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05000313/FIN-2018003-052018Q3Arkansas NuclearFailure to Maintain Main Feedwater Pump B Discharge Pressure in Band Caused a Reactor TripThe inspectors reviewed a self-revealed, Green finding and associated non-cited violation of Arkansas Nuclear One, Unit 1, Technical Specifications 5.4.1.a, for the licensees failure to implement Procedure OP-1102.002, Plant Startup, Revision 106. Specifically, control room operators failed to maintain main feedwater pump discharge pressure in the required band to control flow to the steam generators during a plant startup. As a result, the only operating main feedwater pump tripped on high discharge pressure, causing an automatic reactor trip.
05000313/FIN-2018003-042018Q3Arkansas NuclearFailure to Verify Safety-Related 4160 V Breaker Operability Following Maintenance ActivitiesThe inspectors reviewed a self-revealed Green finding and associated non-cited violation of Arkansas Nuclear One, Unit 1, Technical Specification 5.4.1.a, for the licensees failure to properly preplan maintenance that can affect the performance of safety-related equipment. Specifically, the licensee failed to perform post-maintenance testing to demonstrate component operability for the train A safety-related 4160 V switchgear A-303 breaker that provides power to the swing service water pump B (P-4B) after the breaker was racked in. The breaker subsequently failed to close when attempting to start the pump.
05000313/FIN-2018003-032018Q3Arkansas NuclearFailure to Provide Complete and Accurate Information in a License Amendment Request to Change Emergency Action Level RequirementsThe inspectors identified a Severity Level IV non-cited violation because the licensee provided inaccurate information to the NRC in a license amendment request for an emergency action level scheme change. Specifically, the licensee provided information about the availability of the postaccident sampling system building radiation monitor and the Unit 1 level instrumentation that was material to the licensing decision, but not accurate. The NRC approved an emergency action level scheme change on November 9, 2012 (ADAMS Accession No. ML12269A455) to allow Arkansas Nuclear One to adopt the Nuclear Energy Institute (NEI) 99-01, Revision 5, scheme. Subsequently, the licensee identified that two of their current emergency action level thresholds could not be implemented in accordance with their emergency classification procedure: On May 26, 2017, Condition Report CR-ANO-2-2017-03161 documented that postaccident sampling system building radiation monitor 2RX-9840 should be removed from all regulatory commitments because the postaccident sampling system had been removed from service, and its building would not be monitored for radiological releases. Radiation monitor 2RX-9840 was being used as a means to evaluate emergency action levels AU1, AA1, AS1, and AG1. In addition, it was used in the loss/potential loss of containment (CNB6) for fission product emergency action levels. The condition report noted that requirements for the postaccident sampling system had been removed from Arkansas Nuclear One licenses in August 2000 and the licensee had abandoned the systems valves (March 2003, EC-ANO-1779), removed power from the postaccident sampling system ventilation system (January 2004), and made radiation monitor 2RX-9840 nonfunctional (May 2008, Condition Report CR-ANO-2-2008-01439 and Work Order 150817). On March 15, 2018, Condition Report CR-ANO-C-2018-01121 documented that the Unit 1 level instrumentation set point used in emergency action level CA1 was below the indicating range of the instrument. The emergency action level indicated that a loss of Unit 1s reactor vessel inventory was shown by an indicated level less than 368 feet, 0 inches. Therefore, the lowest level indicated on the instrument would be higher than the level used in making the emergency classification decision. The inspectors reviewed the licensees license amendment request, dated December 1, 2011 (ADAMS Accession No. ML113350317), Proposed Emergency Action Levels Using NEI 99-01, Revision 5, Scheme, and the licensees response to a request for additional information dated July 9, 2012, (ADAMS Accession No. ML12192A090) to determine whether the conditions identified in the corrective action program existed at the time the licensee requested the license amendment and whether the request correctly described the instruments. The inspectors identified: The December 1, 2011, submittal incorrectly indicated that radiation monitor 2RX-9840 was a viable means of classifying emergency action levels AU1, AA1, AS1, and AG1, as well as providing input for the evaluation of fission product barrier emergency action levels. In the response to NRCs request for additional information (RAI) dated July 9, 2012, the licensee provided additional details about the super particulate iodine noble gas (SPING) radiation monitors used in this application. Response to Question 3 associated with emergency action levels AA1, AS1, and AG1 stated: Each SPING is associated with a particular ventilation pathway and provides continuous monitoring of air discharged via the respective release pathway. The license reviewer concluded that all of the SPING monitors included in the license amendment request were operable and continuously monitoring the specified release pathways, thereby being capable of measuring the radiation levels described in the proposed emergency action levels. 17 The December 1, 2011, submittal indicated that loss of Unit 1 reactor vessel inventory for emergency action level CA1 was a vessel level less than 368 feet, 0 inches. This issue was NRC-identified because when the licensee identified the emergency action level errors, they took action to correct the errors, but failed to address the failure to ensure that technical information provided to the NRC in support of the license amendment request was complete and accurate in all material respects. Corrective Actions: To correct the Unit 1 reactor vessel level emergency action level threshold error, the licensee issued communications regarding correct application of the emergency action level on March 15, 2018, followed by implementation of a change to Procedure OP-1903.010, Emergency Action Level Classification, Revision 56, dated June 26, 2018, with the corrected level. The use of radiation monitor 2RX-9840 is being removed from the emergency action levels as part of an emergency action level scheme change submitted to the NRC on March 29, 2018 (ADAMS Accession No. ML18088B412 and ML18094A155). In the interim, the licensee issued communications to emergency director-qualified staff members to ensure they are aware of the error, how to address it if implementing emergency action levels, and to inform them of the corrective actions in progress. Additionally, the licensee issued Condition Report CR-ANO-C-2018-03597, dated September 13, 2018, for the incomplete and inaccurate emergency action level submission examples to address the completeness and accuracy issues identified by the inspectors.
05000313/FIN-2018003-022018Q3Arkansas NuclearFailure to Implement Welding Standard Guidance and Examination ProceduresThe inspectors reviewed a self-revealed Green finding and associated non-cited violation of Arkansas Nuclear One, Unit 1, Technical Specification 5.4.1.a, for the licensees failure to properly preplan maintenance that can affect the performance of safety-related equipment. Specifically, the licensee failed to implement welding standard guidance and examination procedure guidance during the installation of the high pressure injection system drain line containing drain valves MU-1066A and MU-1066B. The drain line weld developed a crack that caused a leak shortly after plant startup that was determined to have been caused by grinding during the welding process, which was not permitted by the welding standard.
05000458/FIN-2018003-012018Q3River BendLicensee-Identified ViolationThis violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a non-cited violation, consistent with Section 2.3.2.a of the NRC Enforcement Policy. Violation: Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that measures shall be established to assure that the design basis for those structures, systems, and components to which Appendix B applies is correctly translated into specifications, drawings, procedures, and instructions. The design basis for the control building air conditioning system, as specified in the updated safety analysis report, requires that the system be capable of performing its safety function in the event of a single failure in any component. Contrary to the above, the licensee failed to assure that the design basis was correctly translated into specifications for the control building air conditioning system. Specifically, while reviewing the control logic for the control building air conditioning system, the licensee discovered that the control logic was designed such that a single failure in a component in the control logic could have prevented the system from performing its specified safety function.
05000313/FIN-2018003-012018Q3Arkansas NuclearFailure to Translate the Design Requirements into Instructions for Refueling Emergency Diesel GeneratorsThe inspectors identified a Green finding and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to translate current design into instructions for Unit 1 and Unit 2 diesel fuel oil transfer system. Specifically, the licensee failed to translate the current diesel fuel oil transfer system design into instructions to refuel Unit 1 and Unit 2 safety-related fuel bunkers, T-57 and 2T-57, if the non-safety bulk diesel fuel oil tank T-25 was unavailable following a design basis event (e.g., tornado, external flooding, or earthquake) for which it was not designed to withstand.
05000313/FIN-2018003-062018Q3Arkansas NuclearReactor Power Transient Caused by the Turbine Bypass Valve Failing OpenThe inspectors reviewed a self-revealed Green finding and associated non-cited violation of Arkansas Nuclear One, Unit 1, Technical Specifications 5.4.1.a, for the licensees failure to properly preplan maintenance that can affect the performance of safety-related equipment. Specifically, the licensee failed to properly pre-plan maintenance for the replacement of air supply tubing for turbine bypass valve CV-6687, which resulted in the failure of the air tubing, causing valve CV-6687 to fail open, which led to a manual reactor trip and a subsequent loss of the main condenser.
05000482/FIN-2017008-022017Q4Wolf CreekFailure to Provide Adequate Emergency LightingThe team identified a non-cited violation of License Condition 2.C.(5) for failure to provide emergency lighting along alternate routes plant operators are allowed to take during implementation of the procedure for control room evacuation due to fire.The failure to provide 8-hour emergency lights along alternate routes used by operators during control room evacuation due to fire is a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external factors attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team evaluated this finding using Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, dated September 20, 2013, because it affected the ability to reach and maintain safe shutdown conditions in case of a fire. The team assigned the finding to the post-fire safe shutdown category since it impacted the alternate shutdown element. The issue screened to Green because the reactor would be able to achieve and maintain hot shutdown because the operators are required to carry flashlights. Specifically, the team had reasonable assurance that the operators would be able to complete the evacuation procedure using handheld flashlights to access safe shutdown equipment. The finding is assigned a cross-cutting aspect in the area of human performance, associated with training, because the operators are not being trained on the access and egress routes that are provided with 8-hour emergency lights during implementation of the control room evacuation procedure due to fire to ensure the time critical actions can be met (H.9).
05000482/FIN-2017008-012017Q4Wolf CreekInadequateEvaluation of Spurious Valve OperationThe team identified a non-cited violation of License Condition 2.C.(5) for failure to implement and maintain in effect all provisions of the approved fire protection program. Specifically, the licensee failed to adequately evaluate the potential impacts on post-fire safe shutdown of two motor operated valves spuriously closing due to fire damage.The failure to adequately evaluate the impact of pressure operated relief valve block valves spuriously closing on post-fire safe shutdown was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external events (fire) attribute of the Mitigating Systems cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Because the finding affected the ability to reach and maintain safe shutdown conditions in case of a fire that led to control room evacuation and because the Phase 2 methodology of Inspection Manual Chapter 0609, Appendix F, was not appropriate for this finding, a senior reactor analyst performed a Phase 3 evaluation to determine the risk significance. The analyst determined this finding was of very low risk significance (Green). There is no cross-cutting aspect associated with this finding since the performance deficiency is not reflective of present performance (i.e., the performance deficiency occurred more than 3 years ago).
05000298/FIN-2017002-032017Q2CooperLoss of Control Room Ventilation Due to Improper Switch ManipulationThe inspectors reviewed a self -revealed, non- cited violation of Technical Specification 5.4.1.a , for the licensees f ailure to implement System Operating Procedure 2.2.38, HVAC Control Building, Revision 43, during control building ventilation testing. Specifically, on December 7, 2016, when directed to turn off control building ventilation recirculation fan, RF- C-1A, operations personnel instead inadvertently turned off the operating control room emergency filtration system supply fan, 1 -SF -C-1A, resulting in the loss of the control room emergency filtration system function. Corrective actions to restore compliance included restoration of the control room emergency filtration supply fan and procedure changes to require peer checks for this surveillance test and similar 4 activities. The licensee entered this deficiency into the corrective action program as Condition Report CR -CNS -2016- 08744. The licensees failure to implement System Operating Procedure 2.2.38 , in violation of Technical Specification 5.4.1.a , was a performance deficiency. The performance deficiency was determined to be more than minor, and therefore a finding, because it was associated with the human performance attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers ( control room envelope) 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, dated June 19, 2012, the inspectors determined that the finding had very low safety significance (Green) because it did not represent a degradation of the barrier function of the control room against smoke or a toxic atmosphere. The finding had a cross -cutting aspect in the area of human performance associated with challenge the unknown, because the licensee did not stop when faced with uncertain conditions, and did not ensure that risks we re evaluated and managed before proceeding. Specifically, despite noting several a bnormalities with the switch being manipulated, operations personnel did not stop to evaluate the uncertain conditions nor did they evaluate the risks associated with proceeding (H.11).
05000440/FIN-2017002-012017Q2PerryFailure to Notify the NRC within Eight Hours of a Non -Emergency Event that Could Have Prevented the Fulfillment of Multiple Safety FunctionsSeverity Level IV. The inspectors identified a Severity Level IV Non- Cited Violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) 50.72(b)(3)(v)(A) and (D), Immediate Notification Requirements for Operating Nuclear Power Reactors, for the licensees failure to report an event to the NRC within eight hours that at the time of discovery could have prevented the fulfillment of a safety function. Specifically, the licensee did not recognize there was a loss of safety function associated with multiple instrumentation functions as a result of a main steam turbine bypass valve opening at 100 percent reactor power. Therefore, the licensee did not make the required non- emergency eight hour report. After the inspectors questioned the licensees conclusion, the licensee recognized there was indeed a loss of safety function and submitted the eight -hour notification report on May 3, 2017. They also and entered this issue into the corrective action program (CAP) as condition report ( CR) 2017 04939, CR 201704868, and CR 201705022. The failure to make an applicable non- emergency eight -hour event notification report within the required time frame was a performance deficiency. The inspectors determined that traditional enforcement was applicable to the issue because it impacted the NRCs regulatory process. In accordance with Section 2.2.2.d, and consistent with the examples included in Section 6.9.d.9 of the NRC Enforcement Policy, this violation was screened as a Severity Level IV violation that was more than minor. In accordance with Inspection Manual Chapter 0612, because this violation involved traditional enforcement and does not have an associated finding that would be considered more- than -minor, a cross-cutting aspect was not assigned to this violation.
05000416/FIN-2017002-022017Q2Grand GulfLicensee-Identified ViolationTitle 10 CFR 72.44(d)(3) requires, in part, that an annual report be submitted to the Commission, specifying the quantity of each of the principal radionuclides released to the environment in liquid and in gaseous effluents during the previous 12 months of operation and such other information as may be required by the Commission to estimate maximum potential radiation dose commitment to the public resulting from effluent releases. The report must be submitted within 60 days after the end of the 12- month monitoring period. Contrary to the above, from March 2, 2017 , until April 27, 2017, the licensee did not submit the annual report within 60 days after the end of the 12- month monitoring period. The NRCs significance determination process is not designed to assess the significance of violations that impact or impede the regulatory process. Therefore, the issue of an untimely annual report submittal was assessed using the traditional enforcement process in accordance with the Enforcement Policy. The inspectors determined the violation to be at Severity Level IV because the licensee submitted the annual report approximately 2 months late, and it is similar to examples in the Enforcement Policy , Section 6.9.d. Since this issue was entered into the licensees corrective action program as Condition Report CR-GGN-1-2017-03092 , compliance was restored within a reasonable period of time, the violation was not repetitive, and the violation was not willful, this violation is being treated as a n on-cited violation (NCV), consistent with Section 2.3.2.a of the Enforcement Policy. Traditional enforcement violations are not assigned a cross -cutting aspect .
05000416/FIN-2017002-032017Q2Grand GulfLicensee-Identified ViolationLicense Condition 2.C (46)(f) requires, during the first two scheduled refueling outages after reaching full EPU (extended power uprate) conditions, Entergy shall conduct a visual inspection of all accessible, susceptible locations of the steam dryer in accordance with BWRVIP -139 and GE inspection guidelines. Entergy shall report the results of the visual inspections of the steam dryer to the NRC staff within 60 days following startup. Contrary to the above, on August 16, 2012 , and May 15, 2014, the licensee did not report the results of the visual inspections of the steam dryer to the NRC staff within 60 days following startup. The NRCs significance determination process is not designed to assess the significance of violations that impact or impede the regulatory process. Therefore, the issue of an untimely inspection results submittal was assessed using the traditional enforcement process in accordance with the Enforcement Policy. The inspectors determined the violation to be at Severity Level IV because it is similar to examples in the Enforcement Policy Section 6.9.d. Since this issue was entered into the licensees corrective action program as Condition Report CR -GGN -1-2017 -03404, compliance was restored within a reasonable period of time, the violation was not repetitive, and the violation was not willful, this violation is being treated as a n on-cited violation (NCV), consistent with Section 2.3.2.a of the Enforcement Policy. Traditional enforcement violations are not assigned a cross -cutting aspect.
05000416/FIN-2017002-042017Q2Grand GulfLicensee-Identified ViolationLicense Condition 2.C (46)( g) requires, at the end of the second refueling outage following the implementation of the EPU, the licensee shall submit a long- term steam dryer inspection plan based on industry operating experience along with the baseline inspection results f or NRC review and approval. Contrary to the above, since May 15, 2014, the licensee did not submit a long -term steam dryer inspection plan based on industry operating experience along with the baseline inspection results for NRC review and approval. The NRCs significance determination process is not designed to assess the significance of violations that impact or impede the regulatory process. Therefore, the issue of an untimely inspection plan submittal was assessed using the traditional enforcement process in accordance with the Enforcement Policy. The inspectors determined the violation to be at Severity Level IV because it is similar to examples in the Enforcement Policy Section 6.9.d. Since this issue was entered into the licensees corrective action program as Condition Report CR -GGN -1-2017 -03404, compliance was restored within a reasonable period of time, the violation was not repetitive, and the violation was not willful, this violation is being treated as a n on-cited violation (NCV), consistent with Section 2.3.2.a of the Enforcement Policy. Traditional enforcement violations are not assigned a cross -cutting aspect.
05000416/FIN-2017002-052017Q2Grand GulfLicensee-Identified ViolationTitle 10 CFR 50.72(b)(2)(iv)(B) requires, in part, the licensee shall notify the NRC as soon as practical , and in all cases within 4 hours of the occurrence, of any event or 24 condition that results in actuation of the reactor protection system (RPS) when the reactor is critical. Contrary to the above, on April 4, 2017, the licensee did not notify the NRC within 4 hours of the occurrence of any event or condition that resulted in actuation of the RPS when the reactor was critical. Specifically, the licensee failed to notify the NRC within 4 hours after they performed a manual scram of the reactor due to a failure in the condensate system. The NRCs significance determination process is not designed to assess the significance of violations that impact or impede the regulatory process. Therefore, the issue of an untimely notification was assessed using the traditional enforcement process in accordance with the Enforcement Policy. The inspectors determined the violation to be at Severity Level IV in accordance with Enforcement Policy Section 6.9.d.9. Since this issue was entered into the licensees corrective action program as Condition Report CR -GGN -1-2017 -03331, compliance was restored within a reasonable period of time, the violation was not repetitive, and the violation was not willful, this violation is being treated as a non-cited violation (NCV), consistent with Section 2.3.2.a of the Enforcement Policy. Traditional enforcement violations are not assigned a cross -cutting aspect.
05000298/FIN-2017010-052017Q2CooperFailure to adopt appropriate procedures in accordance with 10 CFR Part 21Severity Level IV. The team identified a violation of 10 CFR 21.21(a), for the licensees failure to adopt appropriate procedures to evaluate deviations and failures to comply to identify those associated with substantial safety hazards. Specifically, Procedure EN-LI-108, 10 CFR 21 Evaluations and Reporting, Revision 5C0, was inadequate to ensure that the correct reportability call was made for a manufacturing flaw discovered in a relay that had resulted in a loss of safety function for the high pressure coolant injection system on April 25, 2016. In particular, the procedure (1) led the licensee to incorrectly conclude that a substantial safety hazard could not be created, (2) allowed a limited extent of condition in performing the substantial safety hazard evaluation such that similarly dedicated parts were not included in the scope, and (3) included incorrect guidance in Attachment 9.3. Corrective actions to restore compliance included re-evaluation of the defect under Part 21 requirements and a procedure adequacy review of the EN-LI-108-01 procedure. The licensee entered this issue into the corrective action program as Condition Reports CR-17-03936 and CR-17-04143. The failure to adopt appropriate procedures to evaluate deviations and failures to comply to identify those associated with substantial safety hazards, in violation of 10 CFR 21.21(a), was a performance deficiency. The NRCs reactor oversight process considers the safety significance of findings by evaluating their potential safety consequences. Using Inspection Manual Chapter 0612, Appendix B, Issue Screening, dated September 7, 2012, the team determined that the performance deficiency was of minor safety significance under the reactor oversight process because it involved a failure to make a report; however the underlying equipment failure was previously evaluated as having very low safety significance. The traditional enforcement process separately considers the significance of willful violations, violations that impact the regulatory process, and violations that result in actual safety consequences. Traditional enforcement applied to this finding because it involved a violation that impacted the regulatory process. The team used the NRC Enforcement Policy, dated November 1, 2016, to determine the significance of the violation. The inspectors determined that the violation was similar to Examples 6.9.d.10 and 6.9.d.13 of the Enforcement Policy, because although the procedure resulted in an inadequate reportability review and the issue was not reported as a manufacturing flaw, the licensee had reported some aspects of the event under the requirements of 10 CFR 50.73. As a result, the team determined that the violation should be classified as a Severity Level IV violation. Cross-cutting aspects are not assigned to traditional enforcement violations.
05000298/FIN-2017002-012017Q2CooperFailure to Assess Operability of Technical Specification System Functions during Surveillance TestingGreen . The inspectors identified a non- cited violation of Technical Specification 5.4.1.a, for the licensees fail ure to follow Station Procedure 0.26, Surveillance Program, Revision 70, and to assess the operability of alternate shutdown reactor pressure instrumentation during surveillance testing. Specifically, the licensee failed to assess the operability of the hig h pressure coolant injection turbine steam inlet pressure instrument that provides indications of reactor pressure for the alternate shutdown panel when the instrument was isolated during surveillance testing. As a result, operations personnel failed to r ecognize that the instrument was inoperable and failed to enter the appropriate technical specification action statements . As immediate corrective actions, the licensee validated that the alternate shutdown reactor pressure function was inoperable and that Technical Specification 3.3.3.2, Altern ate Shutdown System, Condition A, should have been entered, and generated a procedure change request to ensure T echnical Specification 3.3.3.2 would be entered during future surveillances . The licensee entered this deficiency into the corrective action program as Condition Report CR -CNS -2017- 02280. The licensees failure to assess the operability of alternate shutdown reactor pressure instrument ation when the high pressure coolant injection turbine inlet steam pr essure instrument was isolated for surveillance testing, in violation of Station Procedure 0.26, was a performance deficiency. The performance deficiency was determined to be more than minor , and therefore a finding, because it was associated with the hum an 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. Specifically, the alternate shutdown reactor pressure instrument was inoperable when the high pressure coolant injection turbine inlet pressure instrument was isolated for surveillance testing, and the appropriate technical specification action statement was not entered. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, the inspectors determined that the finding had very low safety significance (Green) because it: was not a design deficiency; did not repr esent a loss of system and/or function; did not represent an 3 actual loss of function; did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time; and did not result in the loss of a high safety -significant nontechnical specification train. The finding had a cross -cutting aspect in the area of human performance associated with work management. Specifically, the licensee failed to implement a process of planning, controlling, and executing work activities such that nuclear safety was the overriding priority, including the identification and management of risk commensurate with the isolation of the high pressure coolant injection turbine inlet pressure instrument during surveillance testing (H.5).
05000298/FIN-2017002-022017Q2CooperLoss of Control Room Ventilation Due to Ineffective Preventive Maintenance StrategyGreen . The inspectors reviewed a self -revealed, non- cited violation of Technical Specification 5.4.1.a , for the licensees failure to maintain work order instructions for control room supply fan maintenance resulting in the loss of the control room emergency filtration system. Specifically, prior to October 23, 2016, work order instructions for periodic preventive maintenance on the SF- C-1A supply fan failed to include industry recommended checks to ensure that the bearings were adequately engaged with the fan shaft, and failed to include proper work sequencing to ensure vibration data trending was meaningful. The ineffective preventive maintenance strategy resulted in the failure of the control room supply fan i nboard bearing during operation and a loss of the control room emergency filtration system function. Corrective actions to restore compliance included repair of the s upply fan and changes to improve the effectiveness of the fans preventive maintenance strategy. The licensee entered this deficiency into the corrective action program as Condition Report CR- CNS -2016- 07426. The licensees failure to maintain work order instructions for control room supply fan maintenance , in violation of Technical Specification 5.4.1.a , was a performance deficiency. The performance deficiency was more than minor , and therefore a finding, because it was associated with the structure, system, and component (SSC) and barrier performance attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers ( control room envelope) protect the public fro m radionuclide releases caused by accidents or events. Using Inspection Manual Chapter 0609, Appendix G, Attachment 1, Shutdown Operations Significance Process Phase 1 Initial Screening and Characterization of Findings, dated May 9, 2014, the inspectors determined that the finding had very low safety significance (Green) because the inspectors answered no to all of the Barrier Integrity screening questions. The finding had a cross -cutting aspect in the area of human performance associated with resourc es, because the licensee failed to ensure that personnel, equipment, procedures, and other resources we re available and adequate to support nuclear safety (H.1).
05000298/FIN-2017002-042017Q2CooperLicensee-Identified ViolationTechnical Specification 5.7.1 states, in part, that high radiation areas w ith dose rates greater than 0.1 rem/hr at 30 centimeters shall be barricaded and conspicuously posted as a high radiation area. Contrary to the above, on November 2, 2016, a high radiation area with does rates greater than 0.1 rem/hr at 30 centimeters was not barricaded and conspicuously posted as a high radiation area. Specifically, a radiation protection technician (RPT) identified an unposted high radiation area at the control rod drive (CRD) A pump filter area on r eactor building 881 feet southea st quadrant. D ose rates of 120 mrem/hr at 30 centimeters from the CRD filter were identified. This issue was identified as a result of a RPTs deliberate and focused observations during the course of performing their normal duties of performing radiological surveys. The licensee documented this issue in the corrective action program as Condition Report CR- CNS -2016 -00788. The finding was determined to be of very low safety significance (Green) because it was not an ALARA planning issue, there was no overexposure or potential for overexposure, and the licensees ability to assess dose was not compromised.
05000298/FIN-2017010-012017Q2CooperFailure to Assign Corrective Actions to Prevent Recurrence of High Pressure Coolant Injection FailureGreen. The team identified a non-cited violation of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to assign corrective actions to preclude repetition of a significant condition adverse to quality associated with the loss of the high pressure coolant injection system. Specifically, between July 28, 2016, and June 29, 2017, the licensee failed to assign or complete corrective actions to prevent recurrence to address the failure of a relay coil that resulted in a loss of safety function for the single train high pressure coolant injection system. Corrective actions to restore compliance included reevaluation of the corrective 3 actions assigned to the root cause of the condition and the creation of corrective actions to prevent recurrence for the condition. The licensee entered this deficiency into the corrective action program as Condition Report CR 17 03544. The licensees failure to assign corrective actions to preclude repetition of a significant condition adverse to quality, in violation of 10 CFR 50, Appendix B, Criterion XVI, was a performance deficiency. The performance deficiency was evaluated using Inspection Manual Chapter 0612, Appendix B, Issue Screening, dated September 7, 2012, and was associated with the Mitigating Systems cornerstone. The team determined that the performance deficiency was more than minor, and therefore a finding, because if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Specifically, the licensees failure to assign corrective actions to preclude repetition of a significant condition adverse to quality could reasonably result in the condition recurring and creating more safety-significant equipment failures. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, the inspectors determined that the finding had very low safety significance (Green) because it: was not a design deficiency; did not represent a loss of system and/or function; did not represent an actual loss of function; did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time; and did not result in the loss of a high safety-significant non-technical specification train. The finding had a cross-cutting aspect in the area of problem identification and resolution associated with resolution, because the licensee failed to ensure that the organization took effective corrective actions to address issues in a timely manner commensurate with their safety significance (P.3).
05000298/FIN-2017010-022017Q2CooperFailure to Perform Timely Operability DeterminationsGreen. The team identified a Green non-cited violation of Technical Specification 5.4.1.a, for the licensees multiple failures to immediately evaluate operability of degraded or nonconforming conditions. The team identified multiple examples of these operability determinations not being performed within one shift, as required by procedure. Further, aggregate data indicated routine noncompliance with procedural requirements to document operability immediately and without delay. The licensee entered this violation into its corrective action program as Condition Report CR-CNS-2017-03937, and began evaluating actions to restore compliance. Multiple failures to perform immediate operability determinations timely as required by station procedures is a performance deficiency. This performance deficiency is 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 system s that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, dated June 19, 2012, the inspectors determined that the finding had very low safety significance (Green) because it did not result in the loss of operability or functionality of any system or train. This finding has a consistent process cross-cutting aspect in the human performance cross-cutting area because operators failed to use a consistent, systematic approach to make decisions regarding operability using the organizations well-defined decision making process (H.13)
05000298/FIN-2017010-032017Q2CooperProgrammatic Failure to Identify and Correct Adverse TrendsGreen. The team identified a Green non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, for the licensees programmatic failure to promptly identify adverse trends and enter them into the corrective action program. Often, when adverse trends were identified, they were addressed using informal processes. This was particularly the case for safety culture-related trends such as adverse trends in organizational behaviors. The licensee entered this violation into its corrective action program as Condition Report CR-CNS-2017-03938, and took action to formalize identification processes for potential adverse trends. The programmatic failure to promptly identify adverse trends as required by station procedures was a performance deficiency. This performance deficiency is more than minor because if left uncorrected, it has the potential to become a more significant safety concern. Specifically, failure to arrest an adverse trend, particularly in organizational behaviors, could lead to increased likelihood of a worker-induced initiating event or a failure to effectively mitigate an accident. Using Inspection Manual Chapter 0609, Appendix A, dated June 19, 2012, the inspectors determined that the finding had very low safety significance (Green) because it did not result in the loss of operability or functionality of any system or train. This finding has a trending cross-cutting aspect in the problem identification and resolution cross- cutting area because the organization failed to use available information in the aggregate to identify programmatic and common cause issues (P.4).
05000298/FIN-2017010-042017Q2CooperFailure to Monitor No. 2 Diesel Generator under 50.65(a)(1) due to Inadequate Maintenance Rule EvaluationGreen. The team identified a non-cited violation of 10 CFR 50.65(a)(1)/(a)(2), for the licensees failure to perform an a(1) evaluation and establish a(1) goals when the No. 2 diesel generator a(2) preventive maintenance demonstration became invalid. Specifically, on April 28, 2017, the No. 2 diesel generator exceeded its performance criteria when it experienced a second maintenance rule functional failure, but the licensee failed to perform an associated a(1) evaluation. The licensee had failed to appropriately evaluate a February 4, 2017, failure associated with the No. 2 diesel generator jacket water heater failure in the Maintenance Rule Program and, as a result, the site failed to evaluate and monitor the equipment under 10 CFR 50.65(a)(1) as required. Corrective actions taken by the licensee to restore compliance included reevaluation of the February 4, 2017, functional failure and performance of an a(1) evaluation. The issue was entered into the licensees corrective action program as Condition Report CR-17-03930. The licensees failure to monitor the No. 2 diesel generator in accordance with the requirements of 10 CFR 50.65(a)(1), due to incorrectly evaluating one maintenance rule functional failure, in violation of 10 CFR 50.65(a)(1)/(a)(2), was a performance deficiency. The inspectors screened the performance deficiency using Inspection Manual Chapter 0612, Appendix B, Issue Screening, dated September 7, 2012, and determined that the issue was more than minor, and therefore a finding, because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, the inspectors determined that the finding had very low safety significance (Green) because it: was not a design deficiency; did not represent a loss of system and/or function; did not represent an actual loss of function; did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time; and did not result in the loss of a high safety-significant nontechnical specification train. The finding had a cross-cutting aspect in the area of problem identification and resolution associated with evaluation, because the licensee failed to ensure that the organization thoroughly evaluated 5 the No. 2 diesel generator issues to ensure that resolutions addressed causes and extent of conditions commensurate with their safety significance (P.2)
05000416/FIN-2017002-012017Q2Grand GulfFailure to Establish an Appropriate Preventative Maintenance Procedure for the HPCS Jockey PumpGreen . The inspectors reviewed a self -revealed, non- cited violation of Technical Specification 5.4.1.a, for the licensees failure to establish appropriate procedural instructions for performing preventative maintenance on the high pressure core spray jockey pump. Specifically, on January 27, 2017, the high pressure core spray jockey pump failed because the licensee did not establish a preventative maintenance procedure that prescribes oil analysis and additional performance trending for the high pressure core spray jockey pump every 6 months consistent with the licensees preventative maintenance strategy template. On January 29, 2017, the licensee completed repairs and returned the high pressure core spray jockey pump and high pressure core spray system to operable status . The licensee has also incorporated oil analysis and performance trending into the preventative maintenance for jockey pumps. This issue has been entered into the licensees corrective action program as Condition Report CR -GGN -2017- 0917. The failure to establish appropriate preventative maintenance instructions for the high pressure core spray jockey pump was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, 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. Specifically, the failure to establish appropriate preventative and predictive maintenance work instructions resulted in the unplanned inoperability and unavailability of the high pressure core spray system. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At -Power, dated June 19, 2012, and Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that the finding resulted in a loss of system and/or function; therefore, a detailed risk evaluation was performed. A senior reactor analyst performed a detailed risk evaluation in accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power. The NRC determined that the increase in core damage frequency for internal initiators was 1.59 E-7/year, and a bounding analysis of external initiators indicated that these events would not result in a 3 change in the color of the finding. Therefore, this finding is of very low safety significance (Green). The analyst also determined that an estimation of large early release frequency (LERF) was required. The result was an increase in LERF of 3.19E -8/year, which is of very low safety significance for LERF (Green). This finding had a cross -cutting aspect in the area of human performance associated with consistent process because the licensee did not use a consistent, systematic approach to make decisions. Specifically, the licensee did not use a consistent approach in developing a preventative maintenance strategy for the high pressure core spray jockey pump by utilizing the approved preventative maintenance strategy template (H.13).
05000440/FIN-2017002-022017Q2PerryImplementation of Enforcement Guidance Memorandum 11003, Revision 3From March 17, 2017, to March 24, 2017, Perry Nuclear Power Plant (PNPP) performed Operations with the Potential to Drain the Reactor Vessel (OPDRV) while in Mode 5 without an operable primary and secondary containment. An OPDRV is an activity that could result in the draining or siphoning of the reactor pressure vessel water level below the top of fuel, without crediting the use of mitigating measures to terminate the uncovering of fuel. Secondary containment was required by TS 3.6.4.1 to be operable during OPDRVs. Primary containment was required by TS 3.6.1.10 to be operable during OPDRVS. The required action for these specifications was to suspend OPDRV operations. Therefore, entering the OPDRV without establishing primary and secondary containment integrity was considered a condition prohibited by TS as defined by 10 CFR 50.73(a)(2)(i)(B).The NRC issued Enforcement Guidance Memorandum (EGM) 11003, Revision 3, on January 15, 2016, to provide guidance on how to disposition boiling water reactor licensee noncompliance with TS containment requirements during OPDRV operations. The NRC considers enforcement discretion related to secondary containment operability during Mode 5 OPDRV activities appropriate because the associated interim actions necessary to receive the discretion ensure an adequate level of safety by requiring licensees immediate actions to (1) adhere to the NRC plain language meaning of OPDRV activities; (2) meet the requirements which specify the minimum makeup flow rate and water inventory based on OPDRV activities with long drain down times; (3) ensure that adequate defense in depth is maintained to minimize the potential for the release of fission products with secondary containment not operable by (a) monitoring RPV level to identify the onset of a loss of inventory event, (b) maintaining the capability to isolate the potential leakage paths, (c) prohibiting Mode 4 (cold shutdown) OPDRV activities, and (d) prohibiting movement of irradiated fuel with the spent fuel storage pool gates removed in Mode 5; and (4) ensure that licensees follow all other Mode 5 TS requirements for OPDRV activities.The inspectors reviewed licensee event report (LER) 201700100 for potential performance deficiencies and/or violations of regulatory requirements. The inspectors also reviewed the stations implementation of the EGM during OPDRVs:The inspectors observed that the OPDRV activities were logged in the control room narrative logs, the log entry appropriately recorded the standby source of makeup water designated for the evolutions, and that defense in-depth criteria were in place.The inspectors noted that the reactor vessel water level was maintained at least 22 feet and 9 inches over the top of the reactor pressure vessel flange as required by TS 3.9.6. The inspectors also verified that at least one safety-related pump was the standby source of makeup designated in the control room narrative logs for the evolutions. The inspectors confirmed that the worst case estimated time to drain the reactor cavity to the reactor pressure vessel flange was greater than 24 hours.The inspectors reviewed Engineering Change documents which calculated the time to drain down during these activities and the feasibility of pre-planned actions the station would take to isolate potential leakage paths during these periods of time. The inspectors verified that the OPDRVs were not conducted in Mode 4 and that the licensee did not move irradiated fuel during the OPDRVs. The inspectors noted that PNPP had in place a contingency plan for isolating the potential leakage path and verified that two independent means of measuring reactor pressure vessel water level were available for identifying the onset of loss of inventory events.The inspectors verified that all other TS requirements were met during the March 17, 2017, to March 24 2017, OPDRVs with primary and secondary containment inoperable.Technical Specification 3.6.4.1 required, in part, that secondary containment shall be operable during OPDRV. Technical Specification 3.6.4.1, Condition C, required the licensee to initiate action to suspend OPDRV immediately when secondary containment is inoperable. Technical specification 3.6.1.10 required, in part, that primary containment shall be operable during OPDRV. Technical specification 3.6.1.10, Condition A, required the licensee initiate action to suspend OPDRV immediately when primary containment is inoperable. From March 17, 2017, to March 24, 2017, PNPP performed OPDRV activities while in Mode 5 without an operable primary or secondary containment. Specifically, the station performed the following OPDRV activities without an operable primary or secondary containment:draining of reactor recirculation loop B; replacement of 18 control rod drive mechanisms (unbolt and install);replacement of six instrument dry tubes;replacement of reactor recirculation pump B seal;replacement of reactor recirculation loop B flow control valve actuator;plugging of drain line appendages on reactor recirculation pump B; andlocal leak rate testing of the reactor water cleanup suction line containment isolation valves.The failure to perform OPDRV activities with operable primary and secondary containments is a violation of TS 3.6.1.10 and TS 3.6.4.1. Because the violation occurred during the discretion period described in EGM 11003, Revision 3, the NRC is exercising 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.In accordance with EGM 11003, Revision 3, each licensee that receives discretion must submit a license amendment request within 12 months of the NRC staffs publication in the Federal Register of the notice of availability for a generic change to the standard TS to provide more clarity to the term OPDRV. The inspectors observed thatPNPP is tracking the need to submit a license amendment request as commitment PYL1712101.This LER is closed. This inspection constituted one event follow-up sample as defined in IP 7115305.
05000293/FIN-2016011-062017Q1PilgrimDesign Change Not Appropriately Reviewed by EntergyThe NRC team identified a preliminary greater than Green finding and apparent violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, associated with Entergys failure to ensure that design changes were subject to design control measures commensurate with those applied to the original design and were approved by the designated responsible organization. Specifically, Entergy received a new style right angle drive for the A emergency diesel generator radiator blower fan from a vendor but failed to adequately review the differences in the design of the drives to identify potential new failure mechanisms for the part or the need for related preventive measures. Entergy entered this issue into the corrective action program as CR-PNP-2016-07443. The performance deficiency was more than minor because it was associated with the design control attribute of the Mitigating Systems cornerstone, and affected the associated cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, the team screened the finding for safety significance and determined that a detailed risk evaluation was required based on the A emergency diesel generator being inoperable for greater than the technical specification allowed outage time. Region I senior reactor analysts performed a detailed risk evaluation. The finding was preliminarily determined to be of greater than very low safety significance (greater than Green). The risk important sequences were dominated by external fire risk. Specifically, a postulated fire in the B 4 kilovolt (KV) switchgear room with a consequential loss of the unit auxiliary generator power supply, non-recoverable loss of off-site power (LOOP) to both safety buses A5 and A6, loss of the B emergency diesel generator with the conditional failure of the A emergency diesel generator, along with the loss of bus A8 feed (from the shutdown transformer or station blackout (SBO) diesel generator) to safety buses A5 and A6. The internal event risk was dominated by weather related LOOPs, failure of the A emergency diesel generator, with failure of the B emergency diesel generator and SBO diesel generator to run, along with failure to recover offsite power or the emergency diesel generators. See Attachment 1, A Emergency Diesel Generator Cooling Water System Degradation Detailed Risk Evaluation, for a detailed review of the quantitative criteria considered in the preliminary risk determination. The NRC team did not assign a cross-cutting aspect to this finding because the performance deficiency occurred in May 2000. Entergys program has undergone changes since May 2000, and the NRC team did not identify any recent examples of this performance deficiency. Other aspects of Entergys performance related to this issue are further discussed in Sections 5.10.3 and 6.3.4.
05000446/FIN-2017001-062017Q1Comanche PeakLicensee-Identified ViolationComanche Peak Unit 2, Operating License NPF-89, Condition 2.G, Fire Protection, requires, in part, that the licensee implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report through Amendment 87, and as approved in the Safety Evaluation Report and its supplements through Supplement 27. The stations approved fire protection program includes Fire Protection Report, Revision 29, Section 3.1 which requires, in part, that when fire detection equipment located inside of the containment building is inoperable then hourly monitoring of air temperature is performed as a compensatory measure. Contrary to the above, on November 22, 2016, licensee personnel identified that compensatory measures implemented for a failed detection system in the Unit 2 containment had not been implemented. The licensee had implemented a compensatory measure on December 3, 2015, to monitor containment temperature in the Unit 2 containment hourly due to a failed thermistor strip. On November 17, 2016, the licensee stopped monitoring temperature after restoring a different component to service. The licensee subsequently realized that the compensatory measure was still required and reinstated it on November 22, 2016. The violation is more than minor because it affected the protection against external events 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. Using Inspection Manual Chapter 0609, Attachment 04, Initial Characterization of Findings, and Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, the inspector determined that the violation is of very low safety significance (Green) because the finding did not affect the ability of either unit to achieve safe shutdown. The violation was entered into the licensees corrective action program as Condition Report CR-2016-009888.
05000445/FIN-2017001-052017Q1Comanche PeakLicensee-Identified ViolationTitle 10 CFR 50.54(q)(2) requires, in part, that licensees shall follow and maintain the effectiveness of an emergency plan that meets the planning standards of 10 CFR 50.47(b). Title 10 CFR 50.47(b)(2) requires, in part, that timely augmentation of response capabilities be available. The licensees emergency plan provides for the ability to augment response capabilities by use of a system to callout additional personnel to fill their emergency response organization (ERO) staffing requirements for declared emergencies. Contrary to the above, from January 5, 2017 until January 17, 2017, the licensee failed to ensure timely augmentation of response capabilities was available. Specifically, on January 5, 2017, the licensees corporate security office removed 32 members of the ERO from the licensees callout system, including eight personnel assigned to minimum staffing positions. The licensee identified the issue when, following an inadvertent actuation of the callout system on January 16, 2017, they discovered that multiple personnel were not called. The licensee restored all required personnel to the callout system on January 17, 2017. The violation is more than minor because it affected the ERO readiness attribute of the Emergency Preparedness cornerstone and impacted the cornerstone objective to ensure that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Using Inspection Manual Chapter 0609, Attachment 04, Initial Characterization of Findings, and Inspection Manual Chapter 0609, Appendix B, Emergency Preparedness Significance Determination Process, the inspector determined that the violation is of very low safety significance (Green) because the finding represented a failure to comply with planning standard (b)(2), and, using table 5.2-1, was screened as a Green finding because the deficiency did not cause more than one required ERO functional area to not be filled. The violation was entered into the licensees corrective action program as CR-2017-001524.
05000445/FIN-2017001-042017Q1Comanche PeakFailure to Promptly Correct a Condition Adverse to QualityGreen. The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, associated with the licensees failure to take timely corrective actions for a previously identified condition adverse to quality. Specifically, the licensee failed to verify the adequacy of the design of the Unit 1 120 VAC vital bus inverter 1PC1 with respect to use of alternate AC power to the inverter. The 120 VAC calculation did not properly account for low voltage when the buses are supplied from their alternate source. This issue does not represent an immediate safety concern because, following the inspectors identification, the licensee performed an operability evaluation which established a reasonable expectation of operability. The licensee implemented immediate corrective actions by entering the issues into the corrective action program for resolution and performed an operability determination for the identified degraded conditions. The licensee entered this issue into their corrective action program as CR-2017-001296. The licensees failure to take timely and adequate corrective actions to correct a condition adverse to quality was a performance deficiency. The performance deficiency is more than minor, and therefore a finding, because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to correct the low voltage susceptibility resulted in delayed restoration of a bus following the failure of the swing inverter to sync. Using Inspection Manual Chapter 0609, Attachment 04, Initial Characterization of Findings, dated October 7, 2016, and Inspection Manual Chapter 0609, Appendix A, Significance Determination Process for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined the finding was of very low safety significance (Green) because the finding: (1) was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality, (2) did not represent a loss of system and/or function, (3) did not represent an actual loss of function of at least a single train for longer than its allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time, and (4) does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant for greater than 24 hours in accordance with the licensees maintenance rule program. The finding has a human performance cross-cutting aspect associated with resources, in that, the licensee failed to ensure that resources were adequate to support nuclear safety (H.1).
05000445/FIN-2017001-032017Q1Comanche PeakUse of Non-Design Fouling Factor for Component Cooling Water Heat Exchanger in Station Service Water Tornado Missile CalculationGreen. The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, involving the failure to use the design fouling factor for the component cooling water heat exchanger in a design basis calculation evaluating a tornado missile strike of station service water system piping. The licensee implemented immediate corrective actions by entering the issues into the corrective action program for resolution and performed an operability determination for the identified degraded conditions. The licensee entered this issue into their corrective action program as Issue Report IR-2017-001465. The inspectors determined that the failure to use the design fouling factor for the component cooling water heat exchanger in the tornado missile analysis of the station service water system discharge piping was a performance deficiency. This finding 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. Specifically, the use of a non-conservative heat exchanger fouling factor in a design basis accident analysis resulted in a more restrictive temperature limit (i.e., less than the technical specification allowed value) of the safe shutdown impoundment. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that (1) did not represent a loss of operability or functionality; (2) did not represent an actual loss of safety function of the system or train; (3) did not result in the loss of one or more trains of non-technical specification equipment; and (4) did not screen as potentially risk-significant due to seismic, flooding, or severe weather. The inspectors determined that this finding did not have a cross-cutting aspect because the most significant contributor to the performance deficiency did not reflect current licensee performance. Specifically, the licensee performed the calculation in 1988, therefore, the performance deficiency occurred outside of the nominal three-year period for present performance.
05000445/FIN-2017001-022017Q1Comanche PeakFailure to Evaluate Heat Loads on Control Room Air Conditioning SystemGreen. The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for the licensees failure to properly evaluate heat loads on the control room air conditioning system. Specifically, the licensee used a non-conservative assumption for the number of persons in the control room envelope when calculating the required capacity of the system. The licensee had assumed there would only be six personnel in the technical support center (which is included in the control room envelope) during a design basis event. However, the emergency plan nominally staffed the technical support center with 25 station personnel, and an additional five NRC personnel. The licensee implemented immediate corrective actions by entering the issues into the corrective action program for resolution and performed an operability determination for the identified degraded condition. The licensee entered this issue into their corrective action program as Condition Report CR-2017-000744. The failure to evaluate heat loads to determine the required system capacity was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the design control attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Attachment 04, Initial Characterization of Findings, dated October 7, 2016, and Inspection Manual Chapter 0609, Appendix A, Significance Determination Process for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined the finding was of very low safety significance (Green) because the finding: (1) was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality, (2) did not represent a loss of system and/or function, (3) did not represent an actual loss of function of at least a single train for longer than its allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time, and (4) does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant for greater than 24 hours in accordance with the licensees maintenance rule program. The inspectors determined that no cross-cutting aspect was assigned because the performance deficiency was not reflective of present performance.
05000445/FIN-2017001-012017Q1Comanche PeakFailure to Maintain B.5.b Equipment in a State of Readiness to Support Mitigation StrategiesGreen. The inspectors identified a non-cited violation of 10 CFR 50.54(hh)(2), Conditions of Licenses, involving the licensees failure to maintain available equipment needed to implement mitigating strategies to provide makeup to steam generators following loss of large areas of the plant due to explosions or fire. Specifically, the licensee failed to maintain available a portable alternate mitigation equipment pump related to the steam generator makeup strategy. As an immediate corrective action the licensee put temporary heaters in place for the alternate mitigation equipment pump to ensure the equipment was stored at temperatures greater than 32 degrees Fahrenheit pending further evaluation. The licensee entered this issue into their corrective action program as Condition Report CR-2016-010832. The failure to maintain all necessary equipment available to implement mitigating strategies as required by regulations and conditions of the operating license was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the design control 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. Using Inspection Manual Chapter 0609, Appendix L, B.5.b Significance Determination Process, dated December 24, 2009, the inspectors determined the finding was of very low safety significance (Green) because it resulted in an unrecoverable unavailability of an individual mitigating strategy but did not result in multiple unavailable mitigating strategies, or loss of all on-site, self-powered, portable pumping capability. The inspectors did not assign a cross-cutting aspect because the performance deficiency was not reflective of present performance.
05000293/FIN-2016011-012017Q1PilgrimFailure to Identify All Root Causes of a Significant Condition Adverse to QualityThe NRC team identified a Green non-cited violation of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion XVI, Corrective Action, because Entergy did not adequately determine all root causes associated with a significant condition adverse to quality related to the failure to identify, evaluate, and correct the A SRVs failure to open upon manual actuation during a plant cooldown on February 9, 2013. Specifically, Entergy did not establish adequate measures to assure that the cause of a significant condition adverse to quality, inadequate shift manager operability determination rigor and its associated causes, were adequately determined and corrective action taken to preclude repetition. Entergys immediate corrective actions included planning to conduct operations management face-to-face conversations with shift manager qualified individuals to reinforce the shift managers responsibility for operability and functionality determination accuracy and rigor. Entergy entered this issue into the corrective action program as CRPNP-2017-00363 and CR-PNP-2017-00828. The performance deficiency was more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Specifically, if left uncorrected, the performance deficiency could have the potential to result in repetition of a failure to identify, evaluate, and correct an SRVs failure to open or a similar significant condition adverse to quality. The NRC team evaluated the finding using Exhibit 2, Mitigating Systems Screening Questions, of IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, and determined this finding did not affect the design or qualification of a mitigating structure, system, or component; represent a loss of system and/or function; involve an actual loss of function of at least a single train or two separate safety systems for greater than its technical specification-allowed outage time; or represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant. Therefore, the NRC team determined the finding was of very low safety significance (Green). The NRC team determined that the finding had a cross-cutting aspect in the area of Human Performance, Avoid Complacency, because individuals did not recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Specifically, Entergy incorrectly assumed that CR-PNP-2013-00825 contained inadequate information to determine that the A SRV had not opened, and this assumption ultimately impacted the root cause results documented in CR-PNP-2016-01621 (H.12).
05000293/FIN-2016011-022017Q1PilgrimFailure to Establish Corrective Actions to Preclude Repetition of a Significant Condition Adverse to QualityThe NRC team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, because Entergy did not implement CAPRs for a significant condition adverse to quality identified in root cause evaluation CR-PNP-2016-00716, Implementation of the Corrective Action Program, Revision 2. Specifically, the team identified that CAPRs for Entergys continued weaknesses in the implementation of the corrective action program were inadequate. Entergy entered this issue into their corrective action program for further evaluation as CR-PNP-2017-00053, CR-PNP-2017-00410, and CR-PNP-2017-01134. The performance deficiency was more than minor because if left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, the failure to preclude repetition of this significant condition adverse to quality could result in continuing weaknesses in implementation of the corrective action program, which was designated as a fundamental problem, and thus a contributing factor for PNPS Column 4 performance. Additionally, weaknesses with corrective action program implementation could result in equipment issues where operability is not maintained. The NRC team evaluated the finding using Exhibit 2, Mitigating Systems Screening Questions, of IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, and determined this finding did not affect the design or qualification of a mitigating structure, system, or component; represent a loss of system and/or function; involve an actual loss of function of at least a single train or two separate safety systems for greater than its technical specificationallowed outage time; or represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant. Therefore, the NRC team determined the finding was of very low safety significance (Green). The NRC team determined that the finding had a cross-cutting aspect in the area of Human Performance, Procedure Adherence, because individuals did not follow processes, procedures, and work instructions. Specifically, Entergy did not follow procedure EN-LI-102, which provides the station standards for crafting a corrective action and states, in part, that the corrective action descriptions must be worded to ensure that the adverse condition or cause/factor is addressed (H.8).
05000293/FIN-2016011-032017Q1PilgrimFailure to Issue Appropriate Corrective Actions to Preclude Repetition for the Causes of the September 2016 ScramThe NRC team identified a Green finding because Entergy did not issue appropriate CAPRs in accordance with Entergy procedure EN-LI-102, Corrective Action Process, Revision 28. Specifically, Entergy did not issue adequate CAPRs associated with Root Cause 1 of the feedwater regulating valve failure in September 2016 that resulted in a manual scram. As a result of the NRC teams questions, Entergy issued procedure 1.13.2, Vendor and Technical Information Reviews, Revision 0, as continuous use to ensure that planners will always have the checklist in-hand when planning work to ensure that appropriate vendor technical information is always included in applicable work instructions. Entergy entered the NRC teams concerns in the corrective action program as CR-PNP-2017-00687 and CR-PNP-2017-00936. The performance deficiency was more than minor because it is associated with the equipment performance attribute of the Initiating Events cornerstone and if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Specifically, if left uncorrected, the performance deficiency could have the potential to result in repetition of a significant condition adverse to quality, loss of control of feedwater regulating valve 642A and a manual scram. The NRC team evaluated the finding using Exhibit 1, Initiating Events Screening Questions, of IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, and determined this finding did not cause a reactor trip or the loss of mitigation equipment relied upon to transition the plant from the onset of a trip to a stable shutdown condition. Therefore, the NRC team determined the finding was of very low safety significance (Green). The NRC team determined that the finding had a cross-cutting aspect in the area of Human Performance, Procedure Adherence, because individuals did not follow processes, procedures, and work instructions. Specifically, Entergy did not follow procedure EN-LI-102, which provides the station standards for crafting a corrective action and states, in part, that the corrective action descriptions must be worded to ensure that the adverse condition or cause/factor is addressed (H.8).
05000293/FIN-2016011-042017Q1PilgrimProgrammatic Issue with Implementation of the Operability Determination ProcessThe NRC team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings. Specifically, the NRC team identified a programmatic issue because in some cases, Entergy did not enter the operability determination process when appropriate, and, when the process was entered, did not adequately document the basis for operability, in accordance with Procedure ENOP-104, Operability Determination Process, Revision 11. In each of the examples discussed, though the basis for operability was not adequate, all components were determined to be operable upon further evaluation. Entergy entered this issue into their corrective action program as CR-PNP-2017-00626. The performance deficiency was more than minor because if left uncorrected, could lead to a more significant safety issue. Specifically, the failure to enter and document a basis for operability could lead to not recognizing inoperable safety-related equipment, and place the reactor at a higher risk of core damage in a design basis accident. The NRC team evaluated the finding using Exhibit 2, Mitigating Systems Screening Questions, of IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, and determined this finding did not affect the design or qualification of a mitigating structure, system, or component; represent a loss of system and/or function; involve an actual loss of function of at least a single train or two separate safety systems for greater than its technical specification-allowed outage time; or represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant. Therefore, the NRC team determined the finding was of very low safety significance (Green). This finding had a cross-cutting aspect in the area of Human Performance, Teamwork. Specifically, the operations and engineering departments did not demonstrate a strong sense of collaboration and cooperation with respect to holding each other accountable when performing operability determinations to ensure nuclear safety is maintained (H.4).
05000293/FIN-2016011-052017Q1PilgrimFailure to Establish Corrective Actions to Address Scope of Procedure Quality IssuesThe NRC team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, because Entergy implemented inadequate corrective actions to address the procedure quality issues identified in CR-PNP-2016-02058. Specifically, Entergy inappropriately limited their corrective actions to those procedures that increased integrated risk above normal, and did not include other types of safety-related procedures that did not meet their procedure quality standards and resulted in procedure quality being a problem area. Entergy entered this issue into their corrective action program for further evaluation as CR-PNP-2017-00400. The performance deficiency was more than minor because it affected the procedure quality attribute of the Mitigating Systems cornerstone, and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Entergy limited corrective actions to procedures that increased integrated risk above normal or trip sensitive and failed to include other procedures associated with safety-related components that reflected the broader population reviewed during the collective evaluation. The NRC team evaluated the finding using Exhibit 2, Mitigating Systems Screening Questions, of IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, and determined this finding did not affect the design or qualification of a mitigating structure, system, or component; represent a loss of system and/or function; involve an actual loss of function of at least a single train or two separate safety systems for greater than its technical specificationallowed outage time; or represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant. Therefore, the NRC team determined the finding was of very low safety significance (Green). The NRC team determined that this finding had a cross-cutting aspect related to Human Performance, Resources, because the leaders failed to ensure that personnel, equipment, procedures, and other resources are available and adequate to support nuclear safety. Specifically, based on available resources, Entergy chose to limit the scope of safety-related procedures being revised to only those that resulted in high integrated risk or were trip sensitive (H.1).
05000293/FIN-2016011-082017Q1PilgrimFailure to Adequately Monitor the Performance of Maintenance Rule Scoped ComponentsThe NRC team identified a Green non-cited violation of 10 CFR 50.65(a)(2), Requirements for monitoring the effectiveness of maintenance at nuclear power plants. Specifically, Entergy did not demonstrate that the performance of 18 maintenance rule scoped components was effectively controlled through the performance of appropriate preventive maintenance, and did not establish goals and monitoring in accordance with 10 CFR 50.65(a)(1). Entergys immediate corrective action was to initiate a CR to evaluate moving the affected systems to 10 CFR 50.65(a)(1) monitoring requirements. Entergy entered this issue in the corrective action program as CR-PNP-2017-00401. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, Entergy failed to demonstrate that the performance of the 18 maintenance rule scoped components was being effectively controlled through the performance of appropriate preventive maintenance which adversely impacts the reliability of those systems. The NRC team evaluated the finding using Exhibit 2, Mitigating Systems Screening Questions, of IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, and determined this finding did not affect the design or qualification of a mitigating structure, system, or component; represent a loss of system and/or function; involve an actual loss of function of at least a single train or two separate safety systems for greater than its technical specificationallowed outage time; or represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant. Therefore, the NRC team determined the finding was of very low safety significance (Green). The finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, in that Entergy failed to thoroughly evaluate and ensure that resolution of the identified issue, maintenance not being performed on maintenance rule scoped components, included reclassifying the components as necessary. Specifically, Entergy failed to demonstrate that the performance of Maintenance rule scoped components was effectively controlled through the performance of appropriate preventive maintenance, or through performance goals and monitoring. (P.2).
05000293/FIN-2016011-092017Q1PilgrimIneffective Corrective Actions to Address Conditions Adverse to Quality Regarding Components in Contact with or Close Proximity to the Drywell LinerThe NRC team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, associated with Entergys failure to correct a condition adverse to quality affecting safety-related equipment. Specifically, during a previous NRC inspection in August 2016, inspectors identified numerous locations in the drywell where non-seismic equipment was either in contact, or close proximity, with the drywell liner and had caused damage. Entergy initiated CRs and performed an operability evaluation for the identified issues. However, following a review of these CRs, the NRC team determined that Entergy failed to take corrective actions to address the condition adverse to quality. Entergy entered this issue into the corrective action program as CR-PNP-2016-09346 and CR-PNP-2016-09377 to perform an extent of condition review, secure the loose grating that had caused damage to the liner, and evaluate the need for a clearance criteria between components such as floor grating and support structures and the containment liner. The performance deficiency was more than minor because it was associated with the configuration control attribute of the Barrier Integrity cornerstone and affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Using IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 3, Barrier Integrity Screening Questions, the NRC team determined that this finding was of very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containment (valves, airlocks, etc.), containment isolation system (logic and instrumentation), and heat removal components. This finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, because the engineering evaluation of the degraded condition identified by the inspectors did not thoroughly evaluate the containment liner issues to ensure that resolutions address causes and extents of condition commensurate with their safety significance (P.2).
05000293/FIN-2016011-102017Q1PilgrimFailure to Promptly Correct a Condition Adverse to Quality for the Residual Heat Removal SystemThe NRC team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, because Entergy did not take timely corrective action for a previously identified condition adverse to quality. Specifically, Entergy failed to adequately resolve, through repair or adequate evaluation, gasket leakage on the B residual heat removal heat exchanger, which resulted in continued degradation and leakage for the heat exchanger gasket. Entergy did not consider this leakage as a degraded condition, with the potential to impact both the operability of the residual heat removal system, and PNPSs licensing basis with regards to leakage of a closed loop system outside of containment. After the NRC team raised the issue, Entergy performed an operability determination that established a reasonable expectation of operability pending implementation of corrective actions. Entergy entered this issue into their corrective action program as CR-PNP-2016-09725. The performance deficiency 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 availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to correct identified gasket leakage resulted in continued degradation and leakage of the heat exchanger gasket. The NRC team evaluated the finding using Exhibit 2, Mitigating Systems Screening Questions, of IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, and determined this finding did not affect the design or qualification of a mitigating structure, system, or component; represent a loss of system and/or function; involve an actual loss of function of at least a single train or two separate safety systems for greater than its technical specification-allowed outage time; or represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant. Therefore, the NRC team determined the finding was of very low safety significance (Green). The finding had a cross-cutting aspect in Human Performance, Conservative Bias, because Entergy failed to use decision making practices that emphasize prudent choices over those that are simply allowable (H.14).
05000293/FIN-2016011-112017Q1PilgrimFailure to Adequately Develop and Implement Targeted Performance Improvement PlansThe NRC team identified a Green finding because Entergy did not adequately develop and implement a CAPR of a root cause related to a Category A CR, as required by Entergy Procedure EN-LI-102, Corrective Action Program. Specifically, Entergy did not adequately develop and implement the Targeted Performance Improvement Plans, which were designated as a CAPR for the root cause for the Nuclear Safety Culture Fundamental Problem. Entergy documented this issue in the corrective action program for further evaluation as CR-PNP-2017-00406. The performance deficiency was more than minor because if left uncorrected, it could lead to a more significant safety concern. Specifically, inadequate implementation of the Targeted Performance Improvement Plans could result in recurrence of a culture in which leaders are not holding themselves and their subordinates accountable to high standards of performance, resulting in continuing performance issues at the station. The NRC team evaluated the finding using Exhibit 2, Mitigating Systems Screening Questions, of IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, and determined this finding did not affect the design or qualification of a mitigating structure, system, or component; represent a loss of system and/or function; involve an actual loss of function of at least a single train or two separate safety systems for greater than its technical specification-allowed outage time; or represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant. Therefore, the NRC team determined the finding was of very low safety significance (Green). This finding had a cross-cutting aspect in the area of Human Resources, Change Management, because leaders did not use a systematic process for evaluating and implementing change so that nuclear safety remains the overriding priority. In this case, PNPS leaders did not apply sufficient rigor in development and implementation of the Targeted Performance Improvement Plans such that they would be an adequate method to drive and sustain positive changes in the stations safety culture (H.3).
05000293/FIN-2016011-122017Q1PilgrimLicensee-Identified Violation10 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, and shall be accomplished in accordance with those structures, procedures, and drawings. Entergy procedure EN-DC-148, Vendor Manuals and Vendor Re-Contact Process, Revision 6, requires, in part, that the station update vendor manuals every three years. Contrary to this, in July 2016, PNPS determined through a self-assessment that they had 13 vendor manuals that had not been evaluated for changes within 3 years. The NRC team determined that this finding did not affect the design or qualification of a mitigating structure, system or component; did not represent a loss of a system and/or function; did not result in loss of a train or two safety systems greater than any technical specification allowed outage time; did not result from an actual loss of safety function; and did not involve loss of any external event mitigating system. Consequently, the NRC team determined that this performance deficiency screened as having very low safety significance (Green). PNPS documented this issue in their corrective action program as CR-PNP-2016-05115.
05000293/FIN-2016011-132017Q1PilgrimLicensee-Identified Violation10 CFR 50.54(q)(2) requires, in part, that the licensee follow and maintain the effectiveness of an emergency plan to meet the planning standard of 10 CFR 50.47(b)(4). Specifically, the licensee was to maintain the necessary equipment to support the effectiveness of EALs. Contrary to these requirements, PNPS identified in CR-PNP-2016-01491 that on three past occasions (March 15 through August 8, 2012; September 4 through October 14, 2012; and June 4 through June 14, 2015) both trains of the H2O2 monitors and the Post-Accident Sampling System were unavailable to ensure the effectiveness of EAL 24, Deflagration concentrations exist inside PC, for the potential loss of the containment barrier within the Fission Product Barrier category of the EALs. This issue meets the criteria for very low safety significance (Green) because, due to other EALs, an appropriate emergency declaration could have been made in an accurate and timely manner.
05000293/FIN-2016011-072017Q1PilgrimFailure to Report Condition Prohibited by Technical Specifications and a Safety System Functional FailureThe NRC team identified a Severity Level IV non-cited violation of 10 CFR 50.73, Licensee Event Report System, associated with Entergys failure to submit a licensee event report within 60 days following discovery of an event meeting the reportability criteria. Specifically, on September 28, 2016, Entergy identified the A emergency diesel generator was inoperable. The NRC team determined that the condition was prohibited by technical specifications and the inoperability of the A emergency diesel generator existed for a period of time longer than allowed by Technical Specification 3.5.F, Core and Containment Cooling Systems. This was also reportable as a safety system functional failure. Entergy entered this issue into the corrective action program as CR-PNP-2016-09552. Because this performance deficiency had the potential to impact the NRCs ability to perform its regulatory function, the NRC team evaluated the performance deficiency using traditional enforcement. The violation was evaluated using Section 2.3.11 of the NRC Enforcement Policy, because the failure to submit a required licensee event report may impact the ability of the NRC to perform its regulatory oversight function. In accordance with Section 6.9.d, Example 9, of the NRC Enforcement Policy, this violation was determined to be a Severity Level IV non-cited violation. Because this violation involves the traditional enforcement process and does not have an underlying technical violation, the NRC team did not assign a cross-cutting aspect to this violation, in accordance with IMC 0612, Appendix B.
05000313/FIN-2016004-012016Q4Arkansas NuclearFailure to Pre-plan Walkdown to Avoid Impacting Safety BusGreen. The inspectors documented a self-revealed finding and associated non-cited violation of Unit 1 Technical Specification 5.4.1.a, for the failure to properly pre-plan and perform a pre-modification walkdown in the Unit 1 train A safety-related switchgear room so that the walkdown would not adversely affect the performance of train. As a result, licensee personnel inadvertently de-energized the A3 switchgear and associated ac buses, which resulted in the loss of one train of spent fuel pool cooling. Operators restored spent fuel pool cooling, the licensee evaluated the human error and performed a training stand-down to ensure pre-job walkdowns did not impact plant equipment. The licensee entered this issue into the corrective action program as Condition Report CR-ANO-1-2016-04356. The failure to perform a plant walkdown in a manner that did not impact safety-related switchgear is a performance deficiency. The performance deficiency is more than minor because it adversely affected the human performance attribute of the barrier integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, de-energizing the safety-related switchgear resulted in the loss of one train of spent fuel pool cooling and an increase in risk level from Green to Yellow. The inspectors evaluated the finding with NRC Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 3, Barrier Integrity Screening Questions, because the appendix provides the most applicable guidance, regardless of whether the unit was at-power or shutdown. The inspectors determined that the finding screened as having very low safety significance (Green) because the finding did not cause the spent fuel pool to exceed the maximum analyzed temperature, did not damage fuel cladding, did not result in a loss pool water inventory below the minimum analyzed level, and did not affect the pool neutron absorber or soluble boron concentration. The inspectors determined this finding has a cross-cutting aspect in the human performance area of Avoid Complacency, because the primary cause of the performance deficiency involved the failure to plan for the possibility of mistakes and use appropriate error reduction tools. (H.12)
05000482/FIN-2016004-012016Q4Wolf CreekFailure to Adequately Establish and Adjust Preventive Maintenance for Emergency Diesel Generator Excitation System DiodesGreen. The inspectors identified a non-cited violation of Technical Specification 5.4.1.a, for the licensees failure to adequately develop and adjust preventive maintenance activities in accordance with Procedure AP 16B-003, Planning and Scheduling Preventive Maintenance, Revision 5. Specifically, the licensee did not create a preventive maintenance task for emergency diesel generator excitation system diodes, which resulted in degradation of the excitation system diodes in emergency diesel generator B. The licensee restored compliance by establishing preventive maintenance tasks 49286, 49287, 49288, and 49289, which refurbish the power rectifier assemblies and replace the diodes on a 12-year replacement frequency. The licensee entered this issue into the corrective action program as Condition Report 88665. The failure to adequately develop and adjust emergency diesel generator excitation system diode preventive maintenance activities in accordance with Procedure AP 16B-003, Planning and Scheduling Preventive Maintenance, was a performance deficiency. This finding is 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 availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated the finding using IMC 0609, Appendix A, Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, this finding was not a deficiency affecting the design or qualification of a mitigating structure, system, or component that maintained its operability or functionality; the finding did not represent a loss of system and/or function; the finding did not represent an actual loss of function of at least a single train for greater than its Technical Specification allowed outage time; and the finding did not represent an actual loss of function of one or more non-Technical Specification trains of equipment designated as high safety-significant. Therefore, the finding was of very low safety significance (Green). This finding has a cross-cutting aspect in the area of problem identification and resolution, operating experience, because the organization did not systematically and effectively evaluate relevant internal and external operating experience in a timely manner. This issue is indicative of current performance because the station did not take any formal corrective actions to address the stations failure to adequately consider operating experience (P.5)
05000313/FIN-2016004-022016Q4Arkansas NuclearFailure to Design Pipe Support for VibrationGreen. The inspectors documented a self-revealed finding and associated non-cited violation of 10 CFR 50 Appendix B Criterion III for the licensees failure to verify that the decay heat removal (DHR) system drain piping configuration and supports could withstand vibrations created during low pressure and high flow conditions. As a result, a cracked weld and unisolable leak in the DHR system occurred due to high cycle fatigue caused by those conditions. To correct this issue, the licensee repaired the leaking weld and designed and installed a new piping support and piping configuration to reduce vibrations during the expected operating conditions. The licensee entered this issue into the corrective action program as Condition Report CR-ANO-1-2016-03225. The failure to design the decay heat removal system piping to withstand expected vibrations from the systems cavitating venturis is a performance deficiency. The performance deficiency is more than minor because it was associated with the design control 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, inadequate design of the DHR system piping support resulted in a leak that could have challenged the capability of both trains of the DHR system during shutdown on September 29, 2016. The inspectors performed an initial screening of the finding in accordance with NRC Inspection Manual Chapter (IMC) 0609, "Significance Determination Process," issued October 7, 2016, and were directed to IMC 0609, Appendix G, Attachment 1, "Shutdown Operations Significance Determination Process Phase 1 Screening and Characterization of Findings, since the finding pertained to a degraded condition while the plant was shutdown. Using IMC 0609, Appendix G, Attachment 1, dated May 9, 2014, the inspectors determined that the finding required a Phase 2 evaluation. A senior reactor analyst performed a Phase 2 evaluation in accordance with IMC 0609, Appendix G, Attachment 2, Phase 2 Significance Determination Process Template for PWR during Shutdown, dated February 28, 2005. The senior reactor analyst performed a Phase 2 evaluation which used realistic break characteristics and plant configuration changes to determine the significance to be of very low safety significance (Green). The inspectors determined this finding did not have a cross-cutting aspect because the most significant contributor did not reflect current licensee performance. Specifically, the licensee last reviewed and modified the pipe support configuration in 1996
05000313/FIN-2016004-032016Q4Arkansas NuclearLicensee-Identified ViolationThe licensee identified that the Unit 1 emergency diesel generator governors were left in droop mode at all times, so that during a loss of offsite power the speed and frequency of the EDGs would decrease as loading increased and cause a reduction in speed and capability from safety-related motors. The licensee determined that some EDG-powered safety-related motors would not have been capable of providing the required flow rate for a short period of time, but this did not prevent them from performing their safety function. Title 10 CFR Part 50, Appendix B, Criterion V, Instruction, Procedures, & Drawings, states, in part, that activities affecting quality shall be prescribed by procedures of a type appropriate to the circumstance. Contrary to the above, as of November 2, 2016, the procedure for Unit 1 EDG operations, an activity affecting quality, was not appropriate to the circumstance. Specifically, Procedure OP-1104.036, Emergency Diesel Generator Operation, Revision 74, did not state to set the speed droop settings for both A and B EDGs to zero when not load sharing with another power source and did not specify this as a requirement for the EDGs when in an emergency standby condition. The licensee immediately set the speed droop settings for both EDGs to zero and changed the procedure. The licensee documented the issue in their corrective action program as Condition Report CR-ANO-1-2016-04333. Using NRC Inspection Manual Chapter 0609, Appendix A, "The Significance Determination Process (SDP) For Findings At-Power, dated June 19, 2012, the inspectors determined the finding to be of very low safety significance (Green) because the deficiency did not result in a loss of a safety function.
05000313/FIN-2016004-042016Q4Arkansas NuclearLicensee-Identified ViolationDuring the fall 2016 Unit 1 refueling outage, the licensee foreign object search and retrieval (FOSAR) inspections in the steam generator bowls and reactor vessel identified a number of foreign objects, including an 8-inch metal rod. Discussions with the licensee indicated that some of the debris constituted foreign material that should have been prevented from being introduced into the RCS by the foreign material exclusion program. The inspectors concluded that the foreign material was most likely introduced during the previous refueling outage. During the prior operating cycle, the licensees chemistry sampling identified increased RCS activity, and subsequent fuel bundle examinations of fuel removed from the core identified wear marks through the cladding of two adjacent fuel pins. The fuel assembly with the damage was not placed back into the RCS. Since there was no evidence of broken components inside the RCS, the licensee concluded that the most likely cause was the introduction of foreign material. While it was not possible to determine whether any of the foreign material had actually caused the fuel damage, the inspectors concluded that the licensee had failed to control foreign material and prevent it from entering the RCS. Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality shall be accomplished in accordance with documented instructions, procedures, or drawings of a type appropriate to the circumstances. Licensee Procedure EN-MA-118, Foreign Material Exclusion, Revision 10, an Appendix B quality-related procedure, provides instructions for controlling foreign material. Procedure EN-MA-118, Step 5.5, requires, in part, that all material and tools that were introduced to the FME zone are accounted for. Contrary to the above, between January 25, and March 1, 2015, the licensee failed to ensure that all material and tools that were introduced to the FME zone were accounted for. Specifically, the licensee failed to maintain adequate FME control, leading to two damaged cladding pins and slightly elevated dose rates in the RCS piping, as well as another piece of metallic FME in the vessel, as documented in CR-ANO-1-2016-03340. This issue was documented in the licensees corrective action program under CR-ANO-1-2016-03521. Corrective actions taken include a search for the foreign material and permanent removal of the fuel assembly from the core. Prior to 2012, the NRCs Significance Determination Process in IMC 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, contained guidance to screen all more than minor performance deficiencies affecting fuel barriers to very low safety significance. The inspection manual chapters were restructured in 2012, and the screening was inadvertently omitted, though the NRC was in the process of reinstating that same guidance. Therefore, after consultation with the Office of Nuclear Reactor Regulation, the inspectors determined that this finding is of very low safety significance (Green).
05000298/FIN-2016008-012016Q3CooperPossible Failure to Ensure that the Assumptions in the Engineering Analysis Remain ValidAs part of the transition to a performance-based, risk-informed fire protection program, the licensee adopted the requirements of NFPA 805. NFPA 805 requires the following in Section 2.6: Monitoring. A monitoring program shall be established to ensure that the availability and reliability of the fire protection systems and features are maintained and to assess the performance of the fire protection program in meeting the performance criteria. Monitoring shall ensure that the assumptions in the engineering analysis remain valid. The team reviewed selected samples of equipment monitored by the licensee using Procedure 3-CNS-DC-357, NFPA 805 Monitoring Program, Revision 0, to ensure that the licensees program properly implemented the requirements of NFPA 805, Section 2.6. The team also reviewed Engineering Report Number ER2015-002, NFPA 805 Fire Protection Monitoring Program, Revision 2. The team observed that for components used in the fire probabilistic risk assessment, the unavailability time for those components was monitored using the existing maintenance rule monitoring program. These components included the: Control rod drive pumps Core spray pumps Emergency diesel generators Emergency station service transformer Startup station service transformer High pressure core spray pump Instrument air compressors Residual heat removal pumps Standby liquid control pumps Service water pumps The team noted that the action levels for availability in the maintenance rule monitoring program were greater than the assumptions in the fire probabilistic risk assessment. With this observation, the team questioned the licensee as to whether this met the requirement in NFPA 805 to maintain the assumptions in the engineering analysis. The licensee informed the team that they had performed a sensitivity analysis to determine the significance of monitoring at a higher level of unavailability via the maintenance rule. This analysis determined an increase in core damage frequency for the additional unavailability time that could be accrued above the assumption for availability in the fire probabilistic risk assessment and up to the maintenance rule monitoring value for unavailability. This increase in core damage frequency was then determined to be acceptable if it did not exceed 1.0E-6/year. The team noted that for an individual component this screening criterion would not exceed more than 2 percent of the licensees baseline fire core damage frequency. The team was aware that some particular aspects of the monitoring program were being discussed between the industry and the NRCs Office of Nuclear Reactor Regulation during periodic public meetings which discussed Frequently Asked Question 10-0059, NFPA 805 Monitoring. The monitoring program and the sensitivity analysis approach used by the licensee are enveloped in these discussions. The team determined that additional information is required to determine if a performance deficiency exists. Specifically, the team needed to determine if the licensees action to set the action levels for the availability of some plant components at the components maintenance rule monitoring values and the performance of a riskinformed sensitivity analysis in an attempt to ensure that the assumptions in the engineering analysis remained valid would be an acceptable approach. Judgment on the suitability of this approach is pending further resolution of the monitoring program during discussions of Frequently Asked Question 10-0059, NFPA 805 Monitoring. The licensee entered this issue of concern into the corrective action program as Condition Report CR-CNS-2016-05109. This issue of concern is being treated as Unresolved Item 05000298/2016008-01, Possible Failure to Ensure that the Assumptions in the Engineering Analysis Remain Valid.
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