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| Quarter | Site | Title | Description | |
|---|---|---|---|---|
| 05000416/FIN-2015001-03 | 2015Q1 | Grand Gulf | Emergency Action Level Scheme for Nonfunctional Seismic Monitor | The inspectors identified a non-cited violation of 10 CFR 50.54(q)(2) for the licensees failure to follow and maintain the effectiveness of an emergency plan that meets the requirements of the planning standard 50.47(b)(4), which requires that a standard emergency classification and action level scheme, is in use by the licensee. Specifically, the licensee had identified, on October 15, 2013, that the seismic monitoring instrumentation was non-functional, but had not further evaluated the plant configuration, and the effect on emergency action level declaration capabilities for seismic events. The licensee documented this issue in Condition Report CR-GGN-2015-00713. The corrective actions, based on CR-GGN-2013-06514, were implemented, and a new seismic monitor was installed, tested, and brought into service on January 30, 2015. The licensees inability to promptly declare Emergency Action Level (EAL) HA6, as required in the approved emergency classification and action level scheme per 10 CFR Part 50.47(b)(4), was a performance deficiency. This performance deficiency is more than minor, and therefore a finding, because it is associated with the procedure quality attribute of the Emergency Preparedness Cornerstone and adversely affects 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. Specifically, it negatively impacts the cornerstone attribute of procedure quality in that the plant configuration prohibited the timely declaration of the facility EALs, as written. Using NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, dated June 19, 2012, the inspectors determined that the issue affected the Emergency Preparedness Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix B, Emergency Preparedness Significance Determination Process, dated September 23, 2014, the inspectors determined that the issue is of very low safety significance (Green) because an Emergency Action Level was rendered ineffective such that HA6 would not be declared, consistent with Table 5.4-1 and Figure 5.4-1. The inspectors determined the finding had a cross-cutting aspect in the area of problem identification and resolution associated with evaluation, in that the organization did not thoroughly evaluate issues to ensure that resolutions address causes, and extent of conditions, commensurate with their safety significance; in that while following Technical Requirements Manual requirements for a non-functional piece of equipment (seismic monitor), the complete effect was not evaluated to ensure the EALs were still capable of being implemented (P.2). |
| 05000416/FIN-2015001-04 | 2015Q1 | Grand Gulf | Failure to Properly Calibrate Main Steam Line Radiation Monitors and Containment/Drywall High Range Radiation Monitors | The inspectors identified a non-cited violation of 10 CFR 20.1501(c) for the licensees failure to properly calibrate the main steam line radiation monitors and the containment/drywell high range radiation monitors. The violation was of very low safety significance and was entered into the licensees corrective action program as Condition Report CR-GGNS-2015-01832. The failure to properly calibrate radiation monitors was a performance deficiency. The performance deficiency is more than minor, and therefore a finding, because it adversely affects the cornerstone objective to ensure adequate protection of employee health and safety and is associated with the cornerstone attribute of plant instrumentation. Specifically, the failure to properly calibrate radiation monitors impacts their ability to be used to assess dose rates. Using Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008, the inspectors determined the finding to be of very low safety significance because it was not an as low as reasonably achievable (ALARA) issue, there was no overexposure or substantial potential for overexposure, and the licensees ability to assess dose was not compromised. This finding has a cross-cutting aspect in the resources component of the human performance area because the licensee did not ensure that calibration procedures were adequate, nor was proper calibration equipment designed, characterized, and made available (H.1). |
| 05000416/FIN-2015001-05 | 2015Q1 | Grand Gulf | Failure to Establish, Implement, and Maintain Appropriate Changes to the Offsite Dose Calculation Manual For REMP Airborne Sampling | The inspectors identified a non-cited violation of Technical Specification 5.5.1, Offsite Dose Calculation Manual (ODCM). Specifically, when changes were made to the Offsite Dose Calculation Manual in 1997, the licensee failed to establish an airborne sampling location for a community with the highest deposition factor (D/Q) for the site. As immediate corrective actions, the licensee evaluated their Offsite Dose Calculation Manual, evaluated the dose differential for the monitoring locations, and developed a plan to meet the environmental sampling requirements. The issue was documented in Condition Report CR-GGNS-2015-01835. The failure to establish an air sampling location in the vicinity of a community having the highest D/Q was a performance deficiency. The performance deficiency is more than minor, and therefore a finding, because it adversely affects the cornerstone objective to ensure adequate protection of public health and safety from exposure to radioactive materials released into the environment and public domain. Specifically, the failure to maintain the Offsite Dose Calculation Manual with appropriate airborne radionuclide sampling requirements adversely impacts the licensee's ability to validate offsite radiation dose assessments for members of the public under certain effluent release conditions. Using Inspection Manual Chapter 0609, Appendix D, dated February 12, 2008, Public Radiation Safety Significance Determination Process, the inspectors determined that the violation had very low safety significance because it involved the environmental monitoring program. This finding has a cross-cutting aspect in the procedure adherence component of the human performance area because licensee personnel failed to follow procedures when they determined the airborne sampling locations for the updated Radiological Environmental Monitoring Program (H.8). |
| 05000416/FIN-2015001-06 | 2015Q1 | Grand Gulf | Failure to Adequately Establish Commercial-Grade Items as Basic Components | The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to verify the suitability of replacement parts that were procured from commercial suppliers. Specifically, the inspectors noted that none of the tests specified by the licensee were sufficient to ensure that the seismic qualification of an auxiliary relay had been maintained. The finding was entered into the licensees corrective action system as Condition Report CR-GGN-2014-05049. The performance deficiency is more than minor, and therefore a finding, 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. In accordance with IMC 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 the licensee performed an operability determination, which evaluated the safety impacts of postulated relay chatter during a seismic event, for the applications in which these relays were installed. The licensees subsequent operability evaluation determined that potential relay chatter would not impact the safety-related functions of the relays in the applications in which they were installed. Thus, all applicable screening questions in Manual Chapter 0609, Appendix A, Exhibit 2, were answered no. A cross-cutting aspect is not being assigned to this finding. |
| 05000416/FIN-2015003-01 | 2015Q3 | Grand Gulf | Failure to Have Appropriate Instructions for Preventative Maintenance on the Division II Diesel Generator Fuel Rack Control Lever | The inspectors reviewed a self-revealing, non-cited violation of Technical Specification 5.4.1.a, for the failure to establish appropriate maintenance instructions to perform maintenance activities on the fuel rack control lever of the division II diesel generator. Specifically, the preventative maintenance instruction did not inspect for foreign material between the fuel rack control lever and the adjacent clamp, which caused the fuel rack control lever to be stuck in the open position. As a result, the division II diesel generator was rendered inoperable and unavailable. On June 28, 2015, the licensee cleaned and lubricated the fuel rack control lever and performed the preventative maintenance instruction to return the division II diesel generator to operable status. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2015-3741. This 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 adversely 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 A, The Significance Determination Process (SDP) for Findings At-Power, and Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that the finding is of very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and (4) did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety significant in accordance with the licensees maintenance rule program. The mechanical standard was last updated in 2006, and the preventative maintenance instruction was last updated in 2012 for editorial purposes only. The inspectors determined that the cause of the deficiency occurred in 2006, and therefore, determined the finding did. not have a cross-cutting aspect since it is not indicative of current licensee performance. |
| 05000416/FIN-2017011-01 | 2018Q1 | Grand Gulf | Failure to Categorize Condition Reports for Significant Conditions Adverse to Quality as Required by Procedures | The inspectors identified five examples of a finding for the licensees failure to categorize and evaluate conditions in accordance with procedural requirements. Specifically, the licensee did not categorize adverse conditions that represented the loss of a safety function as significant conditions adverse to quality as required by Procedure EN-LI-102, Corrective Action Program, Revisions 24 through 28. The licensee entered the conditions into their corrective action program as Condition Report CR-GGN-2017-10896. The licensee initiated corrective actions to re-categorize the conditions and perform the required evaluations. The failure to categorize conditions that represent the loss of a safety function as significant conditions adverse to quality as required by Procedure EN-LI-102 was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because if left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, root cause evaluations, corrective actions to prevent recurrence, and effectiveness reviews are used per licensee Procedure EN-LI-102 to ensure availability and reliability of structures, systems, and components are maintained. Using Nuclear Regulatory Commission Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because it was related to, but was not itself: (1) a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) a loss of system and/or function; (3) an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and (4) an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees Maintenance Rule program. This finding had a cross-cutting aspect in the area of human performance, consistent process, because the licensee did not use a consistent, systematic approach to make decisions. Specifically, the licensees failure to consistently evaluate the conditions during initial screening led to the incorrect categorization of the condition reports (H.13) |
| 05000416/FIN-2017011-02 | 2018Q1 | Grand Gulf | Failure to Disposition Adverse Conditions as Required by Procedures | The inspectors identified a finding for the licensees failure to disposition conditions as required by Procedure EN-LI-102, Corrective Action Program, Revisions 24 through 30. Specifically, the licensee did not identify 72 conditions as either Adverse Category B, C, or D as required by the procedure. As a result, the licensee failed to perform the required cause evaluations and develop corrective actions to address the conditions. The licensee entered the conditions into their corrective action program as Condition Report CR-GGN-2017-10896. The licensee initiated corrective actions to re-categorize the conditions and perform the required evaluations. The failure to disposition conditions as adverse (B, C, or D) as required by Procedure EN-LI-102 was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because if left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, condition reports associated with deficiencies or potential deficiencies involving safety-related equipment are required to be categorized as adverse and appropriate corrective actions are assigned including causal analyses appropriate to the circumstances per licensee Procedure EN-LI-102. The inspectors performed an initial screening of the finding in accordance with Nuclear Regulatory Commission Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using Nuclear Regulatory Commission Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because it was related to, but was not itself: (1) a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) a loss of system and/or function; (3) an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and (4) an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees Maintenance Rule program. This finding had a cross-cutting aspect in the area of human performance, consistent process, because the licensee did not use a consistent, systematic approach to make decisions. Specifically, the licensees failure to consistently disposition identified conditions as adverse led to the failure to fully evaluate the conditions (H.13). |
| 05000416/FIN-2017011-03 | 2018Q1 | Grand Gulf | Failure to Conduct Common Cause Failure Evaluation in Response to Inoperable Emergency Diesel Generator | The inspectors identified three instances of a non-cited violation of Technical Specification 3.8.1, AC Sources Operating, for the licensees failure to take required actions for an inoperable emergency diesel generator. Specifically, after classifying the Division I or Division II emergency diesel generator as inoperable on the basis of nonconforming conditions, and after failing to either verify that the opposite train emergency diesel generator was not inoperable due to common cause failure within 24 hours or conduct a surveillance run on the opposite train emergency diesel generator within 24 hours, the licensee failed to enter Mode 3 within 12 hours as required by Technical Specification 3.8.1, Actions B.3.1, B.3.2, and G.1, respectively. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2017-11393. The licensee initiated corrective actions to conduct an adverse condition analysis. The failure to take required actions for an inoperable emergency diesel generator was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the equipment reliability 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. Actions B.3.1 and B.3.2 of Technical Specification 3.8.1 exist to ensure the availability, reliability, and capability of at least one emergency diesel generator in scenarios where there is a potential for a common cause failure of both emergency diesel generators, and the licensee took neither action. Using Nuclear Regulatory Commission Inspection Manual Chapter 0609, Appendix A, Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because the finding did not represent an actual loss of function of either the Division I or Division II emergency diesel generator for greater than its technical specifications allowed outage time. The finding had a cross-cutting aspect in the area of human performance, consistent process, because the licensee failed to use a consistent, systematic approach to make decisions. Specifically, the licensee failed to review the required actions of the applicable technical specification to ensure that all of those actions would be properly carried out (H.13). |
| 05000416/FIN-2017011-04 | 2018Q1 | Grand Gulf | Failure to Install the Residual Heat Removal Pump A Mechanical Seal in Accordance with Procedures | The inspectors reviewed a self-revealed, non-cited violation of Technical Specification 5.4, Procedures, for the licensees failure to perform maintenance on the residual heat removal pump A mechanical seal in accordance with written procedures. Specifically, on September 22, 2016, maintenance did not install seal assembly drive pins in accordance with Step 7.8.2 of Procedure 07-S-14-279, Revision 007. The licensee entered this issue into their corrective action program as Condition Reports CR-GGN-2017-08269 and CR-GGN-2017-11009. The licensee implemented immediate corrective actions by declaring the pump inoperable and replacing the mechanical seal. The failure to perform maintenance on the residual heat removal pump A mechanical seal in accordance with written procedures was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the human 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, on September 22, 2016, mechanical maintenance installed the residual heat removal pump A seal drive pins backwards. As a result, the drive pins damaged the seal and on August 23, 2017, caused an unisolable leak from the seal. This resulted in unplanned inoperability and unavailability of the residual heat removal pump A from August 23, 2017, through August 25, 2017, when the mechanical seal was replaced. Using Nuclear Regulatory Commission Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, and Nuclear Regulatory Commission Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because it was related to, but was not itself: (1) a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) a loss of system and/or function; (3) an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and (4) an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees Maintenance Rule program. This finding had a cross-cutting aspect in the area of human performance, avoid complacency, because the licensee failed to recognize and plan for the possibility of mistakes, and individuals failed to implement appropriate error reduction tools. Specifically, the licensee failed to use appropriate error reductions tools such as self-check or peer checking which resulted in incorrect performance of procedural steps (H.12) |
| 05000416/FIN-2017011-05 | 2018Q1 | Grand Gulf | Failure to Correct Control Room Boundary Door Resulted in Loss of Safety Function | The inspectors reviewed a self-revealed, non-cited violation of 10 CFR Part 50, Criterion XVI, Corrective Action, for the licensees failure to appropriately correct a condition adverse to quality. Specifically, the control room envelope door had been documented in several condition reports for not consistently working properly. Subsequent to these condition reports, on July 9, 2017, the door was opened and did not close automatically, and therefore the door was left in an unsecured position. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2017-06705. The licensee restored compliance by securing the door and replacing the hinge bushings to ensure the door would close properly. The failure to correct a condition adverse to quality for a control room envelope boundary door was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the structures, systems, and components and barrier performance attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (functionality of the control room) protect the public from radionuclide releases caused by accidents or events. Specifically, on July 9, 2017, since the licensee had not corrected the adverse conditions identified on the control room envelope door, the door was left in an unsecured position and resulted in the station declaring both trains of standby fresh air inoperable. Using Nuclear Regulatory Commission Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, and Nuclear Regulatory Commission Inspection Manual Chapter 0609, Appendix A, Exhibit 3, Barrier Integrity Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because the finding did not represent a degradation of the radiological barrier function provided for the control room, auxiliary building, spent fuel pool, or standby gas treatment system, and did not represent a degradation of the barrier function of the control room against smoke or a toxic atmosphere. The period of the barrier in the open position was of short duration, approximately 1 minute, and therefore did not result in significant risk input. This finding had a cross-cutting aspect in the area of problem identification and resolution, resolution, because the licensee did not take corrective actions in a timely manner commensurate with their safety significance. Specifically, the licensee did not ensure proper priority of corrective actions on the degraded control room envelope boundary door (P.3). |
| 05000416/FIN-2017011-06 | 2018Q1 | Grand Gulf | Failure to Perform Functionality Assessments as Required by Procedures | The inspectors identified a finding for the licensees failure to follow Procedure EN-OP-104, Operability Determination Process, Revisions 10 through 12. Specifically, the licensee did not perform functionality assessments for adverse conditions on the offgas system as required by the procedure. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2017-11265. The licensee initiated corrective actions to perform functionality assessments for the conditions identified and to evaluate any potential programmatic issues. The failure to perform functionality assessments required by Procedure EN-OP-104 was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because if left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, the failure to perform functionality assessments could affect the availability and reliability of the offgas system to maintain the doses associated with releases to the environment as low as reasonably achievable. Using Nuclear Regulatory Commission Inspection Manual Chapter 0609, Appendix D, Public Radiation Safety Significance Determination Process, the inspectors determined that the finding was of very low safety significance (Green) because it involved the Effluent Release Program, it did not impair the ability to assess dose, and did not exceed the 10 CFR Part 50, Appendix I, or 10 CFR 20.1301(d) limits. This finding had a cross-cutting aspect in the area of human performance, consistent process, because the licensee did not use a consistent, systematic approach to make decisions. Specifically, the licensees failure to consistently disposition adverse conditions associated with the offgas system resulted in the station not performing required functionality assessments (H.13) |
| 05000416/FIN-2018002-01 | 2018Q2 | Grand Gulf | Failure to Institute Effective Corrective Action to Preclude Repetition | An NRC-identified,Green non-cited violation of 10CFRPart50, AppendixB, CriterionXVI, Corrective Action, was identified when the licensee failed to institute effective corrective actions to preclude repetition of a significant condition adverse to quality. Specifically, the licensee left a secondary containment personnel hatch in an open configuration for approximately 30 minutes while performing a roof inspection, which rendered secondary containment inoperable. This issue had also previously occurred in 2016, but corrective actions to prevent it from occurring again were ineffective. |
| 05000416/FIN-2018002-02 | 2018Q2 | Grand Gulf | Failure to Follow ASME Requirements for Maintaining Inservice Inspection (ISI) Cycles and Perform ASME Required Inservice Inspections within the Scheduled ISI Cycle | The inspector identified 15 examples of a Green non-cited violation (NCV)of 10 CFR 50.55(a)(g)(4)(ii), which requires that inservice examination of components classified as American Society of Mechanical Engineers (ASME), Section XI, Code Class 1, Class 2, and Class 3 be conducted during successive 120-month inspection intervals, and requires compliance with the requirements of the latest edition and addenda of the ASME Code (and all its paragraphs) applicable to the specific interval, including maintaining the 120-month inspection interval in accordance with the ASME Code, Section XI, Paragraph IWA-2430. Specifically, the licensee inappropriately adjusted its second inservice inspection 120-month cycle, and failed to perform VT-3 and MT examinations of 15 class 1, class 2, and class 3 components, including the high pressure core spray pump attachment weld and reinforcing band before the third inservice inspection cycle expired on November 30, 2017, as required by 10CFR50.55(a)(g)(4)(ii). |
| 05000416/FIN-2018002-03 | 2018Q2 | Grand Gulf | Failure to Adequately Test NUS Temperature Switch | A self-revealed,Green non-cited violationof 10CFRPart50, AppendixB, CriterionIII, Design Control, was identified when the reactor core isolation cooling (RCIC) system automatically isolated due to an inadvertent high temperature input from the leakage detection system. Specifically, the licensee failed to fully test a modification that installed a new type of temperature switches, and the system inappropriately isolated the RCIC system when a loss and subsequent restoration of power occurred. |
| 05000416/FIN-2018002-04 | 2018Q2 | Grand Gulf | High Radiation Area Boundary Violation | A self-revealed, Green non-cited violation of Technical Specification 5.7.1 was identified when an individual received a dose rate alarm when the individual failed to comply with established radiological barriers and protective measures and entered a high radiation area. Specifically, an individual leaned over a high radiation area barricade rope, thereby entering the high radiation area. The individuals radiation work permit (RWP) did not permit entry into a high radiation area. |
| 05000416/FIN-2018002-05 | 2018Q2 | Grand Gulf | Failure to Follow Procedure Requirements Resulting in Unplanned Dose | A self-revealed, Green non-cited violation of Technical Specification 5.4.1 was identified when an individual alarmed a personnel contamination monitor upon exit from the radiologically controlled area. Specifically, the licensee failed to follow procedures to establish a decontamination plan or procedure, conduct a specific pre-job brief addressing appropriate contamination risk, and receive approval by radiation protection supervision prior to conducting decontamination activities on thereactor pressure vessel(RPV) O-rings |
| 05000416/FIN-2018002-06 | 2018Q2 | Grand Gulf | Improper Evaluation and Resolution of Intermediate Range MonitorNoise Leads to Manual Reactor Shutdown | A self-revealed, Green non-cited violation of 10CFRPart50, Appendix B, Criterion XVI, Corrective Action, was identified for the failure of the licensee to identify and correct a condition adverse to quality. Specifically, the licensee failed to implement appropriate corrective actions related to intermediate range monitor (IRM) nuclear instrument (NI) electronic noise spiking. The failure to implement adequate corrective actions over the course of at least 5 years resulted in a plant shutdown due to declaration of multiple IRM channels inoperable while in Mode 2. |
| 05000416/FIN-2018002-07 | 2018Q2 | Grand Gulf | Loss of Shutdown Cooling | A self-revealed,Green non-cited violation of Technical Specification 5.4, Procedures,for the licensees failure to follow written procedures was identified when the residual heat removal (RHR) system automatically isolated due to an inadvertent emergency core cooling system (ECCS) actuation. While the plant was shut down with the RHR system in decay heat removal mode, maintenance personnel inadvertently opened an incorrect valve during a transmitter calibration activity, which caused a false low reactor pressure vessel (RPV) water level signal, an ECCS actuation, and a loss of decay heat removal for approximately 31 minutes |
| 05000416/FIN-2018002-08 | 2018Q2 | Grand Gulf | Performance of Surveillance Testing Following Maintenance on Containment Airlock | The inspectors identified a Green non-cited violation of 10CFRPart50,AppendixB, Criterion XI, Test Control, for the licensees failure to perform surveillance testing of containment airlock seals under appropriate conditions. The licensee failed to appropriately control the sequence of maintenance and testing activities to ensure that surveillance testing was not performed subsequent to maintenance which could affect the validity of surveillance test results. |
| 05000458/FIN-2008004-01 | 2008Q3 | River Bend | Inadequate Procedure for Staging the Station Blackout Diesel Generator during Severe Weather | The inspectors identified a noncited violation of Technical Specification 5.4.1.a involving the failure to have an adequate procedure to ensure the availability of on-site emergency ac power sources following the fourhour coping period of a postulated station blackout. Specifically, station procedures did not ensure that the station blackout diesel generator would be reliably deployed to fulfill its intended function during sustained high winds. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2008-5050. This finding is more than minor because it is associated with the protection against external factors attribute (wind and grid stability) 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. The inspectors evaluated the significance of this finding using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, and determined it to be of very low safety significance because it did not result in an actual loss of safety function and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event (Section 1R01). |
| 05000458/FIN-2008004-02 | 2008Q3 | River Bend | Inadequate Corrective Actions Results in Multiple Failures of Standby Service Water Switchgear Room Ventialtion Fans | A self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified for the licensees failure to take adequate corrective actions in response to a condition adverse to quality resulting in repetitive failures of the standby service water switchgear room ventilation fans. Following failure of the switchgear fans in July 2008, the licensee found that inappropriate flow switch settings on the fans had been identified in a condition report in October 1999, but no actions had been taken to correct the condition. Subsequently, more failures of the standby service water switchgear room ventilation fans occurred, including nineteen in the past three and one half years, many of which were attributed to flow switch issues. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2008-5761. The finding was more than minor because it affected the equipment performance attribute of the mitigating systems cornerstone, and it directly affected the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to preclude undesirable consequences. Using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to have very low safety significance because the condition did not result in an actual loss of safety function of a single train for greater than its Technical Specification allowed outage time. This finding has a crosscutting aspect in the area of human performance associated with resources in that the licensee failed to maintain long term plant safety by minimization of long standing equipment issues (H.2(a)) (Section 1R12). |
| 05000458/FIN-2008004-03 | 2008Q3 | River Bend | Inadequate Risk Assessment for Transformer Yard Maintenance While Shut Down | The inspectors identified a noncited violation of 10 CFR 50.65(a)(4) involving the licensees failure to assess and manage the increase in risk that may result from proposed maintenance activities. Specifically, while conducting maintenance in the transformer yard during severe weather with high pressure core spray inoperable, the licensee did not assess the affects on the shutdown risk. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2008-05383. The inspectors determined this finding was more than minor since it was similar to Manual Chapter 0612, Appendix E, Example 7.e, and since it caused the licensees risk model to change from a Green to Yellow risk window. In accordance with NRC Inspection Manual Chapter 0609, Appendix K, Maintenance Risk Assessment and Risk Management, the inspectors requested that a senior reactor analyst evaluate the risk of this condition. The analyst determined that this finding was of very low risk significance because the associated risk deficit was less than 1.0E-6 (Section 1R13). |
| 05000458/FIN-2008004-04 | 2008Q3 | River Bend | Turbine Building Siding Failure Below Design Specifications | A self-revealing finding was identified for wind induced turbine building siding failure that occurred significantly below design specified stress levels as a result of design and installation deficiencies. This resulted in a forced outage to repair transformer damage and to repair the turbine building siding. The licensee missed prior opportunities to identify turbine building siding design and installation deficiencies following damaging wind events in 1992 and 2005. The licensee entered this issue into the corrective action program as Condition Report CR-RBS-2008-5176. This finding is more than minor because it is associated with the protection against external factors attribute (wind and grid stability) of the initiating events cornerstone and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors evaluated the significance of this finding using Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, and determined it to be of very low safety significance because the finding did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available (Section 4OA3). |
| 05000458/FIN-2014003-01 | 2014Q2 | River Bend | Failure to Follow Tagging Clearance Instructions | The inspectors reviewed a self-revealing non-cited violation of Technical Specification 5.4.1.a., "Procedures," for the failure to adhere to procedural requirements to ensure that other fire suppression ring header valves are/are not correctly positioned. Specifically, on May 19, 2014, the licensee failed to follow the specified instructions in tagging clearance 1C16 / 251-001-O-FPW-P1A, to verify that there were no other ring header valves isolated before implementing the clearance, resulting in the inadvertent isolation of the fire protection ring header. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2014-02489. The failure to follow procedures is a performance deficiency. The performance deficiency is more than minor and, therefore, a finding because it adversely impacted the protection against external factors attribute of the Mitigating System Cornerstone, in that the licensee isolated the fire suppression header to the majority of the plant for approximately 36 hours. Using NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," dated June 19, 2012, the inspectors determined that the issue affected the Mitigating Systems Cornerstone and that the finding pertained to a degraded condition while the plant was in operation. As a result, the inspectors were directed to Inspection Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process," dated September 20, 2013. The inspectors determined that Appendix F did not address the loss of the fire protection ring header to most of the facility and Appendix F, "Assumptions and Limitations," states "the SDP approach is intended to support the assessment of known issues only in the context of an individual fire area. A systematic plant-wide search and assessment effort is beyond the intended scope of the fire protection SDP." Therefore, a senior reactor analyst (SRA) performed a detailed risk evaluation. The total exposure period was 36 hours. The bounding change to the core damage frequency was 2E-7/year. The bounding change to the large early release frequency was 4E-8 per year. The finding was of very low safety significance (Green). The dominant core damage sequences included a fire-induced loss of offsite power, failure of operators to suppress the fire, and damage to Division I, II, and III components. The reactor core isolation cooling system and the short exposure period helped to minimize the risk. The finding has a cross-cutting aspect in the area of human performance associated with avoiding complacency because the licensee failed to recognize and plan for the possibility for mistakes and did not implement appropriate error reduction tools (H.12). |
| 05000458/FIN-2014003-02 | 2014Q2 | River Bend | Licensee-Identified Violation | License Condition 2.C(10), "Fire Protection," requires the licensee to "...comply with the requirements of the fire protection program as specified in Attachment 4 (of the license)." Provision 1 of Attachment 4 states in part: "EOI shall implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report for the facility through Amendment 22 and as approved in the SER dated May 1984 and Supplement 3 dated August 1985 subject to provisions 2 and 3." Section 9.5.1 of the Updated Final Safety Analysis Report (UFSAR), "Fire Protection System," Subsection 9.5.1.4., "Inspection and Testing Requirements," states that "Periodic operational checks, inspections, and servicing required to maintain fire protection systems that protect equipment that is important to safety, including the alarm and detection systems, conform with the RBS Technical Requirements Manual." Technical Requirements Manual, Section TRM 3.7.9.2, Action A.2, states that if "One or more of the...required spray or sprinkler systems (are) inoperable," the licensee would be required to "Establish a continuous fire watch with backup fire suppression equipment for those areas in which redundant systems or components could be damaged," and "Establish an hourly fire watch patrol for other areas," within a completion time requirement of one hour. Contrary to the above, on May 20, 2014, the licensee failed to establish fire watches within the one hour requirement, specified in TRM 3.7.9.2., after it was determined that the fire protection ring header was inoperable. The failure to adhere to the requirements of TRM 3.7.9.2 action A.2, to ensure that all required hourly fire watches are posted within one hour from the time of entry into the TRM, is a performance deficiency. The performance deficiency was more than minor and, therefore, a finding because it adversely impacted the protection against external factors attribute of the Mitigating System Cornerstone, in that the failure to post fire watches, in a timely manner, could result in preventing prompt detection and extinguishing of fires. Using NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," dated June 19, 2012, the inspectors determined that the issue affected the Mitigating Systems Cornerstone and that the finding pertained to a degraded condition while the plant was in operation. As a result, the inspectors were directed to Inspection Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process," dated September 20, 2013. Since the finding affected many fire areas, the inspector consulted with a senior reactor analyst. The analyst determined that Appendix F was not a suitable tool to process this finding, and that a detailed risk evaluation needed to be performed. Although the exposure period for this finding was just a few hours, the risk analyst determined that the detailed risk evaluation performed for the finding described above (Section 4OA2.3.b.) fully bounded this finding as well. As a result of the referenced evaluation, the finding was found to be of very low safety significance (Green). The dominant core damage sequences included a fire-induced loss of offsite power, failure of operators to suppress the fire, and damage to Division I, II, and III components. The reactor core isolation cooling system and the short exposure period helped to minimize the risk. This violation is being treated as a non-cited violation (NCV), consistent with Section 2.3.2.a of the Enforcement Policy because it was of very low safety significance (Green) and it was entered into the licensee's corrective action program as Condition Report CR-RBS-2014-02489 to address recurrence. |
| 05000458/FIN-2015007-02 | 2015Q4 | River Bend | Failure to Obtain Prior NRC Approval for a Change in Reactor Core Isolation Cooling Injection Point | The team identified a Severity Level IV, Green, non-cited violation of 10 CFR 50.59, Changes, Tests, and Experiments, Section (c)(2) which states, in part, A licensee shall obtain a license amendment pursuant to Section 50.90 prior to implementing a proposed change, test, or experiment if the change, test, or experiment would result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the final safety analysis report (as updated). Specifically, prior to October 8, 2015, the licensee failed to correctly evaluate that a spurious reactor core isolation cooling actuation injecting into the feedwater line resulted in a more than minimal increase in the frequency of occurrence of the loss of feedwater heating accident previously evaluated in the updated final safety analysis report. In response to this issue, the licensee initiated a condition report to document completion of a new evaluation under current regulatory guidelines. This finding was entered into the licensees corrective action progam as Condition Report CR-RBS-2015-7259. The team determined that the failure to perform an adequate evaluation of a design change was a performance deficiency. This finding was also evaluated using traditional enforcement because it had the potential to impact the NRCs ability to perform its regulatory function. 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 to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences, and there was a reasonable likelihood that the change would have required NRC review and approval prior to implementation. Specifically, the licensee failed to correctly evaluate that a spurious reactor core isolation cooling actuation injecting into the feedwater line resulted in a more than minimal increase in the frequency of occurrence of the loss of feedwater heating accident previously evaluated in the updated final safety analysis report. 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 where the mitigating structure, system, or component maintained its operability or functionality. Since the violation is associated with a Green reactor oversight process violation, the traditional enforcement violation was determined to be a Severity Level IV violation, consistent with the example in paragraph 6.1.d(2) of the NRC Enforcement Policy. There is no cross-cutting aspect assigned to this performance deficiency because the performance deficiency is not indicative of current performance and also because cross-cutting aspects are not assigned to traditional enforcement violations. |
| 05000458/FIN-2017009-01 | 2017Q2 | River Bend | Failure to Obtain Prior NRC Approval for a Change in Reactor Core Isolation Cooling Injection Point | Green. The NRC identified a Severity Level IV violation for the licensees failure to restore compliance for a non-cited violation (NCV) associated with failure to obtain NRC approval prior to making a change to the reactor core isolation cooling injection point. Specifically, as of April 28, 2017, the licensee had not restored compliance with a violation the NRC identified on October 8, 2015. This violation described a previously made change to the facility without prior NRC approval in violation of 10 CFR 50.59, Changes, Tests, and Experiments. The team determined that the licensees failure to restore compliance within a reasonable amount of time was a performance deficiency. Title 10 CFR 50, Appendix B, Criterion XVI, requires in part that, measures shall be established to assure that conditions 3 adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2017-03505. The finding was more than minor because it is associated with the initiating events aspect of the reactor safety cornerstone and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during power operations. The finding is of very low safety significance (Green) because it did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The finding has a human performance cross-cutting aspect associated with procedural adherence because individuals failed to follow the procedures delineated by the corrective action program (H.8). Originally, the licensee met the criteria for dispositioning the issue (50.59) as a NCV. However, based upon the fact that the condition report, which documented the NCV, was closed without restoring compliance, the licensee no longer met the criteria for a NCV and therefore, this violation is being cited in a notice of violation |
| 05000482/FIN-2018007-01 | 2018Q2 | Wolf Creek | Failure to Provide Adequate Work Instructions for Preventive Maintenance on Safety-Related Equipment | The team reviewed a Green, self-revealed non-cited violation of Technical Specification 5.4.1.a to establish, implement, and maintain written procedures recommended by Regulatory Guide 1.33, Appendix A, Revision 2. Specifically, work instructions for the preventive maintenance for the train B Class 1E electrical equipment A/C unit SGK05B, lacked adequate guidance for preventive maintenance and calibration of the associated thermostat. This resulted in the loss of cooling failure of the A/C unit SGK05B,on February 12, 2018. |
| 05000482/FIN-2018007-02 | 2018Q2 | Wolf Creek | Minor Violation | Performance Deficiency: Failure to promptly identify and correct known-defective switches in inservice safety-related breakers, or to control nonconforming breakers accepted into warehouse stores, as required by 10 CFR 50 Appendix B Criteria XV and XVI. In February 2008, the licensee received a notification from GE Hitachi of reduced reliability of some safety-related circuit breakers due to defective cutoff switches internal to the breakers. The licensee incorrectly screened this information as not applicable to the Wolf Creek Generating Station. In August 2011, after licensee engineers received the information again from industry peers, the licensee screened the information as applicable. The licensee then added steps to its overhaul and pre-install test procedures to check for the defective subcomponent. These steps were performed during subsequent regularly scheduled overhaul or pre-install tests, with the last affected switches being replaced in June 2014 and the last potentially susceptible safety-related breaker being inspected in March 2015. The team determined that because the station had information on the defect in February 2008, but did not correct the condition until 2014 and did not confirm that it was corrected until 2015, the licensee had failed to promptly identify and correct a condition adverse to quality. Further, the licensee failed to inspect or place administrative controls on potentially affected spare breakers that had been accepted into warehouse stores, though the added steps in the pre-install procedure likely would have prevented a defective component from being installed. However, by failing to segregate the potentially affected components until they were inspected, the licensee failed to comply with quality assurance requirements for control of nonconforming components. On June 26, 2018, the licensee put a hold on four potentially affected breakers that were in warehouse stores. The licensee documented this performance deficiency in CR 124693. Screening: The performance deficiency was minor because the licensee did not experience an inservice failure as a result of the defect during the 6 years they remained in service and had a procedure in place that would likely have prevented a defective spare from being issued for installation. Therefore, there was no adverse effect on the mitigating systems cornerstone objective and there was no potential to create a more significant safety concern. Enforcement: This failure to comply with 10 CFR 50 Appendix B Criteria XV and XVI constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy. |
| 05000498/FIN-2012005-01 | 2012Q4 | South Texas | Failure to Perform Pressure Testing of the Reactor Vessel Flange LEAK-OFF Lines | Inspectors identified a non-cited violation of 10CFR50.55a(g)(4) involving the licensees failure to perform a system pressure test of the reactor vessel flange leak-off line of Units 1 and 2, in accordance with the applicable edition of Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code. Contrary to the above, prior to November 1, 2012, the licensee failed to perform the required pressure test of the reactor vessel flange seal leak-off line for both units. Specifically, the licensee failed to implement the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section XI, Class 2 requirements for pressure retaining components as provided by Article IWC 5220, System Leakage Test. The licensee entered the finding into their corrective action program as Condition Report 12-28600. The inspectors determined that the licensees failure to perform a pressure test of the reactor vessel flange leak-off line was a performance deficiency. This finding was more than minor because it affected the Initiating Events Cornerstone attribute of Equipment Reliability and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions. Using Manual Chapter 0609, Attachment A, The Significant Determination Process (SDP) for Findings At-Power, the finding was determined to be of very low safety significance (Green) because the finding did not result in exceeding the reactor coolant system leak rate for a small loss-of-coolant accident, and did not affect other systems used to mitigate a loss-of-coolant accident resulting in a total loss of their function. This issue did not have a crosscutting aspect associated with it because it is not indicative of current performance |
| 05000498/FIN-2012005-02 | 2012Q4 | South Texas | Failure to Follow Procedure for the Control of Tools for Use on Stainless Steel | Inspectors identified a non-cited violation of very low safety significance of Technical Specification 6.8.1.a and Regulatory Guide 1.33, for the failure to follow procedures that ensured abrasive tools for use on stainless steel systems were not contaminated with carbon steel. Specifically, the inspectors determined that the licensee was not maintaining tools as required by Procedure 0PGP03-ZG-0001, Control of Materials and Products By User Groups, Revision 30, and Procedure 0PNP01-ZP-0032, Tools and Measuring &Test Equipment Control, Revision 6, because inspectors observed multiple instances of tools coded for use on stainless steel or aluminum bronze stored with tools marked for use on carbon steel, rust deposits on tools marked for use on stainless steel, and rust deposits on stainless steel components in the plant. This indicated that carbon steel contaminated tools may have been used on these systems. The licensee took corrective actions to segregate the coded tools and trained tool room attendants to properly store and mark abrasive tools designated for use on stainless steel, and evaluated the systems with indications of rust deposits. This issue was entered into the licensees corrective action program as Condition Report 12-28689. Inspectors determined the failure to assure that abrasive tools designated for exclusive use on stainless steel were stored separately from tools used on other materials was a performance deficiency. This finding was more than minor because it affected the Initiating Events Cornerstone attribute of Equipment Reliability and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions. Using Manual Chapter 0609, Attachment A, The Significant Determination Process (SDP) for Findings At-Power, the finding was determined to be of very low safety significance (Green) because the finding did not result in exceeding the reactor coolant system leak rate for a small loss-of-coolant accident, and did not affect other systems used to mitigate a loss-of-coolant accident resulting in a total loss of their function. This finding had a cross-cutting aspect in the area of human performance work practices in that the licensee failed to effectively communicate expectations regarding procedural compliance, and personnel did not follow procedures. Specifically, the inspectors observed that although there were requirements to segregate tools, tools were not consistently segregated when returned to the storage locations as required by procedures |
| 05000498/FIN-2012005-03 | 2012Q4 | South Texas | Failure to Maintain Adequate Fire Penetration Seal Material Thickness | The inspectors identified a non-cited violation of Technical Specification 6.8.1.d, Fire Protection Program Implementation, for the failure to follow work order package instructions requiring the use of Drawing C012-00081-F7F, Detail E-1 Silicone Elastomer Typical Electrical Pen. Seals (Walls & Floors), to establish 6 inches of fire retardant sealant material for penetrations in Units 1 and 2. The inspectors noticed that Unit 1 train B safety-related 4160 Vac switchgear room electrical penetration F4476 had gaps around the edge. A design change installed new electrical cables that required the penetration be sealed using work order package 139376, that stated the penetration seal WILL BE IAW the Penetration Seal Permit and detail Drawing C012-00081-F7F. During the repair activities to correct the gaps, it was discovered that a portion of the seal was only 4.5 inches. The licensee captured this issue as Condition Report 12-28283. Corrective actions included restoring the seal to 6 inches, performing additional analysis to support a 3-hour fire barrier with just 5 inches, and performing extent of condition inspections. The finding was more than minor because it was associated with the Initiating Events Cornerstone attributes of Design Control and Procedure Quality, and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions because it resulted in multiple fire penetration seals being declared nonfunctional as a result of being less than the design thickness. The inspectors used Manual Chapter 0609, Attachment 0609.04, to determine that fire protection issues are processed through Appendix F, Fire Protection Significance Determination Process, dated February 28, 2005. The inspectors used Appendix F, Attachment 1, to determine that the finding was of very low safety significance because it was a Moderate A fire confinement issue that screened out using Task 1.3.2 questions, since the seals would still have provided a 2-hour fire endurance rating or a 20 minute fire endurance rating without the seal being subject to direct flame impingement. In addition, this finding had human performance cross-cutting aspects associated with work practices because the licensee did not communicate human error prevention techniques such as self and peer checking, commensurate with the risk, such that the work activity was performed safely |
| 05000498/FIN-2012005-04 | 2012Q4 | South Texas | Licensee-Identified Violation | Technical Specification 3.0.4 requires, in part, that entry into a mode or other specified condition in the applicability shall only be made when the associated actions to be entered permit continued operation for an unlimited period of time, or after performance of a risk assessment addressing inoperable systems, or when specifically allowed by the specification. Contrary to the above, in April 2010 and November 2011, Unit 2 transitioned from Mode 4 to Mode 3 without all required equipment being operable, without performing a risk assessment, and when not allowed by the specification. Specifically, the turbine trip signal from the reactor trip breakers, the turbine trip signal from the reactor trip signal, and the turbine trip signal from a steam generator HI-HI level were all inoperable due to a jumper being installed for testing when the plant transitioned from Mode 4 to Mode 3. The inspectors used Manual Chapter 0609, Appendix A since the finding was identified after residual heat removal was secured, and determined that the finding was of very low safety significance because the finding did not contribute to both the likelihood of a reactor trip and the loss of mitigation equipment. The licensee entered this issue into the corrective action program as Condition Report 11-27377. |
| 05000528/FIN-2004003-02 | 2004Q2 | Palo Verde | Containment Purge Penetration Nonconformance | A Severity Level IV noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," was identified for the failure to correct a nonconforming condition in a timely manner. Specifically, since June 2001, the licensee discontinued implementation of required Technical Specification surveillance testing for the containment purge valves by declaring the valves inoperable and installing blind flanges. This issue was entered into the corrective action program as CRDR 2711167 The finding is greater than minor because the licensee's failure to submit a license amendment to correct the nonconforming condition impacted the NRC's ability to perform its regulatory function. Therefore, this finding was considered applicable to traditional enforcement. Although the significance determination process is not designed to assess the significance of violations that potentially impact or impede the regulatory process, the finding can be assessed using the significance determination process. Using the Phase 1 worksheet in Manual Chapter 0609, "Significance Determination Process," the finding is determined to have very low safety significance because it only affected the barrier integrity cornerstone and the installation of blind flanges adequately maintained containment integrity. |
| 05000528/FIN-2004003-06 | 2004Q2 | Palo Verde | Failure to Perform a Complete Shut Down Cooling Heat Exchanger Temperature LOOP Channel Calibration | A Severity Level IV noncited violation of Technical Specification 3.3.11 was identified for the failure to include the resistance temperature detectors in the channel calibration for the shutdown cooling heat exchanger temperature instruments. Specifically, prior to the implementation of Improved Technical Specifications, the licensee did not perform testing of the resistance temperature detectors. Following the implementation of Improved Technical Specifications, the licensee did not perform an in-place qualitative assessment of the resistance temperature detectors' behavior. This issue was entered into the corrective action program as CRDR 280178 The failure to perform a complete shutdown cooling heat exchanger temperature loop channel calibration is determined to have greater than minor significance because the licensee's failure to report the condition impacted the NRC's ability to perform it's regulatory function. Therefore, this finding was considered applicable to traditional enforcement. Although the significance determination process is not designed to assess the significance of violations that potentially impact or impede the regulatory process, the finding can be assessed using the significance determination process. Using the Phase 1 worksheet in Manual Chapter 0609, "Significance Determination Process," this finding is determined to be of very low safety significance because it only affected the mitigating system cornerstone and the resistance temperature detectors were found to be within calibration. |