Semantic search

Jump to navigation Jump to search
 QuarterSiteTitleDescription
05000338/FIN-2018003-012018Q3North AnnaLicensee-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 Enforcement Policy. Violation: TS 5.4.1.a, requires in part, that written procedures shall be established per Revision 2 of Regulatory Guide 1.33, Appendix A, of which part 9.a requires written procedures and documented instructions appropriate to the circumstances for performing maintenance that can affect the performance of safety related equipment. Contrary to the above, on June 12, 2018, the licensee failed to adequately establish a procedure appropriate to the circumstances during maintenance on the safety-related main control chillers. Specifically, licensee mechanical preventative maintenance procedure, 0-MPM-0806-02, Inspection of Control Room Chillers, Revision 0, did not provide a proper method to adequately monitor the Freon level in main control room chillers. Consequently, the licensee discovered a low Freon level condition on main control room chiller 1-HV-3-4B, which rendered the chiller inoperable. Significance: The inspectors reviewed Exhibit 2 Mitigating Systems Screening Questions of IMC 0609 Appendix A, The Significance Determination Process (SDP) for findings at Power and determined this finding was of very low safety significance, Green, because there was no design deficiency, it did not represent a loss of system or function, and did not represent an actual loss of function for greater than its TS allowed outage time. Corrective Action Reference: CR109958
05000348/FIN-2018014-022018Q2FarleyFailure to Provide Complete and Accurate Information Related to System Operator RoundsDuring an NRC investigation completed on November 16, 2017, a SL IV NOV of 10 CFR 50.9, Completeness and Accuracy of Information, was identified when system operators failed to provide complete and accurate information related to system operator rounds. Specifically, on multiple occasions occurring from July 2016 through September 2016, information required by regulations to be maintained by the licensee was not complete and accurate in all material respects. Four SOs failed to comply with the procedural requirements of NMP-OS-007-001, Conduct of Operations Standards and Expectations, and FNP-0-SOP-0.11, Watch Station Tours and Operator Logs, in that on multiple occasions the SOs recorded data for certain readings without ever entering the corresponding area.
05000348/FIN-2018014-012018Q2FarleyFailure to Complete System Operator Rounds as Required per ProceduresDuring an NRC investigation completed on November 16, 2017, a SL IV Notice of Violation (NOV) of plant Technical Specification (TS) 5.4.1.a was identified when system operators failed to complete rounds as required per procedures. Specifically, on multiple occasions occurring from July 2016 through September 2016, four system operators (SOs) failed to complete various rounds as prescribed by documented instructions and procedures.Specifically, card reader data showed that the four SOs did not enter the rooms to record operating logs during their watch station rounds in accordance with the approved schedule, as required by NMP-OS-007-001, Conduct of Operations Standards and Expectations, and FNP-0-SOP-0.11, Watch Station Tours and Operator Logs.
05000390/FIN-2018012-012018Q2Watts BarPotential Failure to Implement Reviews of Adverse Employment Actions in Accordance with Confirmatory Order, EA-09-009 and EA-09-203Confirmatory Order EA-09-009, -203 was issued to TVA on December 22, 2009 and requires the following: By no later than ninety (90) calendar days after the issuance of this Confirmatory Order, TVA shall implement a process to review proposed licensee adverse employment actions at TVAs nuclear plant sites before actions are taken to determine whether the proposed action comports with employee protection regulations, and whether the proposed actions couldnegatively impact the SCWE. Such a process should consider actions to mitigate a potential chilling effect if the employment action, despite its legitimacy, could be perceived as retaliatory by the workforce...... The inspectors reviewed TVA procedure NPG-SPP-01.7.4, Adverse Employment Action and the Executive Review Board, Revision 0001, dated December 17, 2017. This procedure does not, in some instances, require TVA to review proposed licensee adverse actions before actions are taken to determine whether the proposed action could negatively impact the SCWE. TVA is required per this procedure to conduct an ERB and review the proposed adverse actions for those actions listed under the ERB Adverse Actions (TVA Employees Only) category. TVA is not required per this procedure to conduct an ERB or perform a SCWE impact review for those proposed adverse actions listed under the Non-ERB Adverse Actions category. In addition, three of the adverse actions listed in the Non-ERB Adverse Actions category (Demotion, Transfer to a Less Desirable Job, and Denial of Access) were previously included in TVAs adverse action process, but excluded from the required ERB review in the latest revision and are therefore no longer required to be reviewed using their current procedure. Planned Closure Action(s): Further inspection is needed in order to determine if the licensee is in compliance with Confirmatory Orders EA-09-009, -203 and EA-17-022. Specifically, inspectors need to review samples of adverse employment actions taken and the licensees review of those actions to determine if the licensee is adequately implementing their reviews in accordance with their procedures and the confirmatory order.
05000259/FIN-2018001-032018Q1Browns FerryFailure to Implement Controls for Locked High Radiation Area (LHRA) AccessA self-revealing, Green, NCVof TS 5.7.2, was identified for the failure to control access to a LHRA. Specifically, a worker installed and climbed a ladder in the Unit 3 drywell without Radiological Personnel (RP) present. In doing so, the worker accessed an area with dose rates >1 rem/hr that had not been posted, locked, or surveyed prior to entry
05000259/FIN-2018001-022018Q1Browns FerryUnauthorized Entry into a High Radiation Area(HRAA self-revealing, Green, NCVof Technical Specification (TS)5.7.1, was identified for a worker who entered a HRA without proper authorization. Specifically, the worker entered the Unit 3 A & C Residual Heat Removal Heat Exchanger Room using an incorrect Radiation Work Permit and without being briefed on the radiological conditions.
05000259/FIN-2018001-012018Q1Browns FerryInadequate Post-Maintenance Testing of 4kV Breaker Stationary SwitchesA self-revealing,Green, NCV of 10 CFR Part 50 Appendix B, Criterion V,was identified when the licensee failed to perform an adequate post-maintenance test in accordance with NPG-SPP-06.3, Pre-/Post-Maintenance Testing. Specifically, the post maintenance testing on the 3C diesel generator output breaker did not ensure that all contacts on replacement stationary switches successfully changed state after installation.
05000296/FIN-2018001-042018Q1Browns FerryInadequate Configuration Control of High Pressure Coolant Injection (HPCI)ValveDesign IssuesA self-revealing, Green, NCV of 10 CFR Part 50, Appendix B, Criterion III,was identified when the licensee failed to ensure adequate control of valve design configurations in accordance with NPG-SPP-9.3, Plant Modifications and Engineering Change Control Revision 6. Specifically, the licensee changed, over time, HPCI discharge valve yoke nut and bearing components contrary to original design without documenting or evaluating the changes
05000366/FIN-2017004-012017Q4HatchContinuous Fire Watch or Compensatory Measures Not Established per FHAAn NRC-identified non-cited violation (NCV) of Unit 2 License condition 2.C.(3)(a) Fire Protection was identified when on October 17, 2017, the licensee failed to establish a continuous fire watch or alternative compensatory measures required by Hatchs Fire Hazards Analysis (FHA), Appendix B, while the carbon dioxide fire protection system was nonfunctional during a routine maintenance outage for the 2C emergency diesel generator. Failure to establish a continuous fire watch or alternative compensatory actions as required by Hatchs Fire Hazards Analysis, Appendix B, when the low pressure carbon dioxide storage system became inoperable on October 17, 2017, was a performance deficiency. The licensee restored compliance on October 25, 2017, when the double fire door was shut, restoring functionality of the carbon dioxide system. The licensee entered this issue into the corrective action program as Condition Report (CR) 10423361.This performance deficiency was more-than-minor because the failure to establish a continuous fire watch or alternative compensatory measures adversely affected the reliability of the carbon dioxide system and/or compensatory measures. The finding screened to green because the alternate train of safe shutdown remained operable. The inspectors determined this performance deficiency had a cross cutting aspect in the Human Performance Area Training attribute because of the observed weakness in the application of FHA applicability statements. (H.9)
05000321/FIN-2017004-022017Q4HatchLack of Requirement for Engineering Evaluation of Scaffolding Near Safety-related PipingAn NRC-identified Green NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified when the licensee failed to ensure engineering evaluations were performed when scaffolding was constructed within 2 inches of safety-related piping. The failure to ensure procedure NMP-MA-010, Erecting, Modifying, and Disassembling Scaffolding, required engineering evaluations when scaffolding was constructed within 2 inches of safety-related piping was a performance deficiency. The violation was entered into the licensees corrective action program as CR 10420643.The performance deficiency was more-than-minor because the licensees procedure, as written, would never require an engineering evaluation of any safety-related piping based on the exceptions granted in the procedure. The inspectors determined that the finding was of very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containment. The inspectors determined that the finding did not have an associated cross-cutting aspect because the discrepancy was introduced during a transition to a fleet standardized procedure, which occurred more than three years ago and was therefore not reflective of current licensee performance.
05000321/FIN-2017004-032017Q4HatchA violation of Technical Specification (TS) 3.4.3 was identified because two of eleven safety relief valves were found to be outside the tolerance allowed by TS Surveillance Requirement (SR) 3.4.3.1 for the opening set-point pressure.Description: During the February 2017 Unit 2 refueling outage, all eleven 3-stage safety relief valves (SRVs) were removed and replaced. The SRVs were Target Rock model 0867F, a 3-stage valve design which was in its first use on Unit 2. This design was adopted as a corrective action to address corrosion bonding experienced by 2-stage SRV model 7687F valves which were previously in use at Hatch. "As-found" testing results indicated two of the eleven SRVs had experienced a setpoint drift during the previous operating cycle which resulted in their failure to meet the Technical Specification (TS) opening setpoint pressure as required by TS Surveillance Requirement (SR) 3.4.3.1. The SRV pilot valves were disassembled and inspected to determine the reason for the drift. The licensee determined that the abutment gap closed pre-maturely most likely due to loose manufacturing tolerances. For the 3-stage design, the pilot disc seating stresses should increase proportionally as reactor pressure increases to where a mechanical gap within the valve stem mechanism, referred to as the abutment gap, is closed. Additional pressure increases will cause the valve stem mechanism to reduce the disc seat pressure until the valve eventually opens. This same cause was previously identified in 2016 (CAR 264544) after two of eleven SRVs removed from Unit 1 also experienced setpoint drift. Because the Unit 2 valves were already installed when the cause was initially identified, there was no opportunity for the licensee to take corrective actions for the valves that are the subject of this LER. Additionally, there were no symptoms available to operators or maintenance personnel to indicate the potential for the set point drift prior to post-service testing. As a corrective action, when the eleven valves were removed for post-service testing, the licensee installed eleven refurbished pilot valves that underwent the corrective actions identified by CAR 264544 which included the vendors usage of revised tolerances.Enforcement: Hatch Unit 2 TS limiting condition for operation 3.4.3, Safety/Relief Valves, required 10 of 11 SRVs be operable in MODES 1, 2 and 3. With two or more SRVs inoperable, the required TS action must be taken by the applicable completion time. Contrary to the above, Unit 2 operated from the initiation of the degraded condition until February 6, 2017, with two SRVs inoperable. The inspectors concluded that the violation would normally be characterized as a Severity Level IV violation because it was of very low safety significance (Green). However, the NRC is exercising enforcement discretion (EA-18-006) in accordance with Section 3.10 of the Enforcement Policy because the violation was not associated with a licensee performance deficiency. This issue was documented in the licensees corrective action program as CR 10382586.
05000395/FIN-2017003-032017Q3SummerFailure to Adequately Assess the Risk for an Activity with Consequent Loss of Core CoolingThe inspectors identified a Green, NCV of 10 CFR 50.65(a)( 4), Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, involving the licensees failure to perform an adequate risk assessment for an activity involving restoration of the B train emergency bus to the normal supply and a subsequent loss of the B train residual heat removal (RHR) pump and a consequent loss of core cooling. The issue was entered into the licensees CAP as condition report, CR- 17- 03696 The inspectors reviewed IMC 0612, Appendix B, Issue Screening, dated September 7, 2012, and determined that the PD was more than minor and therefore a finding because it was associated with the Mitigating Systems Cornerstone, and adversely affected the cornerstone objective to ensure in part the availability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to perform an adequate risk assessment resulted in performance of an activity causing a vulnerability to operability of the running RHR pump and a consequent loss of core cooling. The finding was screened for risk significance using NRC IMC 0609.04 and routed to NRC IMC 0609, Appendix G. A detailed shutdown risk assessment was performed by a regional senior risk analyst using NRC IMC 0609 Appendix G and Attachments 1 and 2. The major analysis assumptions included: treatment of the PD as a loss of RHR with an initiating event likelihood of 1.0 using NRC IMC 0609 Appendix G, Attachment 2, worksheet 9 (loss of RHR in plant operating state 2); and recovery credit was applied for closing the alternate feeder breaker. The dominant sequence was a loss of RHR, failure to recover decay heat removal prior to reactor coolant system (RCS) boiling, failure to initiate RCS injection before core damage and failure to restore power to the B train safety bus. The RCS conditions of time to boil and time to core uncovery and availability of mitigating equipment limited the risk. The detailed risk evaluation determined that the PD represented an increase in core damage frequency <1.0E -6 a GREEN finding of very low safety significance. The inspectors reviewed IMC 0310, Aspects Within Cross Cutting Areas, dated December 4, 2014, and determined that this finding had a cross -cutting aspect in the area of Work Management (H.5), because the licensee did not perform an adequate risk assessment in accordance with their procedure.
05000395/FIN-2017003-022017Q3SummerViolation of NRC Examination Security as Required by 10 CFR 55.49The inspectors identified a Green, NCV of 10 CFR 55.49, Integrity of examinations and tests, for the licensees failure to ensure the paper strip chart recorders were advanced following the completion of Job Performance Measures (JPMs) and simulator scenarios to prevent examination compromise when the same examination was administered to multiple crews on the same day. While observing simulator JPMs, the inspectors identified that simulator staff was not advancing any of the strip chart recorders after each JPM. This left examination material out and visible to the next operator performing the JPM. The inspectors informed the licensee of this issue and they immediately started advancing each chart recorder until no exam 3 related material was visible. An initial review by the licensee indicated that not advancing the paper had been a standard practice for this exam period. In accordance with the NRC Enforcement Policy, this violation was classified as Severity Level IV Violation (Section 6.4.d). This Severity Level IV violation is being treated as a Non -Cited Violation, consistent with Section 2.3.2.a of the Enforcement Policy. This violation is in the licensees corrective action program under CR 17 -04424
05000395/FIN-2017003-012017Q3SummerInadequate Procedures for Inspection of Fire BarriersThe inspectors identified a Green, NCV of Operating License Condition 2.C.(18), Fire Protection Program, for failure to adequately establish a surveillance procedure for fire penetration inspections. The licensee entered the problem into their corrective action program (CAP) as condition report, CR -17-04029. The inspectors used IMC 0612, Appendix B, Issue Screening, dated September 7, 2012, and determined that the performance deficiency ( PD ) was more than minor and therefore a finding because it was associated with the Mitigating System Cornerstone and the respective attribute of protection against external factors (fire). This finding had a credible impact on safety and adversely affected the cornerstone objective because STP -728.043 failed to specify adequate inspection of penetrations added by modification of which one penetration (5045) was degraded. The inspectors used IMC 0609, Appendix F, Fire Protection Significance Determination Process, dated September 20, 2013, and determined that the finding impacted the fire confinement finding category. Based on review of respective Attachment 2, Degradation Rating Guidance Specific to Various Fire Protection Program Elements, dated February 28, 2005, the inspectors determined that the degradation rating was low based on the size of the degradation identified. Specifically, the separation of foam seal material was approximately 3 from the mating conduit surfaces and there was greater than 12 of foam seal material when it was initially installed. As a result, this finding was determined to be of very low safety significance (Green) based on the guidance in IMC 0609, Appendix F, Attachment 1, Fire Protection Significance Determination Process Worksheet, dated September 20, 2013
05000395/FIN-2017002-022017Q2SummerFailure to Provide NRC Staff Complete and Accurate InformationThe inspectors identified a severity level (SL) IV NCV of 10 CFR 50.9(a), Completeness and accuracy of information, involving licensee document,RC-13-0142, dated October 14, 2013. This document was a response to a request for additional information involving a license amendment request (LAR) to adopt NFPA 805 and contained an approval request, L12, associated with oil misting from the reactor coolant pumps. The licensee entered this violation into their corrective action program as CR-17-03961. The inspectors determined that the licensees failure to provide complete and accurate information associated with approval request, L12, was a violation of 10 CFR 50.9(a). Because this violation of 10 CFR 50.9(a) impacted the NRCs ability to perform its regulatory function, the inspectors evaluated this violation using traditional enforcement (TE). Since the TE violation is associated with a previous Green reactor oversight process violation, and the misinformation was identified after the NRC relied on it for issuing a previous operating license amendment, the TE violation was determined to be a SL IV, NCV, consistent with the language of the NRC Enforcement Policy, Section 2.3.11, Inaccurate and Incomplete Information. This violation involved TE; therefore a cross-cutting aspect was not assigned.
05000395/FIN-2017002-012017Q2SummerFailure to Implement Corrective Actions to Restore Compliance for Previous NRC-identified Green NCV 05000395/2013003-03The inspectors identified a Green finding with a cited violation of Operating Licensee Condition 2.C.(18) for failure to ensure that conditions adverse to fire protection as noted in a previous NRC-identified Green NCV, 05000395/2013003-03, Failure to Adequately Design, Install and Maintain Oil Collection Devices for Reactor Coolant Pump Motors, were corrected. Specifically, the licensee failed to implement corrective actions and restore compliance in a timely manner for (1) a failure to ensure an adequate design for the oil lift pump enclosure, and (2) a failure to have oil collection components for internally leaked oil dripping from the motor air discharge ductwork flange. The licensee entered the issue in their corrective action program as condition report CR-17-03962.The inspectors determined that the failure to implement corrective actions for the oil collection system to restore compliance was a performance deficiency (PD). The inspectors used IMC 0612 and determined that the PD was more than minor and therefore a finding because it impacted the mitigating systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences, and the related attribute of protection against external factors such as fire. This finding has a credible impact on safety because the failure to adequately install, maintain and design the oil collection system presented a degradation of a fire confinement component which has a fire prevention function of not allowing an oil leak to reach hot surfaces. This finding had been evaluated and screened to a low safety significance (Green) and documented in the previous NRC-identified Green NCV, 05000395/2013003-03. Because the licensee failed to implement corrective actions and restore compliance in a timely manner, this violation is being treated as a cited violation, consistent with Section 2.3.3 of the NRC Enforcement Policy. The inspectors used IMC 0310 and determined this finding has a cross-cutting aspect in the area of Problem Identification and Resolution because the organization failed to take effective corrective actions to address issues in a timely manner commensurate with their safety significance and restore compliance (P.3).
05000338/FIN-2012007-052012Q3North AnnaInadequate Procedures and Procedure Compliance For Thermal Overload Relay TestingThe team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, involving two examples. In the first example, the licensee failed to ensure that appropriate acceptance criteria was included in procedures for testing motor control center thermal overload relays. In the second example, the licensee failed to ensure that testing was accomplished in accordance with the procedures. The licensee entered these issues into their corrective action program as condition reports 479217, 479281, 479535, 479552, and 480755. The licensees failure to ensure that appropriate criteria was included in procedures for testing motor control center thermal overload relays, and the failure to ensure that testing was accomplished in accordance with the procedures was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the Procedure Quality attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, there was reasonable doubt as to whether safety related motors would continue to operate without tripping during design basis conditions. In accordance with Nuclear Regulatory Commission Inspection Manual Chapter 0609.04, Initial Screening and Characterization of Findings, the team conducted a Phase 1 Significance Determination Process screening and determined the finding to be of very low safety significance (Green) because it was not a design deficiency, did not represent the loss of a system safety function, did not result in exceeding a TS allowed outage time, and did not screen as potentially risksignificant due to a seismic, flooding, or severe weather initiating event. The team identified a crosscutting aspect in the work practices component of the human performance area.
05000338/FIN-2012007-042012Q3North AnnaInadequate Design Control Measures for Thermal Overload RelaysThe team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to verify the adequacy of thermal overload relay settings for motor operated valves and continuous duty motors. The licensee entered this issue into their corrective action program as condition reports 479217, 479281, 479535, 479552, and 480755. The licensees failure to verify or check the adequacy of thermal overload relay settings for motor operated valves and continuous duty motors was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the Design Control 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, there was reasonable doubt as to whether safety related motors would continue to operate without tripping during design basis conditions. In accordance with Nuclear Regulatory Commission Inspection Manual Chapter 0609.04, Initial Screening and Characterization of Findings , the team conducted a Phase 1 Significance Determination Process screening and determined the finding to be of very low safety significance (Green) because it was a design deficiency confirmed not to have resulted in the loss of operability or functionality. The team identified a crosscutting aspect in the corrective action program component of the problem identification and resolution area.
05000338/FIN-2012007-032012Q3North AnnaInadequate Testing of the SW Air SystemThe team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for the licensees failure to test the Service Water (SW) air subsystem capability to perform its design bases function. Specifically, the licensee was not testing the air receiver inlet valves (1-SW-343 and 1-SW- 105), or system integrity to ensure the systems capability to maintain header pressure without crediting the non-safety related air compressors. The licensee entered this issue into their corrective action program as condition report 478568. The licensees failure to test the safety related SW air systems capability to maintain adequate header pressure when the SW air compressors are not available was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to perform testing of the SW air system resulted in a lack of reasonable assurance of the systems capability to maintain adequate header pressure and could have resulted in a premature or complete loss of the screen wash system. If the screen wash system was required to mitigate the effects of a severe weather initiating event, the performance deficiency could have resulted in a common mode failure of the SW system. In accordance with Nuclear Regulatory Commission Inspection Manual Chapter 0609.04, Initial Screening and Characterization of Findings, the team conducted a Phase 1 Significance Determination Process screening and determined that a Phase 3 assessment was required because the finding screened as potentially risk-significant due to a severe weather initiating event which could plug the SW travelling screens requiring the screen wash function. A bounding Significance Determination Process Phase 3 analysis was performed by a regional senior risk analyst which determined the performance deficiency was a Green finding of very low safety significance. The finding was reviewed for cross-cutting aspects and none were identified since the performance deficiency was not indicative of current licensee performance.
05000338/FIN-2012007-022012Q3North AnnaFailure To Implement Design Control Measures For The Service Water Air SystemThe team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to implement design control measures involving two examples. In the first example, the licensee failed to translate the updated final safety analyses report single failure design bases criteria into the service water (SW) air system specifications. In the second example, the licensee failed to verify the SW air system receiver capacity was adequate to support its design basis function. The licensee entered these issues into their corrective action program as condition reports 477213, 478531, 478957, and 478137. The licensees failure to establish design control measures to translate the updated final safety analyses report single failure design basis criteria into SW air system specifications and failure to verify or check the adequacy of the SW air receiver capacity was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the Design Control 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, if the screen wash system was required to mitigate the effects of a severe weather initiating event, the performance deficiency could have resulted in a common mode failure of the SW system. In accordance with NRC IMC 0609.04, Initial Screening and Characterization of Findings, the team conducted a Phase 1 Significance Determination Process screening and determined that a Phase 3 assessment was required because the finding screened as potentially risk-significant due to a severe weather initiating event which could plug the SW traveling screens requiring the screen wash function. A bounding Significance Determination Process Phase 3 analysis was performed by a regional senior risk analyst which determined the performance deficiency was a Green finding of very low safety significance. The finding was reviewed for cross-cutting aspects and none were identified since the performance deficiency was not indicative of current licensee performance.
05000338/FIN-2012007-012012Q3North AnnaFailure to Develop an Adequate Procedure to Test the Quench Spray and Outside Recirculation Spray Pump Discharge Check ValvesThe team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings for the licensees failure to develop an adequate test procedure which demonstrated that the quench spray and outside recirculation spray pumps discharge check valves were capable of performing their design basis function. The licensee entered this issue into their corrective action program as condition report 479661. The licensees failure to develop an adequate test procedure which demonstrated that the quench spray and outside recirculation spray pumps discharge check valves were capable of performing their design bases functions was a performance deficiency. This performance deficiency was more than minor because it was associated with the procedure quality attribute of the mitigating system cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems to respond to initiating events to prevent undesirable consequences. Specifically, the failure to measure the torque required to cycle the check valves and compare these with established limits could result in the failure to detect degraded valve performance and prevent it from performing as designed. In accordance with Nuclear Regulatory Commission Inspection Manual Chapter 0609.04, Initial Screening and Characterization of Findings , the team conducted a Phase 1 Significance Determination Process screening and determined the finding to be of very low safety significance (Green) because it was not a design deficiency, did not represent the loss of a system safety function, did not result in exceeding a Technical Specification allowed outage time, and did not screen as potentially risk-significant due to a seismic, flooding, or severe weather initiating event. The team identified a cross-cutting aspect in the decision making component of the human performance area.
Retrieved from "https://kanterella.com/Special:Ask/-5B-5BCategory:Finding-5D-5D-20-5B-5BInspector::Patrick-20Heher-5D-5D/-3FQuarter/-3FSite/-3FTitle/-3FDescription/mainlabel%3D/order%3Ddescending/sort%3Dstart-20date/offset%3D0/format%3Dtable/searchlabel%3D-5BFindings-20List-5D"