felt by station personnel. The seismic monitoring system at Limerick had previously been declared inoperable due to problems with its power supply, so operators began the compensatory measures which directed the operators to contact the United States Geological Survey to confirm the epicenter and magnitude of the seismic event prior to event classification. The United States Geological Survey has a call queue system to answer inquiries in an orderly manner, and Exelon was on hold until 2:11 PM. Exelon declared the Unusual Event at 2:15 PM and made all appropriate state and local notifications. Exelon entered the untimely event declaration into their corrective action program as IR 1254845. The inspectors determined that the finding was of very low safety significance (Green) in accordance with NRC IMC 0609, Appendix B, Emergency Preparedness Significance Determination Process, Sheet 2, because this was related to an actual event implementation problem for a Notice of Unusual Event.
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