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| Q | Significance | CCA | Identified by | Title | Description | |
|---|---|---|---|---|---|---|
| 05000324/FIN-2018411-01 | 2018Q3 | Green | NRC identified | Security | ||
| 05000324/FIN-2018002-02 | 2018Q2 | Severity level Enforcement Discretion | NRC identified | Enforcement Action 18-080: Implementation of EGM 11-003, Revision 3, Enforcement Guidance Memorandum on Dispositioning Boiling Water Reactor Licensee Noncompliance with Technical Specification Containment Requirements during Operations with a Potential for Draining the Reactor Vessel (OPDRV) | During the Unit 1 spring 2018 refueling outage, the OPDRVs activities are listed below: March 7, 2018: 148 gallons per minute (gpm) leakage associated with local leak rate testing (LLRT) of valves 1-G31-F001 and -F004.March 8, 2018: 82 gpm leakage for A RHR loop draining to support maintenance.March 12, 2018: 81.2 gpm leakage to replace the local power range monitors and intermediate power range dry tubes.March 14, 2018: 71.2 gpm leakage to replace the local power range monitors.March 15, 2018: 25 gpm leakage to replace A recirculation pump seal package.March 22, 2018: 25 gpm leakage to replace A recirculation pump seal package.March 27, 2018: 164 gpm leakage to facilitate control rod drive system venting.March 28, 2018: 288 gpm leakage to account for leakage past scram discharge and vent valves during testing.These activities took place without secondary containment being operable. Corrective Actions: EGM 11-003 allows enforcement discretion regarding secondary containment operability during Mode 5 OPDRV activities provided the licensee meets certain requirements. The licensee met the stipulations of the EGM by executing their procedure 1SP-16-100, EGM 11-003 OPDRV Activities, Rev 001, for each OPDRV activity during the Unit 1 Spring 2018 refueling outage. Additionally, as required by the EGM, the licensee submitted a license amendment request (BSEP 17-0060) on June 29, 2017. The amendment was approved on April 13, 2018, and will be implemented prior to the 2019 Unit 2 spring refueling outage. Corrective Action Reference: The issue was entered into the licensees corrective action program as NCR 2189536. Violation: TS 3.6.4.1, Secondary Containment, requires that secondary containment be operable and is applicable during OPDRVs. The required action if secondary containment is inoperable in this condition is to initiate actions to suspend OPDRVs immediately. Contrary to the above, on activities listed above, the licensee failed to maintain secondary containment operable while performing OPDRVs on Unit 1. Severity/Significance: According to EGM 11-003, the NRC considers enforcement discretion related to secondary containment operability during Mode 5 OPDRV activities provided the licensee meets certain requirements such as monitoring vessel level, maintaining capability to isolate leakage paths, providing minimum makeup flow rate, etc. These requirements provide a reasonable assurance of public health and safety during draining activities in Mode 5 while the secondary containment is inoperable 13 Enclosure Discretion Basis: The NRC exercised enforcement discretion in accordance with Section 3.5, Violations Involving Special Circumstances, of the NRC Enforcement Policy and, therefore, will not issue enforcement action for this violation. These violations were identified during the discretion period described in EGM 11-003, Revision 3, and the licensee met the criteria established in the EGM prior to and during these activities. | |
| 05000325/FIN-2018002-01 | 2018Q2 | Green | H.3 | Self-revealing | Automatic Reactor Trip due to Perceived Loss of Stator Cooling Water | A self-revealing Green finding (FIN) was identified for the failure to properly implement a modification to the turbine control system (TCS). The modification ultimately resulted in an automatic reactor trip on April 7, 2018, due to a turbine trip caused by a perceived loss of stator cooling water. The TCS system improperly generated a loss of stator cooling turbine trip when the TCS measured higher than expected stator cooling water flow rates |
| 05000324/FIN-2018011-03 | 2018Q2 | Licensee-identified | Potential Unjustified Activation Energy for Rosemont Transmitters | The licensee used 0.78 eV as the limiting activation energy for Rosemount transmitters. The activation energy was based upon an academic paper documenting experimental work performed for the early space program and first published in 1965. The paper cautioned the reader that the methods used were experimental and were not validated. A 0.5 eV activation energy for electronics was documented by the Electric Power Research Institute (EPRI) report NP-1558, which attributed it to electron migration of aluminum. Reports published by the Institute of Electrical and Electronics Engineers (IEEE) indicated that activation energies for various electronics and their failure modes could range from 0.5-0.66 eV. The licensee did not document an independent failure modes and effects analysis to justify the activation energy that they used. In addition, the licensee chose to use less limiting activation energies that were not proven to be justified. Finally, the licensee was unable to demonstrate acceptable margins for extrapolation confidence. The IEEE standard 323-1974, section 6.5.2, Mathematical Modeling, stated, the first step in the qualification by analysis is generally the construction of a valid mathematical model of the electric equipment to be qualified. The mathematical model shall be based upon established principles, verifiable test data, or operating experience data. The mathematical model shall be such that the performance of the electric equipment is a function of time and the pertinent environmental parameters. All environmental parameters listed in the equipment specification must be accounted for in the construction of the mathematical model unless it can be shown that the effects of the parameter of interest are dependent on the effects of the remaining environmental parameters. Planned Closure Actions: The team must determine whether the activation energy used for the transmitters was appropriate and, if not, whether the licensee had the responsibility to verify the information provided by their vendors and contractors. The region is discussing this issue with NRC headquarters to find a resolution to this issue. | ||
| 05000324/FIN-2018011-02 | 2018Q2 | Green | NRC identified | Failure to Evaluate Effects of MOV Space Heaters on Qualified Life | The NRC identified a Green finding and associated non-cited violation of 10 CFR 50.49(e)(5) for the licensees failure to evaluate the effects of additional heat rise on the qualified life of Limitorque controls. | |
| 05000324/FIN-2018011-01 | 2018Q2 | Green | NRC identified | Failure to Justify Qualified Life Extension of ASCO Solenoid Operated Valves | The NRC identified a Green finding and associated non-cited violation of 10 CFR 50.49(e)(5) for the licensees failure to justify life extensions of ASCO solenoid operated valves (SOVs | |
| 05000324/FIN-2018001-01 | 2018Q1 | Green | Self-revealing | Inadequate Instruction to Perform Inspections on Emergency Ventilation Dampers | A self-revealing Green NCV of TS 5.4.1a, Procedures, was identified when the licensee failed to properly provide adequate work instructions associated with the control room emergency damper inspections. Specifically, the licensee disconnected the damper air supply line without adequate work instruction guidance, which caused a loss of Control Building Heating, Ventilation and Air Conditioning (HVAC) and Control Room Emergency Ventilation (CREV) Systems resulting in a safety system functional failure. | |
| 05000324/FIN-2017004-01 | 2017Q4 | Green | H.8 | Self-revealing | Loss of Emergency 4160V Bus Due to Failure to Implement Procedure | A self-revealing non-cited violation (NCV) was identified for the licensees failure to properly transfer power to the E-4 4160 volt emergency bus from the E-4 emergency diesel generator (EDG), to the normal switchgear bus 2C, as required by procedure 0OP-50.1 Diesel Generator Emergency Power System Operating Procedure. This resulted in a momentary under voltage condition followed by a re-energization of the E-4 emergency bus by EDG-4. This was entered into the licensees corrective action program (CAP) as nuclear condition report (NCR) 2151329.The licensees failure to parallel across (i.e., reclose) the normal feeder breakers prior to unloading the EDG-4 and opening the EDG-4 output breaker, which resulted in a valid and automatic actuation of the EDG-4, was a performance deficiency. The finding was determined to be greater than minor because it was associated with the human performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during power operations. Using IMC 0609.04, Initial Characterization of Findings, Exhibit 1, the issue was classified as a transient initiator contributor because it was associated with a loss of offsite power (LOOP). Finally, using Appendix A of IMC 0609, SDP for Findings at-Power, the finding was determined to be of very low safety significance (Green) because it did not contribute to both the likelihood of a reactor trip and the likelihood that mitigating systems would not be available. Using Manual Chapter 0310, Aspects Within the Cross-Cutting Areas, the inspectors identified a cross-cutting aspect in the procedural adherence of the human performance area, because the operators failed to properly utilize an existing procedure pertinent to their particular situation and this directly resulted in the momentary loss of an emergency 4160 volt bus. (H.8) |
| 05000324/FIN-2017004-02 | 2017Q4 | Green | Licensee-identified | Licensee-Identified Violation | The following violation of very low safety significance (Green) was identified by the licensee and is a violation of NRC requirements which met the criteria of the NRC Enforcement Policy, for being dispositioned as a NCV. Unit 1 and Unit 2 facility operating license DPR-71 and DPR-62 condition 2.B.(6)requires, in part, that the licensee shall implement and maintain in effect all provisions of the approved fire protection program. Procedure AD-EG-ALL-1522, Duties of a Fire Watch, requires periodic fire watches to be performed within their designated time periods including any allowed grace periods. Contrary to the above, during the spring 2017 Unit 2 refueling outage, between March 1 and March 19, selected periodic fire watches were missed or not performed within the required grace periods. The finding was screened using IMC 0609, Appendix F Fire Protection Significance Determination Process, and was determined to be of very low safety significance (Green), because the reactor was able to reach and maintain safe shutdown. This issue was documented in the licensees CAP as NCR 2115035. | |
| 05000324/FIN-2017003-01 | 2017Q3 | Green | Licensee-identified | Licensee-Identified Violation | 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, states in part that activities affecting quality shall be accomplished in accordance with these instructions, procedures, or drawings. Contrary to the above, from August 2015 until April 2017, Unit 2 SRV pilot valves did not incorporate precision grinding to remove micro -cracking layer as described in licensee procedure OCM -VSR509 , Main Steam Relief Valves Target Rock Model 7567 Air Operators and Pilot Assembly, Disassembly , Inspection, and Reassembly. This resulted in 3 of the 11 SRVs being out of tolerance. Since less than 10 SRVs were operable, Unit 2 operation was prohibited by TS 3.4.3. The licensee took corrective action to replace all of the pilot valves with the correct surface finish. This violation was determined to be of very low safety significance (Green) because the violation did not represent a loss of safety function since this condition was supported by the Brunswick Unit 2 Cycle 22 Reload Safe ty Analysis. Specifically, the analysis concluded that with at least five total SRVs operable, the overpressure safety function would not be challenged. The licensee entered this issue into their CAP as CR 2129416. | |
| 05000324/FIN-2017002-02 | 2017Q2 | Green | Licensee-identified | Licensee-Identified Violation | LCO 3.0.4 states , in part, when an LCO is not met, entry into a MODE or other specified condition in the applicability shall only be made when the associated ACTIONS to be entered permit continued operation in the MODE for an unlimited period of time. Contrary to the above, on April 13, 2017, at 2347, the licensee changed modes on Unit 2 from Mode 4 to Mode 2 with the primary containment inoperable. Specifically, TS 3.6.1.1, Primary Containment, requires primary containment to be operable in Modes 1, 2, and 3, however, the reactor was taken to Mode 2 with the primary containment still aligned for venting. In this condition, the drywell ventilation makes the primary c ontainment inoperable due to the drywell and the suppression chamber being in communication to each other. This condition is allowed for two hours before the required action to shutdown to Mode 4 is required. Since the required action is not permitted for continued operation, a violation of LCO 3.0.4 occurred. The cause of the event was the failure of operations personnel to initiate a tracking LCO for primary containment being inoperable in accorda nce with licensee procedure 0OI -01.01, BNP Conduct of Operations Supplement. The licensee determined a tracking LCO should have been initiated to ensure the primary containment was restored to operable prior to entering Mode 2. As immediate corrective actions, the licensee restored containment to operable on April 14, 2017 , at 0030, and the Shift Manager and Control Room Supervisor were removed from shift. Further, a remediation plan was developed and implemented to assess the Shift Manager and Control Room Supervisor for watch standing and reinstatement . This finding is associated with the Barrier Integrity Cornerstone. Using IMC 0609, Appendix G, Shutdown Operations Significance Determination Process, Attachment 1, Exhibit 4, Barrier Integrity Screening Ques tions, the finding screened to G reen because it did not degrade the ability to close or isolate the containment, it did not degrade the physical integrity of reactor containment, and it did not involve an actual reduction in function of hydrogen control . This issue was documented in the licensee s CAP as NCR 2116753. | |
| 05000324/FIN-2017404-01 | 2017Q2 | Green | NRC identified | Security | ||
| 05000324/FIN-2017002-01 | 2017Q2 | Severity level Enforcement Discretion | NRC identified | Implementation of EGM 11- 003, Revision 3, Enforcement Guidance Memorandum on Dispositioning Boiling Water Reactor Licensee Noncompliance with Technical Specification Containment Requirements During Operations with a Potential for Draining the Reactor Vessel | Implementation of EGM 11- 003, Revision 3, Enforcement Guidance Memorandum on Dispositioning Boiling Water Reactor Licensee Noncompliance with Technical Specification Containment Requirements During Operations with a Potential for Draining the Reactor Vessel a. Inspection Scope The inspectors reviewed the plants implementation of NRC EGM 11 -003, Revision 3, during Unit 2 maintenance activities for operations with a potential for draining the reactor vessel (OPDRVs) , during the Unit 2 refueling outage. Inspectors verified that for all dates, all other TS requirements were met during OPDRVs with secondary containment inoperable. Documents reviewed are listed in the Attachment. b. Findings Description . During the Unit 2 refueling outage, the OPDRVs activities are listed below: March 21, 2017: 6 gallons per minute leakage for maintenance associated with the 2B recirculation pump seal rebuild March 23, 2017: 48 gallons per minute leakage for hydraulic control unit draining to support rod maintenance March 24, 2017: 65 gallons per minute leakage to close the excess flow check valve and cap the drain line March 27, 2017: 71.2 gallons per minute leakage to replace the low power range monitors and intermediate power range monitors These activities took place without secondary containment being operable. Enforcement . TS 3.6.4.1, Secondary Containment, requires that secondary containment be operable and is applicable during OPDRVs. The required action if secondary containment is inoperable in this condition i s to initiate actions to suspend OPDRVs immediately. Contrary to the above, on March 21, 2017, March 23, 2017, March 24, 2017 , and March 27, 2017 , the licensee failed to maintain secondary containment operable while performing OPDRVs. However, because the violations were identified during the discretion period described in EGM 11 -003, Revision 3, and the licensee met the criteria established in the EGM prior to and during these activities, the NRC exercised enforcement discretion (Enforcement Action -17- 123 ) for the dates of March 21, 2017, March 23, 2017, March 24, 2017, and March 27, 2017, in accordance with Section 3.5, Violations Involving Special Circumstances, of the NRC Enforcement Policy and, therefore, will not issue enforc ement action for this v iolation. The inspectors observed that Brunswick has already submitted a license amendment request (BSEP 17- 0060) on June 29, 2017 which was accepted for review by the NRC on July 18, 2017. The licensee entered this issue into the CAP as NCR 2110409. | |
| 05000325/FIN-2017009-01 | 2017Q2 | Green | Self-revealing | Inoperability of EDG1 due to Cyclic Fatigue Failure of Hydraulic Fuel Rack Control | Green . A self -revealing Green non- cited violation ( NCV ) of 10 CFR 50 Appendix B Criterion XVI, Corrective Actions, was identified on February 19, 2017, when emergency diesel generator ( EDG ) number one was determined to be inoperable due to an oil leak o n the linkshaft hydraulic control assembly. This violation of regulatory requirement existed from October 27, 2015 u ntil February 20, 2017. The licensee entered this issue in their corrective action program as nuclear condition report ( NCR) 02101084. The inspectors determined that the finding was more than minor because it was associated with the Mitigating Systems cornerstone attribute of Equipment Performance 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, failu re to correct a condition adverse to quality led to the inoperability of EDG1. The inspectors screened this finding using IMC 0609, Appendix A, The Significant Determination Process (SDP) For Findings At -Power, dated June 19, 2012, Based on Exhibit 2, Q uestion A3, the inspectors determined that a detailed risk evaluation was necessary given the uncertainty over how long EDG1 would have operated while leaking oil. A regional senior reactor analyst (SRA) conducted the risk assessment and screened the issu e to Green based on an increase in risk of less than 1E -6. The inspectors determined that this finding did not have an associated cross cutting aspect because this finding was not reflective of current licensee performance due to enhancements of site procedures guiding creation of work orders. | |
| 05000324/FIN-2017001-05 | 2017Q1 | Green | Licensee-identified | Licensee-Identified Violation | TS limiting condition for operation (LCO) 3.3.3.1, Condition F, Post Accident Monitoring (PAM) Instrumentation, states in part , with the DWHRRMs inoperable, a Special Report shall be submitted to the Commission within the next 14 days. Contrary to the above, the licensee failed to identify the inoperability of the DWHRRMs after the NRC Information Notice 97 -45 Supplement 1 was issued. In particular, the DWHRRMs signals cables are susceptible to thermally induced currents which can degrade the accuracy of DWHRRMs . The required action of LCO 3.3.3.1 , action F, was not perf ormed from 1998 until December 5, 2016. Using IMC 0609, Appendix B, Emergency Preparedness Significance Determination Process, inspectors determined that this violation was of very low safety significance (Green) because the finding is related to Emergenc y Preparedness Requirements that are not associated with a planning standard function (e.g.,10 CFR 50.54(q), 10 CFR 50.54(t), and requirements in Appendix E to 10 CFR Part 50 that do not support a planning standard function ). Other parameters could be used to validate the indications from the DWHRRMs . The corrective action is to restore the monitors to operable. This issue was documented in the licensees CAP as NCR 2066681 | |
| 05000324/FIN-2017007-01 | 2017Q1 | Green | P.2 | NRC identified | Failure to Correct a Nonfunctional Fire Door | Green. The NRC identified a Green non-cited violation (NCV) of Brunswick Operating License Condition (OLC) 2.B(6) for Units 1 and 2, for the licensees failure to correct a nonfunctional fire door in the diesel generator (DG) building. Specifically, on three occasions, NRC inspectors identified door 2-DGB-DR-E L023-118 in the DG building as having a stuck open latch, which prevented the door from closing and latching securely. U pon the third discovery of the nonfunctional fire door, the licensee initiated AR 02100405, entered the appropriate action statement in accordance with site procedure 0PLP-01.2, Fire Protection System Operability, Action, and Surveillance Requirements, and took actions to install a new thumb latch, and to install a new door closure mechanism. The inspectors determined that the licensees failure correct nonfunctional fire door was a performance deficiency (PD). The PD was determined to be more than minor because if left uncorrected, the PD could have the potential to lead to a more significant safety concern. Specifically, if the door was not repaired adequately, it could have the potential to not be able to perform its design function in the case of a fire in diesel generator cell nos. 1 or 2 (FA DG-4 or DG-5). Using IMC 0609, Appendix F, Attachment 1, Fire Protection Significance Determination Process Worksheet, the finding was screened as Green at task 1.4.3-C because there was a fully functional automatic suppression system on at least one side of the fire barrier. The finding has a cross-cutting aspect in the area of problem identification & resolution associated with the Evaluation attribute because the organization did not thoroughly evaluate the condition of the door to ensure that the resolution addressed the underlying cause of the nonfunctional fire door (PI.2) |
| 05000324/FIN-2017001-02 | 2017Q1 | Green | H.4 | NRC identified | Failure to Control a Temporary Fire Ignition Source Near the Unit 2 Standby Liquid Control Pump Motor and Cables | Green . An NRC- identified Green NCV of License Condition 2.B.(6), Fire Protection Program, was identified for the licensees failure to adequately control fire ignition sources in the Unit 2 standby liquid control (SLC) pump ar ea in accordance with licensee procedure AD -EG -ALL -1523, Temporary Ignition Source Control. Specifically, between January 7, 2017, and January 13, 2017, a temporary electric portable heater was energized 2 feet from an SLC pump motor without continuously attending the temporary ignition source or obtai ning a continuous fire watch. The licensees c orrective actions included turning off the heater and removing it from near the SLC pumps. This issue was entered into the licensees CAP as NCR 2091736. The inspectors determined that the licensees failure to control fire ignition sources in accordance with licensee procedure AD -EG -ALL -1523, was a performance deficiency. The finding was more than minor because it was associated with the Protection Against External Events attribute (i.e. fire) of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the temporary ignition source could have affected a nearby safety -related SLC pump motor and cables, which provide a shutdown mitigation function. The finding was screened using NRC IMC 0609, Appendix F, Fire Protection Significance Determination Proc ess, dated September 20, 2013. Using IMC 0609, Appendix F, Attachment 1, Fire Protection SDP Phase 1 Worksheet, dated September 20, 2013, the findi ng was assigned to the Fire Prevention and Administrative Controls category because the portable heater is part of the plants combustible materials control program. Proceeding to Task 1.3.1 of IMC 0609, Appendix F, Attachment 1, the inspectors determined the finding was of very low safety significance (Green), because even if one train of SLC had been inoperable, the reactor was able to reach and maintain safe shutdown. This finding had a cross cutting aspect in the area of human performance associated wi th the teamwork aspect because individuals failed to effectively communicate and coordinate their activities to ensure that the temporary heaters were energized following prescribed fire protection control measures and written instructions. (H.4) |
| 05000324/FIN-2017001-01 | 2017Q1 | Green | H.2 | NRC identified | Nonfunctional Sprinklers in the Service Water Building Without Compensatory Measures | Green . An NRC- identified Green non -cited violation ( NCV ) of License Condition 2.B.(6), Fire Protection Program, was identified for the licensees failure to implement compensatory measures for nonfunctional sprinklers. Specifically, from January 11, 2017, until January 14, 2017, fire sprinklers were impaired when scaffold ing was built over the service water (SW) system discharge valves without the proper fire protection evaluation and compensatory measures , as required by licensee procedure 0PLP -01.2, Fire Protection System Operability, Action, and Surveillance Requirements . The licensees corrective actions included declaring the sprinklers nonfunctional, and implementing an hourly fire watch and backup suppression until the scaffold could be removed. This issue was entered into the licensees corrective action program (CAP) as nuclear condition report (NCR) 2091795. The inspectors determined that the licensees failure to implement compensatory measures for nonfunctional sprinklers , in accordance with procedure 0PLP -01.2, was a performance deficiency. The finding was more than minor because it was associated with the Protection against E xternal Events attribute (i.e. fire) of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, this resulted in nonfunctional sprinklers in a safety -related area without compensatory measures. The finding was screened using NRC IMC 0609, Appendix F, Fire Protection Significance Determination Process, dated September 20, 2013, because the finding affected the fixed fire protection system capability. Using IMC 0609, Appendix F, Attachment 1, Fire Protection S DP Phase 1 Worksheet, dated September 20, 2013, the finding was assigned to the Fixed Fire Protection System category because the nonfunctional sprinklers affected the automatic fire suppression system. Proceeding to Task 1.3.1 of IMC 0609, Appendix F, Attachment 1, the inspectors determined the finding was of very low safety significance (Green), because with the sprinklers nonfunctional, the reactor was able to reach and maintain safe shutdown. The finding has a cross -cutting aspect in the area of human performance associated with the field presence attribute because leaders did not observe, coach, and reinforce standards and expectations regarding scaffolding . Deviations from standards and expectations for building scaffolding near fire protection sprinklers were not corrected promptly. (H.2) |
| 05000324/FIN-2017001-04 | 2017Q1 | Green | P.3 | NRC identified | Failure to Enter the Technical Specification for an Inoperable 1D Control Room Air Conditioning Unit | Green . An NRC- identified Green NCV of Technical Specification (TS) 3.7.4, Control Room Air Conditioning (AC) System, was identified for the failure to declare the 1D control room AC unit inoperable. Specifically, on December 1, 2016, the licensee failed to declare the 1D control room AC unit inoperable due to extensi ve corrosion on the support channels . As a result, the 1D control room AC unit was inoperable from December 1, 2016, until the next time it was inspected on January 30, 2017, and exceeded the TS allowed outage time. As corrective actions, the licensee replaced the supports of the 1D and 2D control room AC units and inspected the 2E control room AC unit for corrosion. The licensee entered this issue into the CAP as NCRs 2113799 and 2113800. The inspectors determined the licensees failure to declare the 1D control room AC unit inoperable on December 1, 2016, and enter TS 3.7.4 was a performance deficiency. The finding was more than minor because it was associated with the structures, systems, and components ( SSC ) att ribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, this resulted in the 1D control room AC unit being inoperable from December 1, 2016, to January 30, 2017. Using IMC 0609, Appendix A, issued June 19, 2012, the SDP for Findings At -Power, the inspectors det ermined the finding was of very low safety significance (Green) because the finding did not only represent a degradation of the radiological barrier function for the control room and the finding did not represent a degradation of the barrier function of the control room against smoke or toxic atmosphere. This finding had a cross cutting aspect in the area of problem identification and resolution associated with the resolution aspect because the licensee failed to take effective corrective actions to addres s issues in a timely manner commensurate with their safety significance. Specifically, the licensee did not correct the degradation of the 1D control room AC unit until the unit was inoperable. (P.3) |
| 05000324/FIN-2017001-03 | 2017Q1 | Green | H.6 | NRC identified | Failure to Install Flood Barrier Seals Around the EDG 2 Four -Day Fuel Oil Tank Vents | Green . An NRC- identified Green NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified for the failure of the licensee to install flood barrier seals around the emergency diesel generator (EDG) 2, four -day fuel oil tank v ent as described in engineering change (EC) 400606. This result ed in a nonfunctional flood barrier into the EDG 2 four -day tank room. As an immediate corrective action, the licensee grouted the opening to prevent water intrusion into the EDG 2 four -day f uel oil tank room. The licensee entered this issue into the CAP as NCR 2093563. The inspectors determined the failure of the licensee to control the design of the installation of the new EDG 2 four -day fuel oil tank vent was a performance deficiency. Th e finding is more than minor because it is associated with the protection against external factors (i.e., flood hazard) 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 (i.e., c ore damage). Specifically, the licensee failed to install flood barrier seals around the EDG 2 four -day fuel oil tank vent designed to mitigate a flood of the EDG 2 four -day fuel oil tank room. Using IMC 0609, Appendix A, issued June 9, 2012, The SDP for Findings At -Power, the inspectors determined the finding screened to Exhibit 4, External Events Screening Questions, since the finding involved the loss of equipment specifically designed to mitigate a flood. The inspectors determined the finding screened to Green since if the flood barrier is assumed to be completely failed, it 4 would not result in the inoperability or degradation of EDG 2, and would not involve the total loss of any safety function that contributes to external event initiated core damage accident sequences. The finding has a cross -cutting aspect in the area of human performance associated with the design margins attribute because the licensee failed to maintain equipment within design margins and failed to change margins through a systematic and rigorous process. Specifically, the licensee changed the installation of the EDG 2 fuel oil tank roof vent without ensuring flood protection during the modification. (H.6) |
| 05000324/FIN-2016004-01 | 2016Q4 | Severity level IV | NRC identified | Failure to Submit a Licensee Event Report for a Condition Prohibited by Technical Specification | SL IV. An NRC-identified severity level (SL) IV NCV of 10 CFR 50.73(a)(2)(i)(B) was identified for the failure of the licensee to provide a written report to the NRC within 60 days of identifying a condition which was prohibited by plant Technical Specifications (TSs). The licensee entered this issue into the corrective action program (CAP) as nuclear condition report (NCR) 2091608. The inspectors had previously evaluated the underlying technical issue and determined the failure to promptly identify and correct a condition adverse to quality, which resulted in the condition prohibited by Technical Specifications (TS), was a performance deficiency. The issue was documented as a Green NCV, 05000325;324/2016002-01, Failure to Identify Broken Auto Start Control Relay on Emergency Diesel Generator 1. The ROPs significance determination process does not specifically consider the regulatory process impact in its assessment of licensee performance. Therefore, it was necessary to address this violation which involved a failure to make a required report to the NRC and was considered to impact the regulatory process, using traditional enforcement to adequately deter non-compliance. Using the NRC Enforcement Policy, Section 6.9.d.9, the SL assigned to this violation was SL IV, because the licensee failed to make a report required by 10 CFR 50.73. This violation also meets the criteria for an NCV because it was not repetitive or willful, and was entered into the licensees CAP. Traditional enforcement violations are not assessed for cross-cutting aspects. | |
| 05000324/FIN-2016004-02 | 2016Q4 | Severity level IV | Licensee-identified | Licensee-Identified Violation | The following violation 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 SL IV NCV. 10 CFR 55.33(e), states in part, To maintain an active status, a license holder shall actively perform the functions of an operator or senior operator on a minimum of seven 8-hour or five 12-hour shifts per calendar quarter. (f) If paragraph (e) is not met, before resumption of function...an authorized representative of the facility licensee shall certify...(2) That the licensee has completed a minimum of 40 hours of shift function under the direction of an operator or senior operator as appropriate and that the position to which the individual will be assigned. The 40 hours must have included a complete tour of the plant and all required shift turnover procedures. 0 OI-01-01.05, License Activation and Maintenance, is a licensee written procedure to ensure 10 CFR 55.53(f) requirements are satisfied. Section 5.3.2(d) states, An inactive licensee must stand a minimum of four 12 hour shifts under the direction of an active license holder and in the position to which the individual will be assigned. The time on shift must have included a complete tour of the plant and all required shift turnover procedures under the direction of a Shift Manager or Control Room Supervisor with an active license. Contrary to the above, on October 16, 2016, the licensee identified that they failed to properly reactivate a licensed senior reactor operator (SRO) on June 10, 2016 and subsequently allowed the operator to stand watch in active licensed positions for a four month period. Because this issue had no adverse impacts to nuclear safety, no incidents or incorrect assessments made during this time frame by the individual in question, and the event has been entered into the licensees CAP as NCR 2070317, this violation is being treated as a SL IV NCV, consistent with the NRC Enforcement Policy. | |
| 05000324/FIN-2016003-05 | 2016Q3 | Severity level IV | Licensee-identified | Licensee-Identified Violation | 10 CFR 50.9, Completeness and Accuracy of Information, requires, in part, that information required by statute or by the Commission's regulations, orders, or license conditions to be maintained by the licensee shall be complete and accurate in all material respects. WOs associated with safety-related activities are designated as quality-related records that are required to be maintained by the licensees Quality Assurance Program. Contrary to the above, on March 23, 2015, a contract QC inspector deliberately falsified the quality record for WO 2281641-05, resulting in the licensees maintaining of information that was incomplete and inaccurate. The WO is material to the NRC, as it provides documented evidence of compliance with QC inspection requirements. Using traditional enforcement, the SL assigned to this violation was SL IV, because there was a deliberate violation of regulatory requirements, and resulted in no or relatively inappreciable potential safety consequences. This violation also meets the criteria for a NCV because the licensee identified the violation and promptly provided the information concerning the violation, to the NRC. This issue was documented in the licensees CAP as NCR 739864. | |
| 05000324/FIN-2016003-01 | 2016Q3 | Green | H.8 | NRC identified | Failure to Implement Risk Management Actions during Elevated Risk | An NRC-identified Green non-cited violation (NCV) of 10 CFR 50.65(a)(4) was identified for the failure of the licensee to implement all necessary prescribed risk management actions (RMAs) during a 2A residual heat removal (RHR) and residual heat removal service water (RHRSW) outage. Specifically, between August 31, 2016, and September 1, 2016, the licensee failed to post protective equipment signs on the 2B RHR/RHRSW motor control centers (MCCs) whose unavailability would have taken Unit 2 into a Yellow risk condition. The licensee took immediate corrective actions to protect the 2B RHR/RHRSW MCCs in the field. The licensee entered this issue into the corrective action program (CAP) as nuclear condition report (NCR) 2059064. The inspectors determined the failure of the licensee to adequately post protected equipment signs for the 2B RHR/RHRSW system, whose unavailability would have taken the unit into a Yellow risk condition, was a performance deficiency. The finding was more than minor because if left uncorrected, the failure to perform RMAs could result in a loss of a safety-related mitigating function, specifically the RHR low pressure coolant injection (LPCI). Using IMC 0609, Appendix K, issued May19, 2005, Maintenance Risk Assessment and Risk Management Significance Determination Process, Flowchart 2, Assessment of RMAs, the inspectors determined the finding screened as very low safety significance (Green) since the incremental core damage probability was less than 1E-6. The finding has a crosscutting aspect in the area of human performance associated with the procedure adherence attribute because the licensee failed to follow plant procedures to fully protect the 2B RHR/RHRSW loop during the 2A RHR/RHRSW loop outage. |
| 05000324/FIN-2016003-04 | 2016Q3 | Green | Licensee-identified | Licensee-Identified Violation | 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires in part that, activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Instructions, procedures, or drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished. WO 2281641-05 instructions required in part that, the craft and quality control (QC) hold points, verify the unistrut bolts were torqued to a minimum torque of 72 in/lbs., verify tagging and marking, verify clearances and tolerances, and verify configuration (including dimensions). Contrary to the above, on March 23, 2015, a contract QC inspector deliberately failed to accomplish an activity affecting quality in accordance with a prescribed, documented instruction. Specifically, the QC inspector signed the QC hold points on WO 2281641-05 associated with the removal/installation of a safety-related pipe support for backup nitrogen in the drywell, without verifying and confirming the work had been satisfactorily accomplished according to the WO. Using IMC 0609, Appendix G, Shutdown Operations Significance Determination Process, inspectors determined this finding to be of very low safety significance (Green) since the finding did not involve the total loss of any safety function. This issue was documented in the licensees CAP as NCR 739864. | |
| 05000324/FIN-2016003-03 | 2016Q3 | Green | H.11 | Self-revealing | Inadequate Procedure for the 2B RHRSW Subsystem Operability Test | A self-revealing Green NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified for the failure of the licensee to have an adequate procedure for the 2B RHRSW pump operability test. Specifically, from July 12, 2001, to August 2, 2016, licensee procedure 0PT-08.1.4A(B), RHR Service Water System Operability Test, did not contain sufficient information to maintain plant status control for the Unit 2 RHRSW subsystem B pressure switch instrument isolation valves, 2-SW-PS-1176B-3 and 2-SW-PS-1176D-3. This resulted in the valves being found mispositioned (closed) and the inoperability of the 2B RHRSW subsystem from June 4 - 15, 2016. This finding resulted in a violation of TS 3.7.1, RHRSW System, since the 2B RHRSW subsystem was inoperable for greater than the TS allowed outage time (AOT). As immediate corrective actions, the licensee opened the 2-SW-PS-1176B(D)-3 valves and ensured the subsystem A pressure switch instrument isolation valves were open. Additionally, the licensee revised procedure 0PT-08.1.4A(B) to maintain plant status control by throttling the drain valves versus the pressure switch instrument isolation valves, and included an independent verification step to ensure the valves are returned to the correct position. The licensee entered this issue into the CAP as NCR 2037920. The inspectors determined the licensees failure to have an adequate procedure for the 2B RHRSW subsystem operability test to ensure configuration control was a performance deficiency. The finding was more than minor because it was associated with the procedural quality attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the inadequate procedure resulted in the inoperability of the 2B RHRSW subsystem. Using IMC 0609, Appendix A, issued June 19, 2012, the SDP for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined the finding screened to a more detailed risk evaluation, since the finding represented an actual loss of safety function of a single train for greater than its TS AOT. A regional Senior Risk Analyst (SRA) performed a detailed risk evaluation for the finding by setting the exposure period for 11 days, and assuming recovery actions that could be taken to mitigate the event. In addition, another later recovery was possible for the dominant sequences because service water, in sufficient quantity, could be pushed through the inoperable pumps to provide adequate cooling in non-loss-of-coolant accident (LOCA) sequences. The dominant contributor involved loss of the heat sink through common cause failure. The risk analysis resulted in a finding that is characterized as very low safety significance (Green). The finding had a cross-cutting aspect in the area of human performance associated with the challenge the unknown attribute because the licensee did not stop when faced with uncertain conditions, and risks were not evaluated and managed before proceeding. Specifically, the licensee continued through the April 2016 2B RHRSW system operability test, even when the procedure was not clear on which valve to manipulate to adjust for flow fluctuations. |
| 05000324/FIN-2016003-06 | 2016Q3 | Severity level Enforcement Discretion | NRC identified | Tornado Missile Vulnerability Results in Condition Prohibited by Technical Specifications | On July 21, 2016, Units 1 and 2 were in Mode 1 at 100 percent of RTP. At that time, the licensee determined that a conduit in the EDG building was vulnerable to a tornado missile. The tornado missile vulnerability has existed since original plant construction. The conduit contains cables associated with Unit 2 NSW pump B. If the cables were disabled by a tornado missile, 2B NSW pump would be inoperable. TS 3.7.2, Service Water System and Ultimate Heat Sink, requires three of the four NSW pumps to be operable. The licensee determined that during other NSW pump outages, TS 3.7.2 AOT was exceeded. Inspectors verified the immediate compensatory measures were taken and intermediate compensatory measures were taken including updating station abnormal procedures to start the NSW pumps via the alternate safe shutdown switches and providing training to address the vulnerability during tornadoes. Enforcement Guidance Memorandum 15-002, Enforcement Discretion for Tornado-Generated Missile Protection Noncompliance, was implemented, and the licensee declared the 2B NSW pump operable but nonconforming. The licensee entered this issue into the CAP as NCR 2028383. The inspectors reviewed the cause evaluation and the LER. Documents reviewed are listed in the Attachment. | |
| 05000324/FIN-2016003-02 | 2016Q3 | Green | H.5 | Self-revealing | Inadequate Procedure to Perform Preventive Maintenance on the HPCI Auxiliary Oil Pump Motor Overload Relay | A self-revealing Green NCV of Technical Specification (TS) 5.4.1a, Procedures, was identified for the failure of the licensee to have an adequate procedure for preventive maintenance (PM) on the Unit 2 high pressure coolant injection (HPCI) auxiliary oil pump motor overload relay 2-2XDA-B11-74. Specifically, from May 26, 2015, to July 6, 2016, the licensee failed to incorporate PM task 482688, a 12-year replacement task for the relays, into procedures, resulting in a shorted relay coil, the loss of control power, and the inoperability of the HPCI pump. The licensee replaced the relay and the HPCI pump was returned to operable. The licensee entered this issue into the CAP as NCR 2043067. The inspectors determined that the failure of the licensee to have an adequate PM procedure to replace the Unit 2 HPCI auxiliary oil pump motor overload relay 2-2XDA-B11-74 was a performance deficiency. The finding was more than minor because it was associated with the procedural quality attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to replace the HPCI auxiliary oil pump motor overload relay resulted in the inoperability of the Unit 2 HPCI pump, and the loss of safety function. Using IMC 0609, Appendix A, issued June 19, 2012, the SDP for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined the finding screened to a more detailed risk evaluation, since the finding represented a loss of HPCI system and/or function. The inspectors used SAPHIRE to conduct a more detailed risk review of the finding. The inspectors determined that the finding was of very low safety significance (Green), because the core damage frequency (CDF) risk was less than 1.0E-6/year. This finding has a cross-cutting aspect in the area of human performance associated with the work management aspect, for failing to implement a process of planning, controlling, and executing work activities such that nuclear safety is an overriding priority. Specifically the licensee failed to effectively plan and coordinate PM strategies associated with operating experience to prevent the failure of the HPCI pump. |
| 05000325/FIN-2016008-01 | 2016Q2 | H.12 | NRC identified | Inadequate Procedures to Perform Maintenance on the SAT non-segregated bus and the 1B RRP VFD cables | A self-revealing finding with two examples was identified for the licensees failure to have adequate procedures to perform maintenance on the startup auxiliary transformer (SAT) non-segregated bus duct and the 1B Reactor Recirculation Pump (RRP) variable frequency drive (VFD) cables. The first example, from May 1997 to the present, procedure 0PM-NSB001, Inspection and Cleaning Non-Segregated Buses, did not contain sufficient information to ensure that deficiencies that could lead to water intrusion in the SAT non-segregated bus duct were identified and corrected. The second example, from October 2003 to June 20, 2016, procedure 0SPP-CBL011, Splicing of Wires and Cables Without Tape, failed to specify use of a depth-limiting cutting tool for removing semi-conductor insulation on the 1B RRP VFD cables. The licensee entered this issue into the corrective action program (CAP) as nuclear condition report (NCR) 1998726. The inspectors determined that the failure of the licensee to have adequate procedures to perform maintenance on the SAT non-segregated bus duct and the 1B RRP VFD cables was a performance deficiency. The finding was more than minor because it was associated with the equipment performance attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the failure to perform adequate maintenance on the SAT non-segregated bus duct and the 1B RRP VFD cables resulted in a SAT differential lockout, a Unit 1 manual reactor SCRAM, and a loss of offsite power (LOOP). Using IMC 0609, Appendix A, issued June 19, 2012, the Significance Determination Process for Findings At-Power, the inspectors determined the finding screened to a more detailed risk evaluation because the finding caused 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 finding could not be screened to Green and is pending an initial significance characterization and is not yet finalized. The finding does not currently present an immediate safety concern because the licensee repaired the A phase fault on the non-segregated bus, resealed the bus duct bank, spliced in new cables to the 1B RRP VFD breaker and replaced the 1B RRP VFD breaker. The finding has a cross-cutting aspect in the area of human performance associated with the avoid complacency attribute because individuals failed to recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes, and individuals failed to implement appropriate error reduction tools. Specifically, the licensee failed to plan for the inherent risk associated with water intrusion into the SAT non-segregated bus duct and failed to implement error reduction tools when inspecting and repairing the duct. (H.12) | |
| 05000324/FIN-2016002-01 | 2016Q2 | Green | P.1 | NRC identified | Failure to Identity Broken Auto Start Control Relay on Emergency Diesel Generator 1 | An NRC-identified Green non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified because the licensee failed to promptly identify and correct a condition adverse to quality (CAQ) on emergency diesel generator (EDG) 1. Specifically, from February 7, 2016, until March 5, 2016, the licensee failed to promptly identify and correct a broken auto start control relay (ASCR) which resulted in reduced capacity of EDG 1 due to load oscillations and inoperability of EDG 1 due to oscillating between droop and isochronous mode. The oscillations could cause the EDG to not meet Technical Specification (TS) frequency and load requirements. The licensee replaced the ASCR and entered this issue into the corrective action program (CAP) as nuclear condition report (NCR) 2007720. The licensees failure to promptly identify and correct the broken ASCR, which resulted in reduced capacity and inoperability of EDG 1 due to load oscillations, was a performance deficiency. The finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to identify and correct the malfunctioning ASCR resulted in reduced capacity of EDG 1 due to load oscillations, and could cause EDG 1 to not meet TS frequency and load requirements. Using IMC 0609, Appendix A, issued June 19, 2012, The Significance Determination Process (SDP) for Findings At-Power, the inspectors determined the finding screened to a more detailed risk evaluation because it represented a loss of system and/or function, and the finding represented an actual loss of a function of a single train for greater than the TS allowed outage time. The regional Senior Reactor Analyst evaluated the finding and determined it to be Green. The risk was low because of the diverse sources of AC power available, and the long duration of some of the sequences allowed a greater potential for recovery of a failed AC power source. The dominant risk sequences contained common cause failure of the diesel generators, with the supplemental EDG aligned to the other unit, and non-recovery of offsite power or of an EDG. The finding has a cross-cutting aspect in the area of problem identification and resolution associated with the identification attribute because the licensee failed to implement a CAP with a low threshold for identifying issues completely, accurately, and in a timely manner in accordance with the program. Specifically, the licensee failed to write a timely NCR and identify the load oscillations as a CAQ. (P.1) |
| 05000324/FIN-2016002-02 | 2016Q2 | Green | P.1 | Self-revealing | Failure to Verify or Check the Adequacy of Design of the EDG 3 Auto-Start Circuitry | A self-revealing Green NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified for the licensees failure to verify or check the adequacy of design of the EDG 3 emergency auto-start circuitry. Specifically, on October 24, 2011, the licensee failed to verify or check the adequacy of design of the fuse block holder modification to the EDG auto-start circuitry. This resulted in the fuse block holder connection becoming loose, a loss of continuity through the circuit, and the inoperability of EDG 3. The licensee replaced the fuse block holder, performed a continuity check, and plans to implement a design change to install continuity indication for continuous verification of continuity. The licensee entered this issue into the CAP as NCR 2007449. The licensees failure to verify or check the adequacy of design of the EDG 3 emergency auto-start circuitry fuse block holder modification was a performance deficiency. The performance deficiency was more than minor because it was associated with the design 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. This resulted in the fuse block holder connection becoming loose, a loss of continuity through the circuit, and the inoperability of EDG 3. Using IMC 0609, Appendix A, issued June 19, 2012, The Significance Determination Process (SDP) for Findings At-Power, the inspectors determined the finding screened to a more detailed risk evaluation because it represented a loss of system and/or function, and the finding represented an actual loss of a function of a single train for greater than the TS allowed outage time. The regional SRA performed a detailed risk review for the finding. The finding was determined to be Green. The limited duration of the EDGs failure of the auto start, the ability to manually recover the EDG, and the availability of the other EDGs and of the supplemental EDG contributed to the low risk value. The dominant risk sequences were of low value, and were Station Blackout with failure to recover offsite power or the EDGs. The finding has a cross-cutting aspect in the area of problem identification and resolution associated with the identification attribute because the licensee failed to implement a CAP with a low threshold for identifying issues completely, accurately, and in a timely manner in accordance with the program. Specifically, the licensee failed to identify EDG 3 was inoperable on February 7, 2016, when the indications were apparent. (P.1) |
| 05000325/FIN-2016001-01 | 2016Q1 | NRC identified | ASME Section IX Weld Procedure Qualification | The inspectors identified an unresolved item (URI) associated with the qualification of the weld procedure specification (WPS) used for replacement of a portion of nuclear service water piping. Description: While conducting buried piping inspections in support of license renewal, the licensee identified pitting on the exterior wall of a portion of the Unit 1 nuclear service water supply header (1-SW-103-30-157). The licensee chose to address this by replacing the section of pipe (WO 12274010-08). The licensees repair/replacement plan for this activity identified that the requirements of ASME Section III, 1986 Edition, Subsection ND were applicable for the repair. By reference (ND-4320), several ASME Section IX Subsection QW requirements also applied. First, QW-200.2(f) allowed the use of multiple Procedure Qualification Records (PQRs) to produce a single WPS, provided that each essential variable is addressed by at least one PQR. Second, QW-403.8 and QW-404.30 established the requirements for two essential variables (base metal thickness qualified and filler metal thickness qualified, respectively) and referred to QW-451, which established the limits for both. The inspectors are opening a URI to review whether the licensees use of PQRs 1, 5, 193A and 193B to qualify the base metal and filler metal thickness ranges identified in WPS 01-1-04 and WPS 01-3-04 in accordance with Code was appropriate, and if a performance deficiency exists. The licensee wrote NCR 2009571 to address this issue. The issue is being tracked as a URI: URI 05000325/2016001-01, ASME Section IX Weld Procedure Qualification. | ||
| 05000324/FIN-2016001-04 | 2016Q1 | NRC identified | Notice of Enforcement Discretion for Replacement of EDG 3 Broken Fuse Block Connection | In accordance with the NRCs NOED process, the inspectors are opening a URI to facilitate prompt tracking, documentation, and closure of inspection, verification, and resolution activities, including enforcement action determinations, associated with the NOED. The inspectors are opening a URI to review the cause determination and determine if a performance deficiency exists. Description: On March 4, 2016, due to the inoperability of emergency bus E1, BOP bus 1D, EDG 1 and EDG 3, the licensee requested the NRC not enforce compliance with TS LCO 3.0.3, which requires, in part, that actions shall be initiated within 1 hour to place the unit, as applicable, in: Mode 2 within 7 hours; Mode 3 within 13 hours; and Mode 4 within 37 hours until March 5, 2016 at 12:35 p.m., for an additional 17 hours for each mode change. The licensee requested and was granted the NOED on March 4, 2016 at 3:35 pm. The LCO extension allowed the site time to complete the replacement of the EDG 3 broken fuse block connection, satisfactorily tested EDG 3, and declare the EDG operable. The licensee entered this issue in the CAP as NCR 2007449. This issue is being tracked as a URI: URI 5000324/2016001-04; Notice of Enforcement Discretion for Replacement of EDG 3 Broken Fuse Block Connection. | ||
| 05000324/FIN-2016001-02 | 2016Q1 | Green | P.2 | NRC identified | Failure to Identify and Correct a Condition Adverse to Quality Associated with the 2B NSW Pump Strainer | The inspectors identified a Green non-cited (NCV) of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, because the licensee failed to identify and correct a condition adverse to quality associated with the 2B nuclear service water (NSW) pump strainer. Specifically, the licensee did not ensure the spacers/shims were filed down or seated appropriately, which resulted in the 2B NSW pump strainer shear key failures, and the unavailability of the 2B NSW pump on three separate occasions. As corrective actions, the licensee ensured the spacers/shims were filed down and seated appropriately for the 2B NSW pump strainer and changed the procedure to ensure these steps were performed in the future. The licensee entered this issue into the corrective action program (CAP) as nuclear condition report (NCR) 1988423. The inspectors determined the licensees failure to ensure the 2B NSW pump strainer spacers/shims were filed down or seated appropriately was a performance deficiency. The finding was 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 of systems that respond to initiating events to prevent undesirable consequences. Specifically, this resulted in the failure of 2B NSW pump strainer shear key, and unavailability of the 2B NSW pump during repairs to the strainer. Using IMC 0609, Appendix A, issued June 19, 2012, the SDP for Findings At-Power, the inspectors determined the finding was of very low safety significance (Green) because the finding did not affect the design or qualification of a mitigating SSC, the finding did not represent a loss of system and/or function, the finding did not represent an actual loss of a function of a single train for greater than the technical specification (TS) allowed outage time, the finding did not represent an actual loss of a function of one or more non-TS trains of equipment, and did not screen as potentially risk-significant due to a seismic, flooding, or severe weather initiating event. The finding has a cross-cutting aspect in the area of problem identification and resolution associated with the evaluation attribute because the licensee failed to thoroughly evaluate issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, the licensee failed to evaluate the cause of the 2B NSW pump strainer shear pin failures to ensure the appropriate repair. (P.2) |
| 05000325/FIN-2016001-03 | 2016Q1 | NRC identified | Alert Declared Due to Electrical Fault Resulting in a Fire/Explosion | The inspectors are opening a URI to facilitate prompt tracking, documentation, and closure of inspection, verification, and resolution activities, including enforcement action determinations, associated with the Alert declaration due to the electrical fault resulting in an explosion/fire in the Unit 1 BOP 4 kV switchgear bus area. The inspectors are opening a URI to review the root cause and determine if a performance deficiency exists. On February 7, 2015, operations personnel declared an Alert for Units 1 and 2, in accordance with Emergency Action Level HA 2.1 due to an explosion/fire in the Unit 1 BOP 4 kV switchgear bus area. A manual reactor SCRAM was initiated due to loss of both recirculation system variable speed drives as a result of an electrical fault. The SAT experienced a lockout fault; interrupting offsite power to emergency buses 1 and 2. EDGs 1, 2, 3, and 4 automatically started and EDGs 1 and 2 synchronized to emergency buses 1 and 2 per design. The licensee restored offsite power to the emergency buses and exited the NOUE. The licensee wrote NCR 1998726 to address this event. This issue is being tracked as a URI: URI 05000325/2016001-03; Alert Declared Due to Electrical Fault Resulting in a Fire/Explosion. | ||
| 05000324/FIN-2015004-01 | 2015Q4 | Green | H.11 | Self-revealing | Inadequate Procedure for the 2C RHRSW Booster Pump Motor Bearings | A self-revealing Green non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified for the failure of the licensee to have an adequate procedure for the 2C residual heat removal service water (RHRSW) pump motor bearing maintenance. Specifically, licensee procedure 0CM-M503, Maintenance Instructions for the RHRSW Booster Pump Motors, did not contain information to ensure proper sealing of the 2C RHRSW motor bearings. This finding resulted in a violation of technical specification (TS) 3.0.4, Limiting Condition for Operation (LCO) Applicability, and TS 3.7.1, RHRSW System. As immediate corrective actions, the licensee applied sealant to the motor bearings. Additionally, the licensee revised procedure 0CM-M503 and added a detailed location for applying the sealant to the RHRSW pump motors. The licensee entered this issue into the Corrective Action Program (CAP) as nuclear condition report (NCR) 742643. The inspectors determined the licensees failure to have an adequate procedure for the 2C RHRSW pump motor bearing maintenance was a performance deficiency. The finding was more than minor because it was associated with the procedural quality attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the inadequate procedure resulted in the inoperability of the Loop A RHRSW subsystem, and the loss of safety function while the Loop B RHRSW subsystem was out for maintenance. Using IMC 0609, Appendix A, issued June 19, 2012, the SDP for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined the finding screened to a more detailed risk evaluation, since the finding represented a loss of system and/or function. The regional Senior Reactor Analyst performed a detail risk review of the finding. The at-power model was conservatively used to bound the risk that would happen at the proposed time of failure, which was many days after shutdown due to the time it takes for the oil leak to cause potential bearing failure. Since the licensee had procedures for running the service water (SW) system without the RHRSW pumps energized, and the decay heat loads at the time of failure would be low, a failure rate of only 0.1 for the loss of function was assumed. This was also conservative, since the adverse conditions that would have prevented refill of the oil were LOCA assumptions, and LOCA sequences did not contribute greatly to the risk in the model. The at-power models solution was more than an order of magnitude below the Green/White threshold for the SDP. Therefore, the finding was determined to be of very low safety significance (Green). The finding has a cross-cutting aspect in the area of human performance associated with the challenge the unknown attribute because the licensee did not stop when faced with uncertain conditions, and risks were not evaluated and managed before proceeding. Specifically, the licensee continued through the 2010 and 2013 2C RHRSW pump maintenance outages, even when the bearings were found without sealant. Additionally, the licensee did not question the procedurally required location for the sealant. |
| 05000324/FIN-2015003-01 | 2015Q3 | Green | Licensee-identified | Licensee-Identified Violation | Technical Specification 5.4.1.a requires written procedures to be established as recommended by Safety Guide 1.33, November 3, 1972. Section I.1 recommends procedures for performing maintenance of safety related equipment should be properly pre-planned and performed in accordance with documented instructions appropriate to the circumstances. Contrary to the above, on June 3, 2015, the licensee failed to establish appropriate work instructions to properly pre-plan and perform maintenance that affected the performance of Unit 1 secondary containment. Specifically, WOs 13304512 and 13304513 failed to ensure that secondary containment was declared inoperable prior to work on the RHRSW motor coolers. The performance deficiency is more than minor, because it adversely affected the procedure quality attribute of the barrier integrity cornerstone to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, and Inspection Manual Chapter 0609, Appendix A, Exhibit 3, Barrier Integrity, the inspectors determined that the finding was of very low safety significance (Green) because it only represented a degradation of the radiological barrier function provided for the SBGT system. The licensee entered this issue into their CAP as NCR 752720 and took actions to make a late LCO entry and change the WO instruction. | |
| 05000324/FIN-2015003-02 | 2015Q3 | Severity level Enforcement Discretion | NRC identified | Implementation of EGM 11-003, Revision 2, Enforcement Guidance Memorandum on Dispositioning Boiling Water Reactor Licensee Noncompliance with Technical Specification Containment Requirements During Operations with a Potential for Draining the Reactor Vessel | The inspectors reviewed the plants implementation of NRC EGM 11-003, Revision 2, during Unit 2 maintenance activities which had the potential to drain the reactor vessel during the Unit 2 refueling outage. The activities included: February 26, 2015: 398 gallons per minute leakage for establishing a clearance for RWCU vent and drain valve maintenance March 14, 2015: 85 gallons per minute leakage for RWCU maintenance activities and excess flow check valve testing through test valve 2-B32-V85 March 17, 2015, 0411: 1 gallon per minute leakage for RWCU isolation valve leak March 17, 2015, 1300: 1 gallon per minute leakage for RWCU isolation valve leak March 18, 2015: 1 gallon per minute leakage for RWCU isolation valve leak, leak stopped when the 2A RWCU pump was put under clearance These activities took place without secondary containment being operable. Inspectors verified compliance with the guidelines of EGM 11-003 prior to and during these activities. Additionally, inspectors verified that, for all dates, all other TSs were met during OPDRVs with secondary containment inoperable. TS 3.6.4.1, Secondary Containment requires that secondary containment be operable and is applicable during OPDRVs. The required action if secondary containment is inoperable in this condition is to initiate actions to suspend OPDRVs immediately. Contrary to the above, on February 26, 2015, March 14, 2015, twice on March 17, 2015, and March 18, 2015, the licensee failed to maintain secondary containment operable while performing OPDRVs. However, because the violations were identified during the discretion period described in EGM 11-003, Revision 2, the NRC is exercising enforcement discretion for the dates of February 26, 2015, March 14, 2015, twice on March 17, 2015, and March 18, 2015, in accordance with Section 3.5, Violations Involving Special Circumstances, of the NRC Enforcement Policy and, therefore, will not issue enforcement action for this violation, subject to a timely license amendment request being submitted. | |
| 05000324/FIN-2015007-01 | 2015Q2 | Green | P.2 | NRC identified | Failure to Identify Conditions Adverse to Quality | An NRC-identified Green non-cited violation (NCV) of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion XVI, Corrective Action, was identified for licensee failure to identify conditions adverse to quality during the evaluation of an emergency diesel generator (EDG) output breaker failure on March 16, 2015. Specifically, the licensee missed that an internal change made to a relay was a condition adverse to quality. Further, the licensee failed to reclassify a corrective action document to higher significance when information arose indicating that the event in question was a loss of safety function. The licensee documented these issues in their corrective action program, completed the necessary reviews for a condition adverse to quality, and reclassified the original event to Significance Level 1. The inspectors determined that the finding was more than minor in accordance with Manual Chapter 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, because, if left uncorrected, additional unqualified relays would likely have been installed in the plant. Using Manual Chapter 0609, Appendix A, Exhibit 1, effective July 1, 2012, the finding screened as Green for each unit by answering no to the questions related to an actual loss of function of a system, a single train, non-technical specification equipment designated as high safety-significant in accordance with the licensees maintenance rule program for >24 hrs. The finding had a cross-cutting aspect for Evaluation in the area of Problem Identification & Resolution because the most likely cause of the missed conditions adverse to quality was a lack of thorough investigation during the evaluations (for cause and reportability) of the relay issue. |
| 05000324/FIN-2015002-01 | 2015Q2 | Green | NRC identified | Degraded Fire Barrier Seals in the Unit 2 Cable Access Way | An NRC-identified Green non-cited violation (NCV) of License Condition 2.B.(6), Fire Protection Program, was identified for the licensees failure to maintain the 3-hour fire seals in the Unit 2 cable access way. Specifically, three cables in the Unit 2 cable access way were not within continuously enclosed conduits, which failed to preserve the integrity of the 3-hour rated barrier. As corrective action, the licensee sealed all three penetrations with a qualified 3-hour seal. This issue was entered into the licensees corrective action program (CAP) as nuclear condition report (NCR) 740606. The inspectors determined that the licensees failure to maintain the 3-hour penetration fire barrier conduits in the Unit 2 cable access way, as required by licensee specification 118-003, Selection and Installation of Fire Barrier and Pressure Boundary Penetration Seals, was a performance deficiency. The finding was more than minor because it was associated with the external factors attribute (i.e. fire) of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, this resulted in the failure of the three conduits to perform their function. The finding was screened using NRC IMC 0609, Appendix F, Fire Protection Significance Determination Process, dated September 20, 2013, because the finding affected the ability to confine a fire. Using IMC 0609, Appendix F, Attachment 1, Fire Protection SDP Phase 1 Worksheet, dated September 20, 2013, the finding was assigned to the Fire Confinement category because the degraded penetrations were located in a fire barrier that separated two fire areas. Proceeding to Task 1.3.1 of IMC 0609, Appendix F, Attachment 1, the inspectors determined the finding was of very low safety significance (Green) because safety significant equipment was located a sufficient distance from the degraded penetrations and the reactors ability to reach and maintain a safe shutdown condition was not impacted. The finding does not have a cross-cutting aspect since the performance deficiency is not indicative of current plant performance. (Section 1R05) | |
| 05000324/FIN-2015007-02 | 2015Q2 | Green | NRC identified | Insufficient Material Evaluation of Commercially Dedicated Allen Bradley Relays | An NRC-identified Green NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control was identified for the licensees inadequate commercial grade dedication technical evaluation that resulted in non-conforming relays being installed in the control circuits for emergency diesel generator output breakers. This led to specification of a relay that was unsuitable for the application being installed in the control circuit for two emergency diesel generator output breakers and failure of one of those breakers to close. The licensee documented this issue in their corrective action program and performed corrective actions to mitigate the effects of the undetected changes on the relay. The inspectors determined that the finding was more than minor in accordance with Manual Chapter 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, because, if the process for detecting commercial grade item changes using material evaluations was left uncorrected, additional undetected design or process changes would likely occur. Using Manual Chapter 0609, Appendix A, issued June 19, 2012, The Significance Determination Process (SDP) for Findings At-Power, the inspectors determined the finding required a detailed risk evaluation because the effect on two emergency diesel generators was considered a loss of function. For Unit 1, the regional Senior Reactor Analyst used demand data to adjust the probability that an emergency diesel generator would fail to start and ran a condition assessment on SAPHIRE. Because of limited exposure time, the finding was determined to be Green for Unit 1. For Unit 2, the conditions for exposure occurred during an outage with the reactor cavity filled, and both EDGs would be available. The SRA determined the significance to be bounded by the at power risk analysis performed for Unit 1. Because of the low exposure time, and the high likelihood of operators recovering the failure to start of the EDGs, this issue was Green for Unit 2. The inspectors did not identify a crosscutting aspect associated with this finding because the original relay evaluation was done in 1999 and was not indicative of current licensee performance. | |
| 05000325/FIN-2015002-02 | 2015Q2 | Green | H.7 | NRC identified | Inadequate Procedure for the 1B Conventional Service Water Pump Strainer Repair | An NRC-identified Green NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified for the licensees failure to have an adequate procedure to perform maintenance on the 1B conventional service water (CSW) pump strainer. Specifically, between August 28, 2009, and May 11, 2015, licensee procedure MNT-NGGC-0009, Application of Protective Coatings, was not adequate to perform repairs on the 1B CSW pump strainer, which resulted in through-wall leaks on three occasions. As corrective actions, the licensee repaired the weld, recoated the inside of the affected strainer area with Belzona coating using qualified individuals, and updated procedure MNT-NGGC-0009. The licensee entered this issue into the CAP as NCR 747712. The inspectors determined that the licensees failure to have an adequate procedure to perform maintenance on the 1B CSW pump strainer was a performance deficiency. The finding was more than minor because if left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, it could have led to a more significant failure of the 1B CSW pump strainer and the service water system. Using IMC 0609, Appendix A, issued June 19, 2012, the SDP for Findings At-Power, the inspectors determined the finding was of very low safety significance (Green) because the finding did not affect the design or qualification of a mitigating SSC, the finding did not represent a loss of system and/or function, the finding did not represent an actual loss of a function of a single train for greater than the TS allowed outage time, the finding did not represent an actual loss of a function of one or more non-TS trains of equipment, and did not screen as potentially risk-significant due to a seismic, flooding, or severe weather initiating event. The finding has a cross-cutting aspect in the area human performance associated with the documentation attribute because the licensee failed to create and maintain complete, accurate and up-to-date documentation to correct the 1B CSW pump strainer through-wall leak issue on three occasions. (H.7) |
| 05000324/FIN-2015002-03 | 2015Q2 | Green | P.2 | NRC identified | Failure to Perform an Adequate Extent of Condition Review for the 1C Conventional Service Water Pump Strainer | An NRC-identified Green finding of licensee procedure CAP-NGGC-0205, Condition Evaluation and Corrective Action Process, was identified for the licensees failure to perform an adequate extent of condition review for the 1C CSW pump strainer stop collar clearance issue. Specifically, between February 21, 2014, and April 8, 2015, the licensee failed to perform an adequate extent of condition to identify the 2C CSW pump strainer stop collar was also installed without being securely positioned. This resulted in the failure of the shear pin and inoperability of the 2C CSW strainer and pump. As corrective actions, the licensee replaced the shear pin securely and scheduled the replacement of the other CSW pump strainer shear pins at the earliest available work window. The licensee entered this issue into the CAP as NCR 742444. The inspectors determined that the licensees failure to perform an adequate extent of condition review for the 1C CSW pump strainer stop collar clearance issue, as required by licensee procedure CAP-NGGC-0205 was a performance deficiency. The finding was 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 of systems that respond to initiating events to prevent undesirable consequences. Specifically, this resulted in the failure of 2C CSW pump strainer shear pin, and inoperability of the 2C CSW strainer and pump. Using IMC 0609, Appendix A, issued June 19, 2012, the SDP for Findings At-Power, the inspectors determined the finding was of very low safety significance (Green) because the finding did not affect the design or qualification of a mitigating SSC, the finding did not represent a loss of system and/or function, the finding did not represent an actual loss of a function of a single train for greater than the TS allowed outage time, the finding did not represent an actual loss of a function of one or more non-TS trains of equipment, and did not screen as potentially risk-significant due to a seismic, flooding, or severe weather initiating event. The finding has a cross-cutting aspect in the area of problem identification and resolution associated with the evaluation attribute because the licensee failed to thoroughly evaluate issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, the licensee failed to evaluate the applicability of the stop collar clearance issue to the other strainers after the failure of the 1C CSW pump strainer shear pin. (P.2) |
| 05000324/FIN-2015002-04 | 2015Q2 | NRC identified | 2C Residual Heat Removal Service Water (RHRSW) Pump Oil Leak | The inspectors opened a URI to review the licensees evaluation of the motor oil leak on the 2C RHRSW pump and determine if there is a performance deficiency. On April 8, 2015, the licensee identified an oil leak on the motor for the 2C RHRSW pump in excess of the amount that would be acceptable for the pump to meet the 30-day mission time, and the pump was declared inoperable. The licensees immediate corrective actions were to apply sealant to the mechanical joints of the bearing housings. The licensee entered this issue in the CAP as NCR 742643. This issue is being tracked as a URI: URI 05000324/2015002-04, 2C Residual Heat Removal Service Water Pump Oil Leak. | ||
| 05000324/FIN-2015404-07 | 2015Q1 | Green | NRC identified | Security | ||
| 05000324/FIN-2015404-03 | 2015Q1 | Green | Licensee-identified | Licensee-Identified Violation | ||
| 05000324/FIN-2015404-04 | 2015Q1 | Green | Licensee-identified | Licensee-Identified Violation | ||
| 05000324/FIN-2015404-01 | 2015Q1 | Green | NRC identified | Security | ||
| 05000324/FIN-2015404-06 | 2015Q1 | Green | Licensee-identified | Licensee-Identified Violation | ||
| 05000324/FIN-2015404-05 | 2015Q1 | Green | Licensee-identified | Licensee-Identified Violation |