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05000255/FIN-2018001-032018Q1PalisadesLicensee-Identified ViolationA violation of very low safety significance (Green) was identified by the licensee, has been entered into the licensees corrective action program, and is being treated as a Non-Cited Violation consistent with Section 2.3.2 of the Enforcement Policy. Enforcement:Violation: Technical Specification 3.7.6 requires that the combined useable volume of the Condensate Storage Tank (CST) and Primary Makeup Storage Tank (T81) shall be greater or equal than 100,000 gallons. LCO 3.7.6, Condition A states that if the useable volume is not within this limit then A.1 Verify OPERABILITY of backup water supplies in 4 hours andA.2 Restore condensate volume to within limit in 7 days. Condition B states that if the Required Action and associated Completion Time is not met then B.1 Be in MODE 3 in 6 hours and B.2 Be in MODE 4 without reliance on steam generators for heat removal in 30 hours. Contrary to the above, on December 7, 2017 and March 3, 2016, the licensee failed to enter and comply with the actions required by LCO 3.7.6 Condition A and Condition B when Primary Makeup Tank Makeup Control Valve CV2008 could not be fully opened, resulting in a combined useable volume of the CST and T81 of less than 100,000 gallons.Significance/Severity Level: The inspectors answered No to all the questions in IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, because even though the CST and T81 volume were considered inoperable by the TS requirements, there was not a loss of safety function because credited backup water sources were available and operable.Therefore, the finding screened as Green.Corrective Action References: The licensee entered these issues into their CAP as CRPLP20175589, CRPLP20175554, CRPLP20175551, and CRPLP20161116
05000255/FIN-2018001-022018Q1PalisadesLicensee Implementation of Enforcement Guidance Memorandum 15002, Enforcement Discretion for Tornado-Generated Missile Protection NoncomplianceOn June 10, 2015, the NRC issued Regulatory Issue Summary (RIS) 201506, Tornado Missile Protection (ML15020A419), focusing on the requirements regarding tornado-generated missile protection and required compliance with the facility-specific licensing basis. The RIS also provided examples of noncompliances that had been identified through different mechanisms and referenced Enforcement Guidance Memorandum (EGM) 15002, Enforcement Discretion For Tornado Generated Missile Protection Non-Compliance, which was also issued on June 10, 2015 (ML15111A269) and revised on February 7, 2017 (ML16355A286). The EGM applies specifically to a structure, system, or component (SSC) that is determined to be inoperable for tornado-generated missile protection. The EGM stated that a bounding risk analysis performed for this issue concluded that tornado missile scenarios do not represent an immediate safety concern because their risk is within the LIC504, Integrated Risk-Informed Decision-Making Process for Emergent Issues, risk acceptance guidelines. In the case of Palisades, the EGM provided for enforcement discretion of up to 3 years from the original date of issuance of the EGM. On December 7, 2017, and as supplemented on January 18, 2018, Palisades submitted a request to the NRC to extend the enforcement discretion from June 10, 2018 to June 10, 2020 (ML17341A415 and ML18018A328, respectively). By letter dated February 16, 2018, the NRC granted the request to extend enforcement discretion until June 10, 2020 (ML18046A675). The EGM permitted NRC staff to exercise this enforcement discretion only when a licensee implements, prior to the expiration of the time mandated by the LCO, initial compensatory measures that provide additional protection such that the likelihood of tornado missile effects were lessened. In addition, licensees were expected to follow these initial compensatory measures with more comprehensive compensatory measures within about 60 days of issue discovery. In accordance with the EGM, the comprehensive compensatory measures are toremain in place until permanent repairs are completed, or until the NRC dispositions the non-compliance in accordance with a method acceptable to the NRC such that discretion is no longer needed. Palisades was licensed prior to issuance of Appendix A to 10 CFR Part 50, General Design Criteria for Nuclear Power Plants (GDC). Specifically, GDC 2, Design Bases for Protection Against Natural Phenomena, and GDC 4, Environmental and Dynamic Effects Design Basis, discuss how SSCs important to safety shall be designed to protect against natural phenomena, such as tornadoes and shall be adequately protected against the dynamic effects of tornadoes, including protection against missiles. Palisades site-specific licensing bases compliance with GDC 2 and GDC 4 are described in the Updated Final Safety Analysis Report (UFSAR) Sections 5.1.2.2 and 5.1.2.4. Palisades protection of SSCs against tornado-generated missiles is also discussed in UFSAR Section 5.5, Missile Protection. On January 31, 2018, the licensee initiated condition report (CR) CRPLP201800556, which identified a nonconforming condition in the Palisades licensing basis. Specifically, the surge line from the component cooling water (CCW) surge tank to the CCW suction line was identified to be potentially vulnerable to a tornado missile through a doorway. The licensee previously identified a CCW system-related vulnerability on March 29, 2017. The March 29, 2017 CCW vulnerability and five additional vulnerabilities of other SSCs, which all received enforcement discretion, are documented in NRC Inspection Report 05000255/2017002 (ML17220A349). The licensee assessed this new vulnerability and concluded that previously established compensatory measures for the CCW system were adequate and no additional comprehensive compensatory actions were required. Therefore, the licensee declared the SSC operable, but nonconforming because no additional compensatory measures designed to reduce the likelihood of tornado-generated missile effects were required and the previously implemented compensatory measures were still in place. Corrective Action: The licensee documented the condition of the SSC in the CAP and documented the SSC as operable but nonconforming.Corrective Action Reference: CRPLP201800556 Enforcement: Violation: Enforcement discretion was applied to the required shutdown actions of the following Technical Specification (TS) Limiting Conditions for Operation (LCOs): TS 3.0.3, General Shutdown LCO (cascading or by reference from other LCOs); andTS 3.7.7, Component Cooling Water (CCW) System.Severity/Significance: The subject of this enforcement discretion associated with tornado missile protection deficiencies was determined to be less than red (i.e., high safety significance) based on a generic and bounding risk evaluation performed by the NRC in support of the resolution of tornado-generated missile non-compliances. The bounding risk evaluation is discussed in EGM 15002, Revision 1, Enforcement Discretion for Tornado-Generated Missile Protection Non-Compliance (ML16355A286). 11 Basis for Discretion:The NRC exercised enforcement discretion in accordance with Section 2.3.9 of the Enforcement Policy and EGM 15002 because the licensee initiated initial compensatory measures that provided additional protection such that the likelihood of tornado missile effects were lessened. The licensee implemented more comprehensive compensatory actions to resolve the nonconforming conditions within the required 60 days. These comprehensive measures were to remain in place until permanent repairs were completed, which for Palisades were required to be completed by June 10, 2020, or until the NRC dispositioned the non-compliance in accordance with a method acceptable to the NRC such that discretion was no longer needed.The disposition of this enforcement discretion closes LER 05000255/201700101, Inadequate Protection from Tornado Missiles Identified Due to Nonconforming Design Conditions.
05000255/FIN-2018001-012018Q1PalisadesFailure to Maintain an Appropriate Documented Work Instruction for Reassembly of Primary Makeup Tank Makeup Control Valve CV2008A self-revealed Green finding and an associated NCV of Technical Specification 5.4.1, Procedures, was identified for the licensees failure to have an adequate maintenance work instruction for the reassembly of Primary Makeup Tank Makeup Control Valve CV2008. Specifically, because a previous CV2008 maintenance activity failed to properly set the height of the CV2008 jam nuts, the valve guide key fell out of place and in December 2017, CV2008 was unable to be manually stroked during surveillance testing
05000456/FIN-2017003-012017Q3BraidwoodFailure to Implement Adequate Radiological Controls for Treated Liquid Radioactive Effluents Containing TritiumThe inspectors identified a finding of very low safety significance and an associated NCV of Title 10 of the Code of Federal Regulations (CFR), Part 20.1406(c), when the licensee failed to conduct operations to minimize the introduction of residual radioactivity onto the site. Specifically, the licensee failed to identify and evaluate the environmental risk and control work practices with a credible mechanism to prevent spills and leaks from reaching groundwater at the circulating water blowdown (CWBD) area, a radiologically unrestricted area in the licensees owner controlled area. Specifically, tritium contaminated sump water was intermittently pumped to the environs. The licensee documented this finding in their corrective action program (CAP) as Issue Report (IR) 4020644. The failure to conduct operations and control work practices with a credible mechanism to prevent spills and leaks to reach groundwater and minimize residual radioactivity onto the site represented a licensee performance deficiency. The performance deficiency was of more than minor significance because it was associated with the Program and Process attribute of the Public Radiation Safety cornerstone and adversely affected the cornerstone objective of ensuring the adequate protection of public health and safety from exposure to radioactive materials released into the public domain as a result of routine civilian nuclear reactor operation. In accordance with IMC 0609, Appendix D, Public Radiation Safety Significance Determination Process, the finding was determined to be of very low safety significance (Green) because the issue involved a radioactive effluent release, but did not: (1) represent a substantial failure to implement the radioactive effluent release program; or (2) result in public exposure that exceeded the dose values in Appendix I to 10 CFR Part 50 and/or 10 CFR 20.1301(e) limits. The inspectors determined that this finding had a cross-cutting component in the area of Human Performance, in the aspect of Challenging the Unknown, because licensee personnel did not stop when faced with uncertain conditions or evaluate and manage risk before proceeding.
05000346/FIN-2016004-012016Q4Davis BesseMispositioned Instrument Air Valves Result in Plant TransientA self-revealed finding of very low safety significance was identified for the licensees failure to appropriately follow station procedures for aligning instrument air valves that support main feedwater (MFW) regulating valve operation. Specifically, two instrument air valves were not aligned to their normal operating position following planned maintenance. As a result, the Steam Generator 2 (SG 12) MFW Regulating Valve momentarily closed during routine steam feedwater rupture control system (SFRCS) surveillance testing and caused a plant transient. Corrective actions taken by the licensee, include but are not limited to, performance of an instrument air valve line up to validate no other valves were out of position; performance of SFRCS Actuation Channel 2 testing to verify no other half trips existed on SFRCS Actuation Channel 2 components; a configuration control stand-down with the instrument and control shop; and revisions to procedural guidance to perform additional valve position verification. The finding was of more than minor significance because it was associated with cornerstone attribute of configuration control and adversely affected the cornerstone objective: To limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The finding was determined to be of very low safety significance because the finding did not cause a reactor scram with the loss of mitigation equipment relied upon to transition the plant from the onset of the scram to a stable shutdown condition (e.g. loss of condenser, loss of feedwater). The inspectors determined that the finding had a cross-cutting aspect in the area of human performance. The inspectors assigned the cross-cutting aspect of Avoid Complacency to the finding because the procedural step to close valve IA1008A was marked as complete but was not performed correctly. Additionally, appropriate human performance error reduction tools were not adequately used to ensure valve manipulations were performed as intended. (H.12)
05000346/FIN-2016004-022016Q4Davis BesseFailure to Adequately Evaluate Degraded Turbine Building Roof VentsA finding of very low safety significance was self-revealed on September 10, 2016, when rainwater intrusion into the automatic voltage regulator caused a generator lockout and reactor trip. Specifically, station management failed to adequately assess the identified degraded condition of the turbine building roof vents in accordance with station expectations and procedures when four roof vents were left stuck open although it was identified by operators that water intrusion was possible onto the stator water cooling skid and automatic voltage regulator on August 17th, 24 days prior to the event. No violation of regulatory requirements was identified because the turbine building roof vents and automatic voltage regulator are not safety related, and the applicable maintenance procedures were not covered under Title 10 of the Code of Federal Regulations (10 CFR) 50, Appendix B. The finding was of more than minor safety significance because it affected the Equipment Reliability attribute of the Initiating Events cornerstone. Specifically, the failure to fully evaluate the risk associated with the stuck open turbine building roof vents affected the availability and reliability of the automatic voltage regulator causing a reactor trip. The inspectors also reviewed the examples of minor issues in IMC 0612, Power Reactor Inspection Reports, Appendix E, Examples of Minor Issues, dated August 11, 2009, and found no similar examples. The finding was determined to be a licensee performance deficiency of very low safety significance because the performance deficiency did not cause a reactor trip with the loss of mitigating equipment. The inspectors determined this finding affected the cross-cutting area of problem identification and resolution and the cross-cutting aspect of evaluation. The licensee did not properly evaluate the problem and assigned an incorrect priority to the work order to address the degraded roof vents. (P.2)
05000237/FIN-2016002-012016Q2DresdenFailure to Implement and Maintain Written Procedures Regarding Breathing Air Quality TestingA finding of very-low safety significance and an associated NCV of Title 10 of the Code of Federal Regulations (CFR), Part 20.1703, was an NRC-identified finding for failure to implement and maintain written procedures regarding breathing air quality that resulted in the failure to perform a continuous in-line breathing air quality test during filling of self-contained breathing apparatus (SCBA) cylinders since 2009. Specifically, on May 4, 2016, during an inspection of the licensees air compressor, the inspectors identified that the in-line carbon monoxide (CO) detector located at the compressor highpressure filling station was inoperable since 2009, the procedure does not specify an alternative method of CO monitoring during the filling of the SCBA cylinders. Without specifying an alternative method of monitoring and only relying on the high-temperature safety shut-off, hazardous CO gas could be introduced into the SCBA cylinders, thus degrading the Grade-D air quality, during a compressor malfunction. The licensees corrective actions included but were not limited to revising the applicable procedures, servicing or replacing the CO monitor by the manufacturer, and installing a new air compressor at the facility. The inspectors determined that that the finding was more than minor in accordance with Inspection Manual Chapter (IMC) 0612, in that the finding impacted the program and process attribute of the Occupational Radiation Safety Cornerstone and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation through the use of SCBAs during an emergency response use by maintaining certified air quality. Specifically, the licensee failed to implement and maintain written procedures regarding an alternative method of monitoring air quality testing to maintain the Grade-D air quality during filling of SCBA cylinders. The finding was determined to be of very-low safety significance in accordance with IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, because it was not an as-low-as-reasonably-achievable planning issue, there was no overexposure nor substantial potential for an overexposure, and the licensees ability to assess dose was not compromised. The inspectors concluded that the cause of the issue involved a cross-cutting component in the area of human performance, resources, in that, the license did not ensure the adequacy of the procedure describing the alternate methods of CO monitoring during filling of Grade D air into the SCBA cylinders. (H.1)
05000346/FIN-2016002-012016Q2Davis BesseMispositioned Instrument Air Valves Result in Plant TransientOn May 31, 2016, at approximately 10:21 a.m., planned testing of Steam and Feedwater Rupture Control System (SFRCS) Actuation Channel No. 1 was in progress. This was the first performance of this test since the unit returned to operation following RFO 19. Unexpectedly, operators in the control room received several overhead annunciator alarms coincident with a rapid swing in plant power and indications that the SG 12 MFW Regulating Valve (SP6A) had gone closed and then reopened. In accordance with established procedures for responding to such an event, control room operators took manual control of integrated control system (ICS) stations for reactor demand, SG/reactor demand, both MFW regulating valves, both MFW startup valves, and both MFW loop demands. The control room crew was then able to arrest the transient and stabilize plant power at approximately 89 percent. Initial evaluation of the transient by the licensee revealed that two instrument air (IA) valves associated with control air for SP6A (IA1008D, SVSP6A1 Bypass; and IA1008A SVSP6A1 Maintenance Isolation) were out of their normal positions. The mispositioned valves had the effect of placing P6A in a half trip condition, such that when SFRCS Actuation Channel No. 1 was being tested SP6A unintentionally responded to the test signal. The licensee entered this issue into their CAP as CRs 201607282, 201607286, 201607337, and 201608363. Because the licensee had yet to complete their investigation and analysis of the event and the IA valve mispositioning by the end of this inspection period, the issue is being treated as an unresolved item (URI) pending the inspectors review of the licensees completed cause evaluation and proposed corrective actions. (URI 05000346/201600201)
05000255/FIN-2016001-032016Q1PalisadesFailure to Meet the Minimum Staffing Requirements of the Fire BrigadeAn NRC-identified finding of very low safety significance and an associated NCV of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Section 48(c) and the National Fire Protection Association (NFPA) Standard 805 Section 3.4.1 was identified for the failure to meet the minimum staffing requirements for the Fire Brigade on January 4 and 5, 2016. Specifically, two nuclear plant operators (NPOs) who had their Fire Brigade qualifications suspended, stood watch as Fire Brigade members during day shift on January 4, 2016 and approximately one half of day shift on January 5, 2016. The licensee entered this issue into their Corrective Action Program (CAP) as CR-PLP-2016-00198, performed an apparent cause evaluation, successfully performed a fire drill to requalify the Fire Brigade members with suspended qualifications on January 6, 2016, and planned to update the tracking method used to validate drill completion for Fire Brigade qualifications. The performance deficiency was determined to be more than minor because it was associated with the Protection Against External Factors attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). The finding screened as having very low safety significance based on using qualitative criteria located in IMC 0609, Appendix M, Significance Determination Process Using Qualitative Criteria. The finding had a cross-cutting aspect of Documentation in the Human Performance cross-cutting area because the licensee informally tracked drill completion and this information was not accessible to each individual Fire Brigade member to validate their qualifications (H.7).
05000255/FIN-2016001-012016Q1PalisadesDesign Review of Modification to Track Alley Wall for Dry Fuel Storage ActivitiesThe inspectors identified a unresolved item (URI) associated with the design review of a modification to the Track Alley wall for dry fuel storage (DFS) campaign activities. Specifically, the licensee is currently revising the process applicability determination (50.59 and 72.48 screenings), and reviewing any necessary actions, associated with altering the newly modified wall in support of upcoming DFS campaign activities. The wall, a protective barrier with safety functions per the UFSAR, in its newly modified condition, will be altered when the steel plate covering the opening cut into it will be raised to accommodate the DFS transporter. The DFS campaign is currently on hold pending resolution of other issues. In January 2016, the licensee began work on an engineering change to permanently modify the west wall of Track Alley in order to accommodate the new transporter used for moving the casks associated with the dry fuel storage campaign. This modification removed a section of the reinforced concrete wall by cutting out an opening approximately 9 feet wide by 4 feet high by 18 inches deep into the existing wall. A three inch thick steel plate was mounted onto vertical rails which can slide down to cover the window cut into the wall and raised to open the window for when the transporter is brought into Track Alley. The west wall of Track Alley is also the east wall of the Technical Support Center (TSC). This wall is designed to withstand seismic, high wind, and tornado missile loads. It also serves as a radiation protection barrier for personnel in the TSC during emergency situations. The permanent modification of cutting the opening in the wall and installing the steel plate, to provide equivalent protection of the 18 inches of concrete that were cut out, was evaluated in Engineering Change 59170 and calculation EAEC5917001. The inspectors reviewed these documents, the supporting process applicability determination (50.59 screening), and risk assessment of implementing the design change. During this review, the inspectors identified that the licensee did not assess the alteration of the wall, a protective barrier with safety functions per the UFSAR, when the steel plate covering the window would need to be raised to accommodate the DFS transporter. The inspectors questioned this condition and the licensee subsequently completed a process applicability determination (PAD) form (72.48 and 50.59 screening). When reviewing the PAD, the inspectors questioned the licensees underlying assumption that moving the steel plate to uncover the window was considered to be in support of a maintenance activity and, hence, screened out of the 50.59 process, including not requiring certain compensatory actions for the walls safety functions during the period of time in which the opening was exposed. At the end of the inspection period the licensee was reviewing their assessment. Once their review is completed, including any changes that may be made, the inspectors will re-assess their evaluation and determine what actions, if any, will need to be accomplished in support of the DFS campaign. Since the campaign is on hold, a URI is being opened to track resolution of this issue.
05000255/FIN-2016001-052016Q1PalisadesLicensee-Identified ViolationTS Limiting Condition for Operation (LCO) 3.0.6 states, in part, that when a supported system LCO is not met solely due to a support system LCO not being met, the Conditions and Required Actions associated with this supported system are not required to be entered; only the support system LCO actions are required to be entered. TS LCO 3.0.6 further specifies that an evaluation shall be performed in accordance with TS 5.5.13, Safety Function Determination Program. Palisades Administrative Procedure 4.11, Safety Function Determination Program, step 5.4.3 requires documentation of entry into TS LCO 3.0.6 for the inoperable supported system in the Operations Log. Contrary to the above, on January 19, 2016, the licensee failed to document entry into TS LCO 3.0.6 in the operations log when work was commenced on breaker 521214, Motor Control Center (MCC) 22 and MCC24 480 Volt feeder breaker. The licensee identified this issue when a similar condition was entered on January 22, 2016 and documented the missed entry into TS LCO 3.0.6 in CRPLP201600413, Operations Failed to Log Entry into LCO 3.8.1B and LCO 3.5.2B or LCO 3.0.6. The licensee provided coaching to the individuals involved. The inspectors screened the issue using IMC 0609, Appendix A, The Significance Determination Process for Findings at Power, Exhibit 2, Mitigating System Screening Questions, and answered No to all the questions. Therefore, the finding screened as having very low safety significance (Green).
05000255/FIN-2016001-042016Q1PalisadesLicensee-Identified ViolationTitle 10 CFR 50.54(m)(2)(iii), Condition of Licenses, states that when a nuclear power unit is in an operational mode other than cold shutdown or refueling, as defined by the units technical specifications, each licensee shall have a person holding a senior operator license for the nuclear power unit in the control room at all times. TS 5.2.1 states in part, that during any absence of the Shift Supervisor from the control room while the plant is in Mode 1, an individual with an active Senior Reactor Operator (SRO) license shall be designated to assume the control room command function. Contrary to the above, at approximately 2:00 a.m. on September 2, 2015, with the unit in Mode 1, the Command SRO left the control room without another SRO being present in the control room and without turning over the command function. A few minutes prior to the event, the shift Command SRO turned over to the Shift Technical Advisor (STA) the Command SRO function of the control room so that the shift Command SRO could take a break outside the control room boundary. A minute or so after the STA (who had the Unit Command SRO function at the time) left the control room, a control room reactor operator observed that there were no SROs in the control room and summoned the Shift Manager from an office across the hall to the control room. The Shift Manager then assumed the Command SRO function and the STA was called back to the control room. This issue was identified by the licensee on September 2, 2015, and documented in CRPLP201503637, The SRO with Command and Control Momentarily Left the Control Room. There were no risk-significant plant evolutions in progress and no adverse reactor plant operations occurred during the SROs absence. The STA was relieved from shift responsibilities until corrective actions were taken. The inspectors screened the issue using IMC 0609, Appendix A, The Significance Determination Process for Findings at Power. The inspectors reviewed the screening questions under all three Cornerstones and all of the logic questions did not apply, therefore the finding screened as having a very low safety significance (Green).
05000255/FIN-2016001-022016Q1PalisadesMovement of Radioactive Material Results in an Unposted and Un-Barricaded High-Radiation AreaA self-revealed finding of very low safety significance and an associated NCV of Technical Specification 5.7.1 was identified when movement of a bag of radioactive material caused an area to become a high radiation area without the proper posting and barricades. The licensee immediately moved this bag of radioactive material to a posted locked high-radiation area and entered this issue into their CAP as CRPLP201505019. The performance deficiency was determined to be more than minor because it was associated with the Program and Process attribute of the Occupational Radiation Safety cornerstone and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation. Specifically, the movement of the bag from an area that was a high-radiation area to an area that was not posted and barricaded as a high-radiation area removed a barrier that was intended to prevent workers from receiving unexpected dose. The finding was determined to be of very low safety significance in accordance with IMC 0609 Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008. The violation was of very low safety significance because: (1) it did not involve as-low-as-reasonably-achievable planning or work controls, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. The finding had a cross-cutting aspect of Teamwork in the Human Performance cross-cutting area because the individuals and work groups involved did not communicate or coordinate their activities within and across organizational boundaries to ensure nuclear safety was maintained (H.4).
05000440/FIN-2015003-022015Q3PerryFailure to Properly Implement Steps Outlined in a Technical Specification Surveillance ProceA finding of very low safety significance and associated NCV of Technical Specification (TS) 5.4.1., Procedures, was self-revealed on August 5, 2015, when an unexpected isolation of the reactor core isolation cooling (RCIC) system occurred as a result of the licensees failure to properly implement the steps outlined in TS Surveillance Procedure, SVIE31T5395B, RCIC Steam Line Flow High Channel Functional for 1E31N684B. Specifically, during performance of the surveillance, several steps were marked as not applicable that were applicable to prevent the isolation of the RCIC system. As a result, the licensee failed to lift leads as required by the procedure and the RCIC steam supply inboard isolation valve then closed when the isolation trip signal was applied during the test. The licensee took immediate actions to restore system operability and availability and conducted a human performance event response investigation. A standing order for both Operations and Instrumentation and Controls personnel was initiated addressing interim actions for control room surveillance performance and to reinforce maintenance fundamentals and human performance behaviors. The licensees failure to properly implement the steps in the procedure was a performance deficiency that was determined to be more than minor, and thus a finding, because it was associated with the Mitigating Systems Cornerstone attribute of equipment performance and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was determined to be of very low safety significance because it 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 for greater than 24 hours. This finding has a cross-cutting aspect in the area of human performance, avoid complacency, for failing to recognize and plan for the possibility of mistakes, and for failure to implement appropriate error reduction tools, such as proper self-checks and peer checks, which resulted in an isolation of the RCIC system.
05000440/FIN-2015003-012015Q3PerryInadequate Operating Procedure for Diesel Generator Building Ventilation SystemThe inspectors identified a finding of very low safety significance and associated non-cited violation (NCV) of Title 10 of the Code of Federal Regulations Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure as of July 8, 2015, to establish and maintain an adequate procedure for operation of the Diesel Generator Building Ventilation System (DGBVS). Specifically, the DGBVS operating procedure did not ensure that diesel room temperature would remain below limits during testing. The failure to establish and maintain an adequate procedure was a performance deficiency and resulted in the Division 2 Diesel Generator room temperatures exceeding specified limits. The performance deficiency was more than minor, and thus a finding, because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined that the finding was of very low safety significance because the finding is a deficiency affecting the design or qualification of a mitigating structure, system, and component (SSC) that maintained its operability. This finding has a cross-cutting aspect in the area of human performance, design margins, because the licensee did not incorporate the degree of redundancy specified in the Updated Safety Analysis Report for DGBVS into the applicable operating procedures.
05000440/FIN-2015003-032015Q3PerryLicensee-Identified ViolationDuring a review of items entered in the licensees CAP, the inspectors identified that corrective action item (CR 201416769) documented direct and root causes for the reactor scram that occurred on November 7, 2014. The inspectors reviewed the root cause analysis and the corrective actions taken to prevent recurrence. The direct cause of the event was injection of a false feed flow runback signal, caused by the redundant reactivity control system (RRCS) self-test feature, into the digital feedwater control system DFWCS) which caused both contacts in the A and B divisions to close simultaneously, thus actuating a real feedwater runback. The licensee determined that the design was not adequate to prevent this event from occurring and that the root cause of the event was a latent design flaw from the original digital upgrade design package. The latent design flaw was identified by the licensee as a violation of 10 CFR 50, Appendix B, Criterion III, Design Control, which requires in part, design control measures for verifying or checking the adequacy of the design. The corrective actions, which consisted of physical modifications to plant equipment, were previously reviewed by inspections conducted during the refueling outage, March and April of 2015, and documented in Perry Integrated IR 2015002. The inspectors evaluated the licensee-identified violation using IMC 0612, Appendix B, Issue Screening, and determined that the deficiency was more than minor because it was associated with the Initiating Events Cornerstone attribute of design control and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors further evaluated the issue in accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, dated June 19, 2012, and determined that the safety significance of this event was very low because, in accordance with the initiating events screening questions, all safety systems functioned as designed and the scram was not complicated. This issue is also discussed in Sections 4OA3.1 and 4OA7.
05000255/FIN-2013005-072013Q4PalisadesPeriodic Design Basis Testing of Safety-Related Electrical ComponentsThe licensees Quality Assurance Program Manual stated they were committed to Regulatory Guide 1.30, Quality Assurance Requirements for the Installation, Inspection, and Testing of Instrumentation and Electric Equipment. This Regulatory Guide endorsed Institute of Electrical and Electronics Engineers (IEEE) Standard 336-1971 (also known as American National Standards Institute (ANSI) N45.2.4-1972) as adequate for demonstrating compliance with the pertinent quality assurance requirements of 10 CFR Part 50, Appendix B. In addition, Section C.3 of Regulatory Guide 1.30 stated, Although Subdivision 1.1 of ANSI N45.2.4-1972 states the requirements promulgated apply during the construction phase of a nuclear power plant, these requirements are also to be considered applicable for the installation, inspection, and testing of instrumentation and electric equipment during the operation phase of a nuclear power plant. IEEE Standard 336-1971, Section 3.3, Procedures and Instructions, required the licensee to produce documents that shall be kept current by controlled supervision so that installation, inspections, and tests are performed in accordance with the latest approved design and manufacturers instructions. However, while reviewing the licensees management of component design life, the inspectors noted the licensee did not periodically test safety-related electrical components to the design requirements. The licensee interpreted the intent of Section C.3 of Regulatory Guide 1.30 as to apply IEEE 336 requirements only to modifications and activities that were similar to initial construction activities. This issue is a URI pending further review, including consultation with the Office of Nuclear Reactor Regulation, and determination of further NRC actions to resolve the Issue.
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