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| Quarter | Site | Title | Description | |
|---|---|---|---|---|
| 05000237/FIN-2018003-02 | 2018Q3 | Dresden | Licensee-Identified Violation | Violation: Dresden Technical Requirements Manual (TRM) Control Program (Appendix G of TRM), Section 1.5, Program Implementation, requires that proposed changes to the TRM are screened and reviewed under the 10 CFR 50.59 process in accordance with plant specific procedures. Contrary to the above, in October 2017 Dresden station approved and implemented an extension to the surveillance frequency of DIS 150020, Division I & II Low Pressure Coolant Injection (LPCI) Pumps Suction and Injection Valves Circuitry Logic System Functional Test, on Unit 2 per the Surveillance Frequency Control Program (SFCP) without the required 50.59 review. |
| 05000249/FIN-2018003-01 | 2018Q3 | Dresden | Failure to Follow Maintenance Procedures for Assembling Unit 3 HPCI Room Cooler Fan | A self-revealing, Green non-cited violation (NCV) of Technical Specification (TS) 5.4, Procedures, was identified for the licensees failure to follow maintenance procedures DMP 570004, LPCI and HPCI Room Cooler Maintenance, and DEP 570004, HPCI Room Cooler Fan Preventive Maintenance, when assembling the Unit 3 HPCI room fan. Specifically, on one occasion when maintenance was performed on the fan, technicians installed the cam locking collar in the opposite direction of the fan shaft rotation, and on the other occasion, technicians tensioned the fan belt to the wrong value and misadjusted the alignment of the shaft sheave. Over time, this improper maintenance caused the inboard and outboard fan bearings to wear on the shaft, causing increased vibrations, and eventually leading to HPCI being declared inoperable to emergently work on the fan |
| 05000373/FIN-2018002-04 | 2018Q2 | LaSalle | Minor Violation - Follow-up of Events and Notices of Enforcement Discretion | Minor Violation: For S/RV 2B21F013L, serial number N63790050012 (hereafter referred to as S/RV 12), the licensee completed a work group evaluation as documented in AR 03975216ACIT No. 3 to investigate the cause for two S/RVs that failed a set pressure lift test out of specification low. For ACIT No. 3, the licensee staff incorporated a vendor letter that documented the results of the S/RV vendors review of the S/RV 12 condition and which recorded an out of tolerance spring condition. It stated that The spring was measured and rate tested. The free height was found to be below the minimum original equipment manufacturer specified tolerance. The licensees vendor subsequently replaced the nonconforming spring with a new spring. In prior vendor correspondence with the licensee (reference E-mail dated June 24, 2015), the vendor stated that Typically we contribute a low as-found lift to an out-of-tolerance spring rate or free height dimension. Therefore, the nonconforming spring free height dimension may have caused the low as-found lift setpoint failure for this valve and as such was relevant (e.g. material) to the determination of a failure cause that was reported in LER 05000374/201700400 and 01. However, the licensee failed to identify this during their cause investigation and erroneously reported in LER 05000374/201700400 and 01 that The vendor reported for both valves that all the spring tolerances were within the acceptance limits. The licensee documented this violation in AR 04134591, Potential Minor Violation for Unit 2 LER 20170401. The licensee also submitted a revision to the LER as LER 05000374/201700402 Screening: The significance determination process does not specifically consider the regulatory process impact in its assessment of licensee performance. Therefore, it is necessary to address this violation which could impede the NRCs ability to regulate using traditional enforcement to adequately deter non-compliance. The inspectors determined that this issue was a Severity Level IV violation based on Example 6.9.d.10 in the NRC Enforcement Policy which states, A failure to identify all applicable reporting codes on a Licensee Event Report that may impact the completeness or accuracy of other information (e.g. performance indicator data) submitted to the NRC. In accordance with the Section 2.2.1.c of the NRC enforcement policy, the severity level of a violation involving the failure to make a required report to the NRC will depend on the significance of and the circumstances surrounding the matter that should have been reported. The NRC had not relied on information in this LER report to make a regulatory decision, and the inspector answered no to each of the more than minor screening questions in Appendix B of IMC 0612 for the issue of concern. Therefore, the NRC determined this was a minor violation because it was associated with a minor performance deficiency. Violation: Failure to comply with 10 CFR 50.9 Completeness and accuracy of information and accurately report the nonconforming S/RV 12 spring tolerance in LER 05000374/201700400 and 01 to the NRC constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy. |
| 05000373/FIN-2018002-03 | 2018Q2 | LaSalle | Licensee-Identified Violation | This violation of very low safety significant was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a Non-Cited Violation, consistent with Section 2.3.2 of the Enforcement Policy. Violation: Technical Specification LCO 3.4.4 (applicable for Modes 1, 2 and 3) states: The safety function of 12 safety relief valves (S/RVs) shall be OPERABLE, and Action Statement A states that One or more required S/RVs inoperableA.1 be in mode 3 in 12 hours and A.2 be in Mode 4 in 36 hours. Technical Specification SR 3.4.4.1 states that Verify the safety function lift setpoints of the required S/RVs are as follows Number of S/RVs Setpoint (psig 2 1205 36. 3 1195 35. 2 1185 35. 4 1175 35. 2 1150 34. Contrary to the above, during portions of previous Unit 1 and 2 operating cycles from 2012 through January of 2017, two main steam S/RVs did not meet these lift pressure setpoint requirements. Specifically S/RV 2B21F013C lifted at 1131 psig instead of from 1139.8 to 1210.2 psig and S/RV 2B21F013L lifted at 1130 psig instead of from 1159.2 to 1230.8 psig (reference: Licensee Event Report 05000374/201700400; 01, Two Main Safety Relief Valves Failed Inservice Lift Inspection Pressure Test. Significance/Severity: This licensee identified finding affected the Initiating Events Cornerstone and was screened in accordance with Exhibit 1 of IMC 0609, Appendix A, The Significance Determination Process for Findings At Power. The two affected SRVs lifted low outside of their setpoint band, which was conservative with respect to maintaining the reactor coolant system overpressure protection safety function of these valves. Therefore, the inspectors determined that this finding is of very low safety significance (Green) because after a reasonable assessment of degradation, the finding would not have resulted in exceeding the reactor coolant system leak rate for a small LOCA and did not affect other systems used to mitigate a loss-of-coolant accident. Corrective Action Reference: AR 3974669 |
| 05000373/FIN-2018002-02 | 2018Q2 | LaSalle | Failure to Follow Procedure and Perform Database Revision Review Requirements | The inspectors identified a Green finding of very low safety significance for the licensees failure to follow procedure NSWPWM03, Predefine Database Revisions, Revision 0, for retiring procedure LESGM108, Inspection of 480V Motor Control Center Equipment, that performed bus bar inspection on Division 3 motor control centers. Specifically, instead of completing NSWPMW03, step 6.5, Database Revision Review Requirements, to retire the bus bar inspections for Division 3 motor control centers, the licensee retired the procedure based solely on having previously retiring the bus bar inspections for Division 1 and Division 2 in 2002,and did not performthe required review. |
| 05000373/FIN-2018002-01 | 2018Q2 | LaSalle | Failure to Implement a Preventative Maintenance Strategy for Residual Heat Removal Service Water Pump Shorting Relays | A self-revealed Green finding of very low safety significance was identified for the licensees failure to implement a preventative maintenance (PM) strategy for the residual heat removal service water (RHRSW) pump shorting relays in accordance with procedure MAAA716210, Performance Centered Maintenance (PCM) Process, Revision 11. Specifically, a PCM template was issued in 2002 that required periodic as-found testing and calibration for control and timing relays, but a maintenance strategy was never implemented. As a result, one of the normally closed contacts on the Unit 1 D RHRSW pump shorting relay developed a high contact resistance and prevented the Unit 1 D RHRSW pump from starting. |
| 05000255/FIN-2018010-01 | 2018Q1 | Palisades | Licensee-Identified Violation | Violation: Title 10 of theCode of Federal Regulations (CFR) Part 50.55a(g)(4), Inservice Inspection Standards Requirement for Operating Plants, requires that, throughout the service life of a boiling or pressurized water-cooled nuclear power facility, components (including supports) that are classified as ASME Code Class 1, Class 2, and Class 3 must meet the requirements set forth in Section XI of the 2006 edition through 2008 addenda of the ASME Boiler and Pressure Vessel Code. This edition of the AMSE Code requires that a VT3 visual examination of supports other than piping supports be performed once every 10year inservice inspection (ISI) interval. Contrary to the above, since the beginning of plant operation, the safety-related CCW and SW pump lateral supports (classified as ASME Code Section XI Class 3) had never been included in the ISI program and therefore had never had the required VT3 examination performed during each 10year ISI interval. Corrective actions included incorporating the supports into the ISI program, scheduling the inspections as required, and validating that the supports were still capable of performing their safety function and that the CCW and SW systems remained operable.Significance/Severity Level: The inspectors determined that the failure to perform ASME Code Section XI required inspections of the CCW and SW pump lateral supports was a performance deficiency. The inspectors determined the performance deficiency was more than minor because it adversely affected the Design Control attribute of the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the failure to periodically inspect the pump lateral supports could result in the failure to identify a nonfunctional support that could increase the risk of a pump failure.The inspectors assessed the significance of the finding using Appendix A of the SDP. The finding was determined to be of very low safety significance (Green) because although it was a deficiency affecting the design or qualification of a mitigating structure, system, or component (SSC), the SSC remained operable. Corrective Action Reference: CRPLP201705784, OE Review Identified Palisades Failure to Inspect ASME Class 3 Pump Supports for SW and CCW Pumps, 1/26/2018 Safety Conscious Work Environment Observations Based on interviews with plant staff and reviews ofthe latest safety culture survey results to assess the safety conscious work environment on site, the team determined that, in general, plant personnel appeared willing to raise nuclear safety concerns through at least one of the several means available. Most of those interviewed had an adequate knowledge of the CAP process and would initiate a CR, or work with someone who would do so on their behalf, if they knew of a safety concern. A weakness was identified in plant personnel knowledge ofhow to use the electronic CR system. Specifically, there were some personnel who were not familiar with how to generate a CR or how to track the resolution of a CR. Personnel also expressed an overall frustration with feedback provided on a CR; either with difficulties in being able to see how something was resolved or with not being able to understand the decision-making process for the resolution of issues.Most individuals expressed a willingness to raise safety concerns without fear of retaliation and all employees knew the importance of having a strong safety conscious work environment. There were some instances where the free flow of information or a willingness to raise concerns through an individuals direct line of supervision were hampered due to the perception that supervision was not receptive to receiving the concern or addressing the issue. In some cases, this presented an uncomfortable work environment for the affected individuals. However, when presented with this situation, all individuals knew of other supervisors that they could bring their concerns to or other avenues to use to address anissue. All plant personnel were aware of the Employee Concerns Program (ECP), knew who the ECP coordinator was, and most were willing to use it as an avenue to raise concerns, if desired. However, some individuals believed that the ECP lacked the appropriate level of confidentiality to effectively address concerns. |
| 05000255/FIN-2018001-03 | 2018Q1 | Palisades | Licensee-Identified Violation | A 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-2018001-02 | 2018Q1 | Palisades | Licensee Implementation of Enforcement Guidance Memorandum 15002, Enforcement Discretion for Tornado-Generated Missile Protection Noncompliance | On 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-2018001-01 | 2018Q1 | Palisades | Failure to Maintain an Appropriate Documented Work Instruction for Reassembly of Primary Makeup Tank Makeup Control Valve CV2008 | A 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 |
| 05000255/FIN-2017004-01 | 2017Q4 | Palisades | Improperly Connected M&TE Leads to Unexpected AFU Fan Trip | A finding of very low safety significance and an associated NCV of Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed when the licensee failed to follow step 5.4.4.b of Technical Specification surveillance procedure RT85DA, Control Room Emergency Ventilation Filtration Testing A Train. Specifically, the licensee failed to properly connect maintenance and test equipment (M&TE) across flow transmitter test taps which caused V26A, the air filter unit (AFU) VF26A fan, to stop 17 seconds after operators started the fan from the control room. The licensee entered this issue into their Corrective Action Program (CAP) as condition report (CR) CRPLP201705234. Corrective actions included coaching the vendor on ensuring M&TE is properly connected to plant equipment and ensuring suitable field oversight of the vendor during re-performance of the surveillance.The issue was determined to be more than minor in accordance with IMC 0612, Appendix B, Issue Screening, because it was associated with the Barrier Integrity cornerstone attribute of Human Performance and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The finding screened as having very low safety significance (Green) in accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 3, because the inspectors answered "No" to all screening questions. The finding had across-cutting aspect in the area of Human Performance, in the Field Presence aspect, for the failure to ensure supervisory and management oversight of work activities, including contractors and supplemental personnel (H.2). |
| 05000255/FIN-2017003-03 | 2017Q3 | Palisades | 12 Diesel Generator Trip During Maintenance Resulting in Additional Unavailability of the 12 DG | A finding of very low safety significance and an associated NCV of Technical Specification (TS) 5.4.1, Procedures, was self -revealed on March 31, 2017, when the 12 Diesel Generator ( DG ) tripped during performance of monthly TS surveillance procedure MO 7A 2, Emergency Diesel Generator 1 2. Specifically, during conduct of the monthly surveillance procedure, restoration activities associated with maintenance of breaker 152 213, 1 2 DG to Bus 1D, were being performed. When maintenance personnel closed the trip cutouts for the Z -phase of the 1 2 DG differential overcurrent relay, an unbalanced current flow into the differential relay resulted in relay actuation. This actuation resulted in a trip of the output breaker and subsequently the 1 2 DG. The trip caused a delay in the TS surveillance activities and resulted in the extended unavailability and inoperability of the 1 2 DG. The licensee entered this issue into their corrective action program (CAP) as condition report (CR) CR PLP 2017 01291. Corrective actions included retesting the 1 2 DG and updating the work instructions associated with the differential overcurrent relays to include caution statements that opening or closing trip cutouts for the relays while the output breaker s from the DGs to the associated buses were closed could cause the differential relay s to actuate and trip the DG . The issue was determined to be more than minor in accordance with IMC 0612, Appendix B, Issue Screening, because it was associated with the Mitigating System s cornerstone attribute of Procedure Quality and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding screened as having very low safety significance (Green) in accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings At -Power, Exhibit 2, since the inspectors answered No to all screening questions. The finding had a cross- cutting aspect in the area of Human Performance, in the Work Management aspect , for the licensees failure to identify and manage risk commensurate to the work (H.5). |
| 05000255/FIN-2017003-01 | 2017Q3 | Palisades | Left Train Emergency Diesel Generator Load Sequencer Failure | Introduction: The inspectors identified an Unresolved Item ( URI ) associated with the failure of the left train emergency DG load sequencer to run its program. Since this sequencer is required for left train DG operability, this condition resulted in an unanticipated entry into a TS shutdown action statement. The cause of this failure is currently unknown, pending the results of a vendor evaluation of a failed load sequencer component. Description : On August 3, 2017, the control room received alarm EK 1145, Sequencer Trouble, unexpectedly. The operators identified that the indication lights were not lit on the left channel load sequencer, MC -34L101; declared the associated DG inoperable; and entered the appropriate TS action statement. The failed sequencer was removed and replaced with a new module that was satisfactorily post -maintenance tested and the left train EDG was subsequently declared operable on August 4, 2017. The failed sequencer was sent to an on -site lab for further troubleshooting. No obvious visual signs of failure were identified and the electrolytic capacitors in the module all tested satisfactorily. The module was then bench tested using a test program, which identified that although it would power up, no program would run. The licensee completed an equipment failure evaluation to review the bench test data, along with information collected in the failure modes analysis, and determined that the direct cause of the failure was a memory fault within the sequencer module that caused the sequencer to lock -up and not run its program. A fault in the memory module, memory processing interface circuitry, or the executive module could have caused the sequencer to lock up. At the end of the inspection period, further examination by t he vendor was required and in progress to determine the exact initiating point of the fault. In addition to replacing the failed sequencer, the licensees immediate corrective actions included inspecting the right train load sequencer and completing the quarterly surveillance test to ensure proper operation; the results of which were satisfactory. A plant operating experience review was conducted and did not identify any prior memory failures on the load sequencers. Once the vendors evaluation is complete, the licensee plans to re-assess the failure mechanism and any additional corrective actions required. This item is considered unresolved, pending the inspectors review of the vendor analysis and any changes made to the equipment failure evaluation, to determine if this issue constitutes a performance deficiency and/or violation of NRC requirements. (URI 05000255/2017003 01, Left Train Emergency Diesel Generator Load Sequencer Failure ) |
| 05000255/FIN-2017003-02 | 2017Q3 | Palisades | Cause of 422/RPS Breaker Failure to Open | Introduction: The inspectors identified an URI associated with the failure mechanism of the 42 -2/RPS control rod clutch breaker failure to open. Specifically, at the end of the inspection period the licensee was working to understand the cause of the breaker failure and determine the actions required to address the failure mechanism. Description : On May 17, 2017, the licensee conducted a shutdown to complete emergent repairs to a leaking seal identified on control rod drive mechanism 40. In accordance with GOP 8, Power Reduction and Plant Shutdown to Mode 2 or Mode 3 525 F, the operators depressed the reactor trip pushbutton from the EC 06, reactor protection system panel. When the pushbutton was depressed, the reactor did not trip as expected. The operators successfully tripped the reactor using the reactor trip pushbutton on the EC 02, primary process and reactor controls console. The licensee identified that the 42 1/RPS breaker tripped as expected when the reactor trip pushbutton on the EC 06 panel was depressed, however, the 42 2/RPS breaker did not trip as expected. This resulted in the reactor trip not occurring as expected when the reactor trip pushbutton on the EC 06 panel was depressed as both breakers a re required to open to result in a reactor trip. The licensee performed troubleshooting activities to determine the cause of the 42 2/RPS breaker failure. The direct cause of the breaker failure was found to be the 42 2/RPS breaker undervoltage release mechanism failing to provide enough downward force to fully depress the trip plunger. This resulted in a physical failure of the breaker to open. At the end of the inspection period, the cause of this physical failure mode was unknown. The licensees equipment failure evaluation identified that it could be age- related degradation or a physical degradation of the breaker. As a corrective action, a failure analysis of the breaker was planned. Once the failure analysis i s complete, the licensee plans to re-assess the failure mechanism and determine any additional corrective actions that are required to address the issue. This item is considered unresolved, pending the inspectors review of the failure analysis and any changes made to the equipment failure evaluation, to determine if this issue constitutes a performance deficiency and/or violation of NRC requirements. (URI 05000255/2017003 02, Cause of 42 2/Reactor Protection System Breaker Failure to Open) |
| 05000255/FIN-2017003-04 | 2017Q3 | Palisades | Licensee-Identified Violation | The licensee identified a finding of very low safety significance (Green) and an associated NCV o f TS 5.7.2, which requires, in part, that each entryway into High Radiation Areas ( HRAs) with dose rates greater than 1.0 rem/hour at 30 centimeters from the radiation source or any surface penetrated by the radiation, but less than 500 rads/hour at 1 meter from the radiation source or from any surface penetrated by the radiation source shall be provided with a locked or continuously guarded door or gate that prevents unauthorized entry. Contrary to the above, on May 4, 2017, the licensee failed to lock or continuously guard an entryway into a HRA with dose rates greater than 1.0 rem/hour at 30 centimeters from the radiation source or any surface penetrated by the radiation, but less than 500 rads/hour at 1 meter from the radiation source or from any surface penetrated by the radiation source. Specifically, an entryway was left unguarded when the individual assigned to guard the entryway left the area prior to another guard being stationed. This issue was identified by a radiation protection technician who immediately stationed another guard. This issue was entered into the licensees CAP as CR PL 2017 02160. The failure to continuously guard the HRA entryway was a performance deficiency that was within the licensees ability to foresee and should have been prevented. The performance deficiency was 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 the adequate protect ion of worker health and safety from exposure to radiation. The finding was determined to be of very low safety significance (Green) because it did not involve as -low -as-reasonably -achievable planning or work controls, there was no overexposure or substantial potential for an overexposure, and the licensees ability to assess dose was not compromised. |
| 05000255/FIN-2017002-01 | 2017Q2 | Palisades | Inadequate Protection from Tornado Missiles Identified Due to Non- Conforming Design Conditions | A finding and an associated violation of 10 CFR, Part 50, Appendix B, Criterion III, Design Control, was identified based upon the lack of adequate tornado missile protection to the safety -related equipment listed above. The finding 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 Enforcement Guidance Memorandum 15 002, Revision 1, Enforcement Discretion for Tornado- Generated Missile Protection N on- Compliance, and can be found in ADAMS Accession No. ML16355A286. Because this finding and violation was identified during the discretionary period covered by Enforcement Guidance Memorandum 15002, Revision 1, Enforcement Discretion for Tornado Missile Protection Non-Compliance and because the licensee, prior to the expiration of the associated LCO, took initial compensatory measures that provided additional protection such that the likelihood of tonado missile effects were lessoned, followed by more comprehensive compensatory measures that w ere completed within approximately 60 days of issue discovery , and has final corrective actions planned, the NRC is exercising enforcement discretion by not issuing an enforcement action, as discussed in Section 1R15.2 of this report. |
| 05000293/FIN-2016011-05 | 2017Q1 | Pilgrim | Failure to Establish Corrective Actions to Address Scope of Procedure Quality Issues | The NRC team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, because Entergy implemented inadequate corrective actions to address the procedure quality issues identified in CR-PNP-2016-02058. Specifically, Entergy inappropriately limited their corrective actions to those procedures that increased integrated risk above normal, and did not include other types of safety-related procedures that did not meet their procedure quality standards and resulted in procedure quality being a problem area. Entergy entered this issue into their corrective action program for further evaluation as CR-PNP-2017-00400. The performance deficiency was more than minor because it affected the procedure quality attribute of the Mitigating Systems cornerstone, and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Entergy limited corrective actions to procedures that increased integrated risk above normal or trip sensitive and failed to include other procedures associated with safety-related components that reflected the broader population reviewed during the collective evaluation. The NRC team evaluated the finding using Exhibit 2, Mitigating Systems Screening Questions, of IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, and determined this finding did not affect the design or qualification of a mitigating structure, system, or component; represent a loss of system and/or function; involve an actual loss of function of at least a single train or two separate safety systems for greater than its technical specificationallowed outage time; or represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant. Therefore, the NRC team determined the finding was of very low safety significance (Green). The NRC team determined that this finding had a cross-cutting aspect related to Human Performance, Resources, because the leaders failed to ensure that personnel, equipment, procedures, and other resources are available and adequate to support nuclear safety. Specifically, based on available resources, Entergy chose to limit the scope of safety-related procedures being revised to only those that resulted in high integrated risk or were trip sensitive (H.1). |
| 05000293/FIN-2016011-07 | 2017Q1 | Pilgrim | Failure to Report Condition Prohibited by Technical Specifications and a Safety System Functional Failure | The NRC team identified a Severity Level IV non-cited violation of 10 CFR 50.73, Licensee Event Report System, associated with Entergys failure to submit a licensee event report within 60 days following discovery of an event meeting the reportability criteria. Specifically, on September 28, 2016, Entergy identified the A emergency diesel generator was inoperable. The NRC team determined that the condition was prohibited by technical specifications and the inoperability of the A emergency diesel generator existed for a period of time longer than allowed by Technical Specification 3.5.F, Core and Containment Cooling Systems. This was also reportable as a safety system functional failure. Entergy entered this issue into the corrective action program as CR-PNP-2016-09552. Because this performance deficiency had the potential to impact the NRCs ability to perform its regulatory function, the NRC team evaluated the performance deficiency using traditional enforcement. The violation was evaluated using Section 2.3.11 of the NRC Enforcement Policy, because the failure to submit a required licensee event report may impact the ability of the NRC to perform its regulatory oversight function. In accordance with Section 6.9.d, Example 9, of the NRC Enforcement Policy, this violation was determined to be a Severity Level IV non-cited violation. Because this violation involves the traditional enforcement process and does not have an underlying technical violation, the NRC team did not assign a cross-cutting aspect to this violation, in accordance with IMC 0612, Appendix B. |
| 05000293/FIN-2016011-13 | 2017Q1 | Pilgrim | Licensee-Identified Violation | 10 CFR 50.54(q)(2) requires, in part, that the licensee follow and maintain the effectiveness of an emergency plan to meet the planning standard of 10 CFR 50.47(b)(4). Specifically, the licensee was to maintain the necessary equipment to support the effectiveness of EALs. Contrary to these requirements, PNPS identified in CR-PNP-2016-01491 that on three past occasions (March 15 through August 8, 2012; September 4 through October 14, 2012; and June 4 through June 14, 2015) both trains of the H2O2 monitors and the Post-Accident Sampling System were unavailable to ensure the effectiveness of EAL 24, Deflagration concentrations exist inside PC, for the potential loss of the containment barrier within the Fission Product Barrier category of the EALs. This issue meets the criteria for very low safety significance (Green) because, due to other EALs, an appropriate emergency declaration could have been made in an accurate and timely manner. |
| 05000293/FIN-2016011-12 | 2017Q1 | Pilgrim | Licensee-Identified Violation | 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, and shall be accomplished in accordance with those structures, procedures, and drawings. Entergy procedure EN-DC-148, Vendor Manuals and Vendor Re-Contact Process, Revision 6, requires, in part, that the station update vendor manuals every three years. Contrary to this, in July 2016, PNPS determined through a self-assessment that they had 13 vendor manuals that had not been evaluated for changes within 3 years. The NRC team determined that this finding did not affect the design or qualification of a mitigating structure, system or component; did not represent a loss of a system and/or function; did not result in loss of a train or two safety systems greater than any technical specification allowed outage time; did not result from an actual loss of safety function; and did not involve loss of any external event mitigating system. Consequently, the NRC team determined that this performance deficiency screened as having very low safety significance (Green). PNPS documented this issue in their corrective action program as CR-PNP-2016-05115. |
| 05000293/FIN-2016011-11 | 2017Q1 | Pilgrim | Failure to Adequately Develop and Implement Targeted Performance Improvement Plans | The NRC team identified a Green finding because Entergy did not adequately develop and implement a CAPR of a root cause related to a Category A CR, as required by Entergy Procedure EN-LI-102, Corrective Action Program. Specifically, Entergy did not adequately develop and implement the Targeted Performance Improvement Plans, which were designated as a CAPR for the root cause for the Nuclear Safety Culture Fundamental Problem. Entergy documented this issue in the corrective action program for further evaluation as CR-PNP-2017-00406. The performance deficiency was more than minor because if left uncorrected, it could lead to a more significant safety concern. Specifically, inadequate implementation of the Targeted Performance Improvement Plans could result in recurrence of a culture in which leaders are not holding themselves and their subordinates accountable to high standards of performance, resulting in continuing performance issues at the station. The NRC team evaluated the finding using Exhibit 2, Mitigating Systems Screening Questions, of IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, and determined this finding did not affect the design or qualification of a mitigating structure, system, or component; represent a loss of system and/or function; involve an actual loss of function of at least a single train or two separate safety systems for greater than its technical specification-allowed outage time; or represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant. Therefore, the NRC team determined the finding was of very low safety significance (Green). This finding had a cross-cutting aspect in the area of Human Resources, Change Management, because leaders did not use a systematic process for evaluating and implementing change so that nuclear safety remains the overriding priority. In this case, PNPS leaders did not apply sufficient rigor in development and implementation of the Targeted Performance Improvement Plans such that they would be an adequate method to drive and sustain positive changes in the stations safety culture (H.3). |
| 05000293/FIN-2016011-10 | 2017Q1 | Pilgrim | Failure to Promptly Correct a Condition Adverse to Quality for the Residual Heat Removal System | The NRC team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, because Entergy did not take timely corrective action for a previously identified condition adverse to quality. Specifically, Entergy failed to adequately resolve, through repair or adequate evaluation, gasket leakage on the B residual heat removal heat exchanger, which resulted in continued degradation and leakage for the heat exchanger gasket. Entergy did not consider this leakage as a degraded condition, with the potential to impact both the operability of the residual heat removal system, and PNPSs licensing basis with regards to leakage of a closed loop system outside of containment. After the NRC team raised the issue, Entergy performed an operability determination that established a reasonable expectation of operability pending implementation of corrective actions. Entergy entered this issue into their corrective action program as CR-PNP-2016-09725. The performance deficiency was more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to correct identified gasket leakage resulted in continued degradation and leakage of the heat exchanger gasket. The NRC team evaluated the finding using Exhibit 2, Mitigating Systems Screening Questions, of IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, and determined this finding did not affect the design or qualification of a mitigating structure, system, or component; represent a loss of system and/or function; involve an actual loss of function of at least a single train or two separate safety systems for greater than its technical specification-allowed outage time; or represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant. Therefore, the NRC team determined the finding was of very low safety significance (Green). The finding had a cross-cutting aspect in Human Performance, Conservative Bias, because Entergy failed to use decision making practices that emphasize prudent choices over those that are simply allowable (H.14). |
| 05000293/FIN-2016011-09 | 2017Q1 | Pilgrim | Ineffective Corrective Actions to Address Conditions Adverse to Quality Regarding Components in Contact with or Close Proximity to the Drywell Liner | The NRC team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, associated with Entergys failure to correct a condition adverse to quality affecting safety-related equipment. Specifically, during a previous NRC inspection in August 2016, inspectors identified numerous locations in the drywell where non-seismic equipment was either in contact, or close proximity, with the drywell liner and had caused damage. Entergy initiated CRs and performed an operability evaluation for the identified issues. However, following a review of these CRs, the NRC team determined that Entergy failed to take corrective actions to address the condition adverse to quality. Entergy entered this issue into the corrective action program as CR-PNP-2016-09346 and CR-PNP-2016-09377 to perform an extent of condition review, secure the loose grating that had caused damage to the liner, and evaluate the need for a clearance criteria between components such as floor grating and support structures and the containment liner. The performance deficiency was more than minor because it was associated with the configuration control attribute of the Barrier Integrity cornerstone and affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Using IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 3, Barrier Integrity Screening Questions, the NRC team determined that this finding was of very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containment (valves, airlocks, etc.), containment isolation system (logic and instrumentation), and heat removal components. This finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, because the engineering evaluation of the degraded condition identified by the inspectors did not thoroughly evaluate the containment liner issues to ensure that resolutions address causes and extents of condition commensurate with their safety significance (P.2). |
| 05000293/FIN-2016011-08 | 2017Q1 | Pilgrim | Failure to Adequately Monitor the Performance of Maintenance Rule Scoped Components | The NRC team identified a Green non-cited violation of 10 CFR 50.65(a)(2), Requirements for monitoring the effectiveness of maintenance at nuclear power plants. Specifically, Entergy did not demonstrate that the performance of 18 maintenance rule scoped components was effectively controlled through the performance of appropriate preventive maintenance, and did not establish goals and monitoring in accordance with 10 CFR 50.65(a)(1). Entergys immediate corrective action was to initiate a CR to evaluate moving the affected systems to 10 CFR 50.65(a)(1) monitoring requirements. Entergy entered this issue in the corrective action program as CR-PNP-2017-00401. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, Entergy failed to demonstrate that the performance of the 18 maintenance rule scoped components was being effectively controlled through the performance of appropriate preventive maintenance which adversely impacts the reliability of those systems. The NRC team evaluated the finding using Exhibit 2, Mitigating Systems Screening Questions, of IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, and determined this finding did not affect the design or qualification of a mitigating structure, system, or component; represent a loss of system and/or function; involve an actual loss of function of at least a single train or two separate safety systems for greater than its technical specificationallowed outage time; or represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant. Therefore, the NRC team determined the finding was of very low safety significance (Green). The finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, in that Entergy failed to thoroughly evaluate and ensure that resolution of the identified issue, maintenance not being performed on maintenance rule scoped components, included reclassifying the components as necessary. Specifically, Entergy failed to demonstrate that the performance of Maintenance rule scoped components was effectively controlled through the performance of appropriate preventive maintenance, or through performance goals and monitoring. (P.2). |
| 05000293/FIN-2016011-04 | 2017Q1 | Pilgrim | Programmatic Issue with Implementation of the Operability Determination Process | The NRC team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings. Specifically, the NRC team identified a programmatic issue because in some cases, Entergy did not enter the operability determination process when appropriate, and, when the process was entered, did not adequately document the basis for operability, in accordance with Procedure ENOP-104, Operability Determination Process, Revision 11. In each of the examples discussed, though the basis for operability was not adequate, all components were determined to be operable upon further evaluation. Entergy entered this issue into their corrective action program as CR-PNP-2017-00626. The performance deficiency was more than minor because if left uncorrected, could lead to a more significant safety issue. Specifically, the failure to enter and document a basis for operability could lead to not recognizing inoperable safety-related equipment, and place the reactor at a higher risk of core damage in a design basis accident. The NRC team evaluated the finding using Exhibit 2, Mitigating Systems Screening Questions, of IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, and determined this finding did not affect the design or qualification of a mitigating structure, system, or component; represent a loss of system and/or function; involve an actual loss of function of at least a single train or two separate safety systems for greater than its technical specification-allowed outage time; or represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant. Therefore, the NRC team determined the finding was of very low safety significance (Green). This finding had a cross-cutting aspect in the area of Human Performance, Teamwork. Specifically, the operations and engineering departments did not demonstrate a strong sense of collaboration and cooperation with respect to holding each other accountable when performing operability determinations to ensure nuclear safety is maintained (H.4). |
| 05000293/FIN-2016011-03 | 2017Q1 | Pilgrim | Failure to Issue Appropriate Corrective Actions to Preclude Repetition for the Causes of the September 2016 Scram | The NRC team identified a Green finding because Entergy did not issue appropriate CAPRs in accordance with Entergy procedure EN-LI-102, Corrective Action Process, Revision 28. Specifically, Entergy did not issue adequate CAPRs associated with Root Cause 1 of the feedwater regulating valve failure in September 2016 that resulted in a manual scram. As a result of the NRC teams questions, Entergy issued procedure 1.13.2, Vendor and Technical Information Reviews, Revision 0, as continuous use to ensure that planners will always have the checklist in-hand when planning work to ensure that appropriate vendor technical information is always included in applicable work instructions. Entergy entered the NRC teams concerns in the corrective action program as CR-PNP-2017-00687 and CR-PNP-2017-00936. The performance deficiency was more than minor because it is associated with the equipment performance attribute of the Initiating Events cornerstone and if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Specifically, if left uncorrected, the performance deficiency could have the potential to result in repetition of a significant condition adverse to quality, loss of control of feedwater regulating valve 642A and a manual scram. The NRC team evaluated the finding using Exhibit 1, Initiating Events Screening Questions, of IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, and determined this finding did not cause a reactor trip or the loss of mitigation equipment relied upon to transition the plant from the onset of a trip to a stable shutdown condition. Therefore, the NRC team determined the finding was of very low safety significance (Green). The NRC team determined that the finding had a cross-cutting aspect in the area of Human Performance, Procedure Adherence, because individuals did not follow processes, procedures, and work instructions. Specifically, Entergy did not follow procedure EN-LI-102, which provides the station standards for crafting a corrective action and states, in part, that the corrective action descriptions must be worded to ensure that the adverse condition or cause/factor is addressed (H.8). |
| 05000293/FIN-2016011-01 | 2017Q1 | Pilgrim | Failure to Identify All Root Causes of a Significant Condition Adverse to Quality | The NRC team identified a Green non-cited violation of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion XVI, Corrective Action, because Entergy did not adequately determine all root causes associated with a significant condition adverse to quality related to the failure to identify, evaluate, and correct the A SRVs failure to open upon manual actuation during a plant cooldown on February 9, 2013. Specifically, Entergy did not establish adequate measures to assure that the cause of a significant condition adverse to quality, inadequate shift manager operability determination rigor and its associated causes, were adequately determined and corrective action taken to preclude repetition. Entergys immediate corrective actions included planning to conduct operations management face-to-face conversations with shift manager qualified individuals to reinforce the shift managers responsibility for operability and functionality determination accuracy and rigor. Entergy entered this issue into the corrective action program as CRPNP-2017-00363 and CR-PNP-2017-00828. The performance deficiency was more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Specifically, if left uncorrected, the performance deficiency could have the potential to result in repetition of a failure to identify, evaluate, and correct an SRVs failure to open or a similar significant condition adverse to quality. The NRC team evaluated the finding using Exhibit 2, Mitigating Systems Screening Questions, of IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, and determined this finding did not affect the design or qualification of a mitigating structure, system, or component; represent a loss of system and/or function; involve an actual loss of function of at least a single train or two separate safety systems for greater than its technical specification-allowed outage time; or represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant. Therefore, the NRC team determined the finding was of very low safety significance (Green). The NRC team determined that the finding had a cross-cutting aspect in the area of Human Performance, Avoid Complacency, because individuals did not recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Specifically, Entergy incorrectly assumed that CR-PNP-2013-00825 contained inadequate information to determine that the A SRV had not opened, and this assumption ultimately impacted the root cause results documented in CR-PNP-2016-01621 (H.12). |
| 05000293/FIN-2016011-02 | 2017Q1 | Pilgrim | Failure to Establish Corrective Actions to Preclude Repetition of a Significant Condition Adverse to Quality | The NRC team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, because Entergy did not implement CAPRs for a significant condition adverse to quality identified in root cause evaluation CR-PNP-2016-00716, Implementation of the Corrective Action Program, Revision 2. Specifically, the team identified that CAPRs for Entergys continued weaknesses in the implementation of the corrective action program were inadequate. Entergy entered this issue into their corrective action program for further evaluation as CR-PNP-2017-00053, CR-PNP-2017-00410, and CR-PNP-2017-01134. The performance deficiency was more than minor because if left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, the failure to preclude repetition of this significant condition adverse to quality could result in continuing weaknesses in implementation of the corrective action program, which was designated as a fundamental problem, and thus a contributing factor for PNPS Column 4 performance. Additionally, weaknesses with corrective action program implementation could result in equipment issues where operability is not maintained. The NRC team evaluated the finding using Exhibit 2, Mitigating Systems Screening Questions, of IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, and determined this finding did not affect the design or qualification of a mitigating structure, system, or component; represent a loss of system and/or function; involve an actual loss of function of at least a single train or two separate safety systems for greater than its technical specificationallowed outage time; or represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant. Therefore, the NRC team determined the finding was of very low safety significance (Green). The NRC team determined that the finding had a cross-cutting aspect in the area of Human Performance, Procedure Adherence, because individuals did not follow processes, procedures, and work instructions. Specifically, Entergy did not follow procedure EN-LI-102, which provides the station standards for crafting a corrective action and states, in part, that the corrective action descriptions must be worded to ensure that the adverse condition or cause/factor is addressed (H.8). |
| 05000293/FIN-2016011-06 | 2017Q1 | Pilgrim | Design Change Not Appropriately Reviewed by Entergy | The NRC team identified a preliminary greater than Green finding and apparent violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, associated with Entergys failure to ensure that design changes were subject to design control measures commensurate with those applied to the original design and were approved by the designated responsible organization. Specifically, Entergy received a new style right angle drive for the A emergency diesel generator radiator blower fan from a vendor but failed to adequately review the differences in the design of the drives to identify potential new failure mechanisms for the part or the need for related preventive measures. Entergy entered this issue into the corrective action program as CR-PNP-2016-07443. The performance deficiency was more than minor because it was associated with the design control attribute of the Mitigating Systems cornerstone, and affected the associated cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, the team screened the finding for safety significance and determined that a detailed risk evaluation was required based on the A emergency diesel generator being inoperable for greater than the technical specification allowed outage time. Region I senior reactor analysts performed a detailed risk evaluation. The finding was preliminarily determined to be of greater than very low safety significance (greater than Green). The risk important sequences were dominated by external fire risk. Specifically, a postulated fire in the B 4 kilovolt (KV) switchgear room with a consequential loss of the unit auxiliary generator power supply, non-recoverable loss of off-site power (LOOP) to both safety buses A5 and A6, loss of the B emergency diesel generator with the conditional failure of the A emergency diesel generator, along with the loss of bus A8 feed (from the shutdown transformer or station blackout (SBO) diesel generator) to safety buses A5 and A6. The internal event risk was dominated by weather related LOOPs, failure of the A emergency diesel generator, with failure of the B emergency diesel generator and SBO diesel generator to run, along with failure to recover offsite power or the emergency diesel generators. See Attachment 1, A Emergency Diesel Generator Cooling Water System Degradation Detailed Risk Evaluation, for a detailed review of the quantitative criteria considered in the preliminary risk determination. The NRC team did not assign a cross-cutting aspect to this finding because the performance deficiency occurred in May 2000. Entergys program has undergone changes since May 2000, and the NRC team did not identify any recent examples of this performance deficiency. Other aspects of Entergys performance related to this issue are further discussed in Sections 5.10.3 and 6.3.4. |
| 05000255/FIN-2017001-01 | 2017Q1 | Palisades | Licensee-Identified Violation | The licensee Identified a finding of very low safety significance (Green) and an associated NCV of 10 CFR 50, Appendix R, Section III.G.2, which requires, in part, that where cables or equipment of redundant trains of systems necessary to achieve and maintain hot shut down conditions are located within the same fire area outside of primary containment, one means of ensuring that one of the redundant trains is free of fire damage shall be provided. Contrary to the above, as of October 1, 2010, the licensee failed to ensure that one of the redundant trains was free of fire damage in areas where cables or equipment of redundant trains of systems necessary to achieve and maintain hot shutdown conditions are located within the same fire area outside of primary containment. Specifically, the licensee failed to analyze a fire scenario in the 1C switch gear room, screen- house room, and component cooling water pump room that could potentially damage the control cable before the load cable, and therefor e result in the loss of safety -related 2400 volt alternating current (VAC) bus 1C and/or 1D, with subsequent loss of equipment credited for Appendix R compliance to support safe shutdown in the event of such a fire. The licensees failure to analyze an Appendix R fire scenario for the three fire areas described above w as a performance deficiency . 21 The performance deficiency was more- than- minor because it was associated with the Mitigating Systems cornerstone attribute of Protection Against External Factors (Fire) 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 was determined to be of very low safety significance (Green) because it did not impact the licensees ability to reach hot shutdown because operator manual actions would have allowed operators to shut down the plant following a fire. The licensee identified this issue during the transition to NFPA 805, entered the issue into their CAP as CR PLP 2010 04255, and implemented compensatory measures, including fire watches. The violation was not willful and routine licensee efforts, such as normal surveillance or quality assurance activities, were not likely to have previously identified the violation due to the specific sequence of fire cable damage required for such an Appendix R fire scenario. As a result, the inspectors concluded that the violation met all four criteria for exercising enforcement discretion established by Section 9.1 of the NRCs Enforcement Policy Regarding Enforcement Discretion for Certain Fire Protection Issues; therefore, the NRC is exercising enforcement discretion to not cite this violation |
| 05000255/FIN-2016004-01 | 2016Q4 | Palisades | Failure to Have Appropriate Controls in Place for Combustible Materials | Green. A 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) was identified by the inspectors for the licensees failure to appropriately implement the requirements of procedure ENDC161, Control of Combustibles. Specifically, between January 1, 2016 and October 22, 2016, the inspectors identified several examples of the licensees failure to have appropriate controls in place for the storage of combustible materials in excess of the limits required for those respective areas without a completed transient combustible evaluation (TCE). Also, on several occasions from October 19, 2016 to October 22, 2016, the required compensatory actions for a TCE related to the dry fuel storage cask transporter vehicle were not appropriately implemented as required by procedure ENDC161. The licensee entered these issues in their corrective action program (CAP) as condition reports (CRs) CRPLP201603633, CRPLP201605148, and CRPLP20160564. Corrective actions for these issues included completing the required TCEs, ensuring the combustible materials in the areas were addressed by the combustible loading calculations, and ensuring appropriate compensatory measures were implemented. The issue was determined to be more than minor in accordance with IMC 0612, Appendix B, Issue Screening, because it was associated with the Protection Against External Factors attribute, in the area of Fire, 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). Specifically, transient combustible materials without required TCEs were stored in the charging pump cubicles and in the refueling and spent fuel pool areas. The finding screened as having very low safety significance (Green) in accordance with IMC 0609, Appendix F, Fire Protection Significance Determination Process, since none of the stored materials were self-igniting, low flashpoint liquids, or heat sources and was therefore assigned a Low degradation rating. The finding had a cross-cutting aspect of Training in the Human Performance cross-cutting area due to the common element of a lack of knowledge of the individuals with the control of combustibles process and understanding their roles in that process (H.9). |
| 05000255/FIN-2016004-02 | 2016Q4 | Palisades | Failure to Correct an Adverse Condition Associated with Diesel Generator Load Sequencer Module | Green. A finding of very low safety significance and an associated NCV of 10 CFR, Part 50, Appendix B, Criterion XVI, Corrective Action, was self-revealed for the licensees failure to promptly correct a condition adverse to quality. Specifically, the licensee failed to correct an adverse condition associated with the emergency diesel generator (DG) load sequencer and power supply module as revealed when the electrolytic capacitor failed two days after installation. The 12 DG was declared inoperable, the licensee replaced the failed module, and an equipment apparent cause evaluation was completed for the equipment failure. An internal operating experience review revealed that a similar issue occurred in 2005 and corrective actions to address that failure, which included establishing shelf life and age requirements for electrolytic capacitors that were part of power supply modules, were not applied to this module. The licensee entered this issue into their CAP as CRPLP201603260. The issue was determined to be more than minor in accordance with IMC 0612, Appendix B, because the performance deficiency was associated with the Equipment Performance 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). Specifically, the licensee failed to correct a condition adverse to quality, which rendered the 12 DG inoperable. This condition would have prevented the DG from automatically starting and loading on the prescribed signal. The finding was screened in accordance with IMC 0609, Appendix A, and was determined to have very low safety significance (Green) based on answering No to all the screening questions under the Mitigating Structure, System and Components, and Functionality section. The inspectors concluded that the corrective actions for the adverse condition of the aging electrolytic capacitors should have been implemented greater than three years ago, so the finding was not reflective of current licensee performance. Therefore, no cross-cutting aspect was identified. |
| 05000255/FIN-2016004-03 | 2016Q4 | Palisades | Failure to Translate Design Analysis Stack-up Configuration into Specifications, Drawings, Procedures, and Instructions | Green. A finding of very low safety significance and an associated NCV of 10 CFR, Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors for the licensees failure to establish measures to assure that the applicable regulatory requirements and the design basis were correctly translated into specifications, drawings, procedures, and instructions. Specifically, the licensee failed to provide instructions in procedures to construct the spent fuel dry cask loading stack-up, in the safety-related auxiliary building, in the configuration that had been analyzed for in the stack-up seismic design basis calculation. In addition, the licensee failed to provide instructions in revised procedures to construct the stack-up without certain gaps as 4 specified in the stack-up seismic design basis document. The licensee documented these issues in their CAP as CRPLP201600646, CRPLP201601308, CRPLP201601558, CRPLP201604497, and CRPLP201604826; revised the stack-up seismic analysis to address the identified issues; and translated the analyzed stack-up design configuration into stack-up installation procedures prior to performing stack-up operations with spent nuclear fuel in the multi-purpose canister. The issue was determined to be more than minor in accordance with IMC 0612, Appendix B, Issue Screening, because it was associated with the Design Control attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the performance deficiency resulted in a stack-up configuration that did not ensure stack-up dynamic stability or Auxiliary Building structural integrity to maintain radiological barrier functionality during a design basis seismic event. The finding screened as having very low safety significance (Green) because it did not result in the loss of operability or functionality of the Auxiliary Building. The finding had a cross-cutting aspect of Field Presence in the Human Performance cross-cutting area, because licensee senior managers failed to ensure effective supervisory and management oversight of contractor activities related to the seismic analysis and installation of the stack-up configuration (H.2). |
| 05000255/FIN-2016003-01 | 2016Q3 | Palisades | Failure to Appropriately Select and Review for Suitability of Application the Control Switch and Circuit Design of the Engineered Safeguards Room Cooler Fans | A self-revealed finding of very low safety significance and an associated non-cited violation (NCV) of Title 10 of the Code of Federal Regulations, Part 50, Appendix B, Criterion III, Design Control, was identified for the failure to appropriately select and review for suitability of application the control switch and circuit design of the engineered safeguards room cooler fans. Specifically, on July 27, 2016, when the licensee was conducting troubleshooting activities for the tripping of engineered safeguards room cooler fan V27B, it was revealed that the control switch design was break before make and as the hand switch was transitioned from one position to the next, the supply voltage and the motor became out of phase and caused an overcurrent trip of the breaker. This resulted in an unplanned entry into a 72 hour limiting condition for operation (LCO) for the right train of the emergency core cooling system (ECCS). In the apparent cause evaluation (ACE) for this issue, the licensee determined that the contributing cause had not previously addressed this particular failure mode (i.e. the control switch and circuit design) when similar overcurrent events occurred in the past. Prior corrective actions included adding guidance to system operating procedures to pause between hand switch movements and replacing other components within those systems. These actions were not successful in eliminating this failure mode. The licensee documented the issue in their CAP, planned to revise the control circuit and switch design, and added specific procedural steps on how to operate these fans until the design change was implemented. The finding was more than minor in accordance with IMC 0612, Appendix B, because it was associated with the Mitigating Systems Cornerstone attribute of Equipment Reliability and adversely impacted the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, as a result of the overcurrent trip of its breaker, V27B was declared non-functional and unavailable and the equipment in the room it cooled was declared inoperable, which included the A high pressure safety injection (HPSI) pump and the A containment spray (CS) pump. This led to an unplanned entry into a 72 hour LCO for the right train of ECCS. The finding had a cross-cutting aspect in the area of Problem Identification and Resolution and was related to the cross-cutting component of Evaluation, which required that the licensee thoroughly evaluate issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. As discussed above, in the ACE for this issue the licensee determined that the corrective actions associated with the identified contributing cause following similar overcurrent events that occurred in the past had not addressed or been successful in eliminating this failure mode (PI.2). |
| 05000255/FIN-2016003-02 | 2016Q3 | Palisades | Hourly Fire Tour Discrepancies | The inspectors identified an unresolved item (URI) related to discrepancies found during fire tour daily log sheet and corresponding badge record reviews. Specifically, the NRC is in the process of reviewing the licensees evaluation of the root and contributing causes of the issue, as well as the corrective actions to prevent recurrence. Also, the NRC will verify that the licensees actions taken to address the issue are sustainable. On May 24 and 25, 2016, while the inspectors were observing a maintenance activity on a service water pump in the screenhouse, they noted that hourly fire tours were not being conducted consistently by security personnel. The inspectors requested plant room badging records and copies of the hourly fire tour daily log sheets from the licensee for hourly fire tours completed on May 24 and 25, 2016. The inspectors identified that some areas on the fire tour log sheets were annotated as complete, yet there were no corresponding badge records for these areas. The inspectors requested additional fire tour daily log sheets and badge records for May 31 and June 1, 2016 for an extent of condition review. Additional issues were identified with the fire tour log sheets not corresponding with badge records for certain plant areas required to be covered by the hourly fire tours. On June 8, 2016, the inspectors discussed these discrepancies with the licensee. The licensee entered this issue into the CAP and promptly began an extent of condition review of the fire tour daily log sheets and plant room badging records for the period of March 1, 2016 through June 8, 2016. The condition report included actions to conduct a root cause evaluation to determine the root and contributing causes of the discrepancies identified in the fire tour and badging records and formulating corrective actions to prevent recurrence. The licensees immediate interim corrective actions included direct supervisor observation of all hourly fire tours being conducted, newly formatted fire tour log sheets with additional detail added, and re-training of personnel conducting the tours on the requirements and expectations for completion of the activity. Pending NRC review of the licensees evaluation of the issue, subsequent corrective actions to prevent recurrence, and verification that the actions are sustainable, this issue is unresolved. |
| 05000255/FIN-2016001-03 | 2016Q1 | Palisades | Failure to Meet the Minimum Staffing Requirements of the Fire Brigade | An 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-2016001-04 | 2016Q1 | Palisades | Licensee-Identified Violation | Title 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-2016001-05 | 2016Q1 | Palisades | Licensee-Identified Violation | TS 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). |
| 05000528/FIN-2016008-01 | 2016Q1 | Palo Verde | Operations Department Failure to Document Conditions Adverse to Quality in Condition Reports | The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for the licensees failure to document conditions adverse to quality in the corrective action program. Previous similar failures to initiate condition reports led to, or contributed to, two significant conditions adverse to quality over the last 15 months. The failure of the operations department to document identified conditions adverse to quality in condition reports, as required by Procedure 01DP-0AP12, Condition Reporting Process, Revision 23, was a performance deficiency. This 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, on two other occasions since January 2015, failures by operations personnel to write condition reports for equipment-related problems resulted in or contributed to significant conditions adverse to quality. This performance deficiency demonstrated a continued gap within Palo Verde Nuclear Generation Stations operations department in understanding condition report initiation criteria. This performance deficiency is associated with the mitigating systems cornerstone. Using NRC Inspection Manual Chapter 0609, Appendix A, the team determined that this finding was of very low safety significance (Green) because it did not affect the operability or functionality of a mitigating structure, system, or component. This finding has a resolution cross-cutting aspect in the area of problem identification and resolution because the licensee failed to take effective corrective actions to address issues in a timely manner commensurate with their safety significance (P.3). |
| 05000528/FIN-2016008-02 | 2016Q1 | Palo Verde | Licensee-Identified Violation | Title of 10 CFR 50.55a(f)(4), requires, in part, that pumps and valves classified as ASME Code Class 1, 2, or 3 must meet the inservice test requirements set forth in the ASME Operation and Maintenance (OM) Code and addenda to the extent practical within the limitations of design, geometry, and materials of construction of the components. The inservice testing program is incorporated into the Palo Verde Nuclear Generation Station licensing basis under Technical Specification 5.5.8 and governed by the procedures controlled under that specification. ASME OM Code Case OMN-1 was adopted by Palo Verde Nuclear Generating Station per Valve Relief Request number 1, and approved by the NRC as an alternative for performing Code-required valve and pump testing for the second and third 10-year testing intervals (January 1998-2018). ASME OMN-1, Section 3.3.1(b) requires that, if insufficient data exist to determine the inservice test frequency...then (motor operated valve) MOV inservice testing shall be conducted every two refueling cycles or three years until sufficient data exist to determine a more appropriate test frequency. Palo Verde Nuclear Generating Station Procedure 73DP-9ZZ12, Motor Operated Valve Program, Section 4.5.4.5, and Appendix H, define when sufficient test data exists to justify increasing test frequencies beyond 3 years. This criteria includes completing at least two complete diagnostic testing cycles constituting a baseline pre-service test and two subsequent as-found tests. These testing requirements are invoked after complete replacement of the valve, installation of a new valve, or major maintenance, which could substantially change the valve/actuator performance. Contrary to the requirements listed above, the licensee failed to perform Code-required testing for a total of 17 valves between 2008 and 2016. The licensee identified an issue in May 2015 with the testing frequency of five valves after a modification installed new motor operated valves in the charging system. An extent of condition was performed and 11 additional valves were identified as being noncompliant. An engineering evaluation was performed to assess and manage the risk of not completing the required ASME testing per Technical Specification Surveillance Requirement 3.0.3. A prompt operability determination was also performed to provide reasonable assurance of operability until the valves could be tested again. In January 2016, an additional valve was identified as being non-compliant and a separate operability evaluation was completed to provide reasonable assurance that the valve would still perform its function. There are currently seven valves that are still in non-compliance with the Coderequired testing frequency; all other valves have been tested satisfactorily and are now in compliance. Those still requiring testing are scheduled during their respective next available system windows. This violation is of very low safety significance (Green) because the non-conforming valves were determined to have reasonable assurance of operability. The licensee entered the condition into its corrective action program and initiated corrective actions to restore compliance under Condition Report 15-02470. |
| 05000255/FIN-2016001-01 | 2016Q1 | Palisades | Design Review of Modification to Track Alley Wall for Dry Fuel Storage Activities | The 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-2016001-02 | 2016Q1 | Palisades | Movement of Radioactive Material Results in an Unposted and Un-Barricaded High-Radiation Area | A 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). |
| 05000255/FIN-2015004-02 | 2015Q4 | Palisades | Failure to Identify Components Required to be Covered by the Quality Assurance Program | The inspectors identified a finding of very-low safety significance, and an associated NCV of 10 CFR, Part 50, Appendix B, Criterion II, Quality Assurance Program, for the licensees failure to identify all component cooling water (CCW) structures, systems, and components (SSC), which were required to be covered by the Quality Assurance Program (i.e., be safety-related). As a result, the licensee incorrectly credited nonsafety-related CCW components to remain functional during and following a design basis event (DBE). The licensee entered this finding into their CAP and, after performing operability determinations, concluded the system would still be capable of performing its function. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of equipment performance, and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding screened as having very-low safety significance (Green) because, although it was a deficiency affecting the design or qualification of a mitigating SSC, the SSC maintained its operability. The inspectors did not identify a cross-cutting aspect associated with this finding because it was determined not to be representative of current performance. |
| 05000255/FIN-2015004-01 | 2015Q4 | Palisades | Inadequate Dye Penetrant Examination of Pipe Lug Welds | The inspectors identified a finding of very-low safety significance (Green), and an associated NCV of Title 10, Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion IX, Control of Special Processes, for the licensees failure to perform a dye penetrant (PT) examination of the Safety Injection System (SIS) pipe lug welds in accordance with the American Society of Mechanical Engineers (ASME) Code Section XI requirements. The licensee entered this issue into the Corrective Action Program (CAP) as CR-PLP-2015-04191, repeated the PT examination of the affected SIS lug welds to meet the full extent of coverage required by the ASME Code, repeated examinations of other welds conducted by the PT examiner during the outage, and removed the PT examiner from further weld examination activities. This performance deficiency was determined to be more than minor because, if left uncorrected, the failure to perform a PT examination in accordance with the ASME Code requirements could result in acceptance and return to service of a component with an undetected crack that would increase the possibility of pipe leakage or failure. In addition, the failure to perform a PT examination in accordance with the ASME Code adversely affected the Mitigating System Cornerstone attribute of Equipment Performance, because it could result in failure to detect cracks in pipe welds, which would reduce the availability and reliability of the SIS mitigating system. The inspectors evaluated the finding in accordance with IMC 0609, Appendix A, The SDP for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, and answered yes to screening question number 1. Although this finding adversely affected the design or qualification of the SIS pipe lugs, the finding screened as very-low safety significance (Green), because it did not result in the loss of operability or functionality of the affected SIS pipe segment. This finding had a cross-cutting aspect in the Field Presence component of the Human Performance cross-cutting area. Specifically, licensee leaders were not observed in the work areas of the plant to coach and reinforce standards or expectations for the licensees vendor staff to ensure deviation from standards and expectations were promptly corrected (H.2). |
| 05000255/FIN-2015004-03 | 2015Q4 | Palisades | Failure to Provide Bases to Determine Changes Did Not Involve Unreviewed Safety Questions | The inspectors identified a Severity Level (SL) IV, NCV of 10 CFR, Part 50, Section 59, Changes, Tests, and Experiments, for the licensees failure to maintain records of written safety evaluations, which provide the bases for concluding the nonsafety-related portions of the CCW system inside containment could be credited to perform their function during and following a DBE, and that the change would not result in an unreviewed safety question. The licensee entered this issue into their CAP and, after performing operability determinations, concluded the system would still be capable of performing its function. The violation was determined to be more than minor because the inspectors could not reasonably determine that the changes would not have ultimately required NRC prior approval. The violation was categorized as a SL IV in accordance with Section 6.1.d.2 of the NRC Enforcement Policy because the resulting changes were evaluated by the SDP as having very-low safety significance (i.e., green finding). The resulting changes, the violations underlying technical concerns, impacted the Mitigating Systems cornerstone, and were evaluated separately as the Green finding with the associated 10 CFR, Part 50, Appendix B, Criterion II, NCV discussed above. The inspectors did not identify a cross-cutting aspect because cross-cutting aspects are not assigned to traditional enforcement violations. |
| 05000255/FIN-2015004-04 | 2015Q4 | Palisades | Failure to Perform a Required 50.59 Evaluation for Declassification of the CVCS | The inspectors identified a SL IV, NCV of 10 CFR, Part 50.59, Changes, Tests, and Experiments, and an associated finding of very-low safety significance (Green) for the licensees failure to maintain a record of the declassification of the Chemical Volume and Control System (CVCS) from safety-related to nonsafety-related, which includes a written evaluation that provides the bases for the determination that the change did not require a license amendment. The licensee entered this issue into their CAP, and after a review of the system, determined there was reasonable assurance that it could perform its function. The inspectors determined the underlying technical concern was a performance deficiency associated with the Mitigating Systems cornerstone that was more than minor because, if left uncorrected, would become a more significant safety concern. The underlying technical concern screened as a finding with very-low safety significance (Green) because, although it affected the design or qualification of the CVCS, it did not result in the loss of functionality of the CVCS. The violation was determined to be more than minor because the inspectors could not reasonably determine that the changes would not have ultimately required NRC prior approval. The violation was categorized as a SL IV in accordance with Section 6.1.d.2 of the NRC Enforcement Policy because the changes were evaluated by the SDP, described above, as having very-low safety significance (i.e., Green finding). The inspectors did not identify a cross-cutting aspect associated with the finding because the finding was not representative of current performance. |
| 05000255/FIN-2015004-05 | 2015Q4 | Palisades | Licensee-Identified Violation | Title 10 CFR 50.65(a)(1), requires, in part, that the holders of an operating license shall monitor the performance or condition of structures, systems, and components (SSCs), against licensee-established goals, in a manner sufficient to provide reasonable assurance that these SSCs, as defined in 10 CFR 50.65(b), are capable of fulfilling their intended functions. Title10 CFR 50.65(a)(2) states that monitoring as specified in 50.65(a)(1) is not required, where it has been demonstrated that the performance or condition of a SSC is being effectively controlled through the performance of appropriate preventive maintenance, such that the SSC remains capable of performing its intended function. Contrary to the above, as identified after the November 14, 2014, TDAFW pump trip, the licensee failed to demonstrate the performance or condition of the safety-related auxiliary feedwater system steam traps had been effectively controlled through the performance of appropriate preventive maintenance. Specifically, some of the safety-related steam traps, one relief valve, and one check valve associated with the steam supply piping of the turbine-driven AFW system were inappropriately classified in the maintenance rule program, resulting in inadequate and/or untimely maintenance being performed on these components, which probably contributed to the overspeed trip event. The licensee found 3 steam traps and one relief valve classified as non-critical components that were reclassified as high critical components and one steam trap and one check valve classified as run-to-failure components that were reclassified as high critical components. Some of these components also had no preventive maintenance (PM) strategies or ones that were not the correct frequency based on the component classification. The licensee identified this issue while conducting the equipment apparent cause evaluation for the overspeed trip event and documented actions to correct the issue in CR-PLP-2014-5477. The licensee performed inspections of all the steam traps required for the TDAFW pump operation and identified some issues with steam cutting, foreign material exclusion in the traps, and incomplete seat contact. These issues were corrected and PM changes have been made for all the system components mentioned above. The inspectors determined that the inconsistent equipment classifications and ineffective preventive maintenance strategy for the safety-related steam traps in the turbine-driven auxiliary feedwater system is considered a performance deficiency. The performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely impacted the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events. Specifically, the licensee identified that the degraded condition of the moisture removal system could have led to excess condensate being present in the steam supply line which had the potential to adversely affect the operation of the turbine for the TDAFW pump, contributing to the overspeed trip event. The inspectors screened the issue using IMC 0609, Appendix A, The SDP for Findings at Power, Exhibit 2, Mitigating Systems Screening Questions, and answered Yes to the question of does this finding represent a loss of system and/or function? This trip of the TDAFW pump on overspeed was evaluated as a failure that impacted the ability of the AFW system to provide the specific function, which could only be accomplished by this train, of decay heat removal via steaming of the A Steam Generator. The turbine-driven AFW pump was also determined to not be in a condition to meet performance requirements defined by the probabilistic risk assessment success criteria, which for AFW is a 24 hour mission time. Therefore, the issue was screened further in a detailed risk evaluation. A Region III Senior Reactor Analyst performed a detailed risk evaluation using the NRCs Standardized Plant Analysis Risk Model for Palisades, Revision 8.20. The SRA assumed the turbine driven AFW pump was unavailable to perform its function for a period of 3 days because the pump was successfully tested and returned to service on November 16, 2014. Given the short exposure period, the calculated delta core delta frequency was less than 1.0E-7/yr. As a result of the low calculated delta core delta frequency, no additional analysis of external event risk contribution or large early release risk contribution was necessary. The dominant core damage sequence was a station blackout followed by the failure of the turbine driven AFW pump and the failure to recover onsite or offsite power. Therefore, the finding screened as very low safety significance (Green). |
| 05000255/FIN-2015003-01 | 2015Q3 | Palisades | Failure to Justify Continued Service of Safety-Related Electrolytic Capacitors Installed Beyond Their Service Life | An 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, Appendix B, Criterion III, Design Control, was identified for the failure to justify continued service of safety-related electrolytic capacitors that were installed beyond their recommended service life associated with the safety-related containment floor level indicating transmitters (LITs). Specifically, on June 21, 2015, containment floor LIT LIT0446B and LIT0446A did not satisfy the acceptance criteria of the technical specification surveillance monthly channel checks and LIT0446B was declared inoperable. Further troubleshooting identified a failure of the electrolytic capacitor within the transmitters converter module and that this failure was most likely due to age since the transmitter had been in service for greater than its recommended service life. In addition to entering this issue into their Corrective Action Program (CAP) as CRPLP201504972, the licensee replaced the failed components and planned to develop a replacement schedule for non-critical, safety-related electrolytic capacitors. The performance deficiency was determined to be more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the 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 answering No to all of the screening questions in the Mitigating Structures, Systems, and Components (SSCs) and Functionality section of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 1, Mitigating Systems Screening Questions. The finding had a cross-cutting aspect of Operating Experience in the Problem Identification and Resolution cross-cutting area because the licensee did not effectively and thoroughly evaluate and implement relevant industry operating experience and guidance for age-related electrolytic capacitor degradation. |
| 05000255/FIN-2015003-02 | 2015Q3 | Palisades | Failure to Establish, Implement, and Maintain the Offsite Dose Calculation Manual | A finding of very low safety significance and an associated NCV of Technical Specification (TS) 5.5.1, Offsite Dose Calculation Manual, was identified for the failure to establish, implement, and maintain the Offsite Dose Calculation Manual (ODCM) relative to dose calculation parameters. Specifically, the licensee failed to modify the parameters used in public radiation calculations when changes in the use of unrestricted areas were identified. As a result, the quarterly and annual doses that were calculated every 31 days, as required by the ODCM, were incorrect and non-conservative. In addition to entering this issue into their CAP as CRPLP20152972, the licensee recalculated the dose using the correct calculation parameters. The performance deficiency was determined to be more than minor 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. The finding was determined to be of very low safety significance in accordance with IMC 0609, Appendix D, Public Radiation Safety Significance Determination Process, because the issue did not represent a significant deficiency in evaluating a planned or unplanned effluent release since the resulting dose was not grossly underestimated. The finding had a cross-cutting aspect of Training in the Human Performance cross-cutting area because the licensee did not ensure adequate knowledge transfer to maintain a knowledgeable, technically competent workforce. |
| 05000255/FIN-2015012-02 | 2015Q3 | Palisades | Operability Evaluation Not Performed in Accordance with Station Procedure | An NRC-identified finding of very low safety significance and an associated NCV of Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, was identified for the licensees failure to adhere to the site procedure for performing operability determinations during the evaluation of a nonconforming condition associated with nine primary coolant system (PCS) welds susceptible to primary water stress corrosion cracking (PWSCC). The licensees corrective actions for this finding included completion of an operability determination in accordance with the site operability procedure to include a new analysis which demonstrated the AMSE Code acceptance criteria would continue to be met for the affected welds during the remainder of the operating cycle. The licensee entered the failure to comply with the operability procedure into the CAP (CR-PLP-2015-03434). This finding was determined to be more than minor because it was similar to the not minor if aspect of Example 3j in IMC 0612, Appendix E, Example of Minor Issues, because the errors in operability evaluation CA-1 of CR-PLP-2015-01239 resulted in a condition in which there was a reasonable doubt on the operability of the systems and components that were the subject of the evaluation and dissimilar from the minor because aspect of this example since the impact of the errors on the operability evaluation was not minimal. In addition, the performance deficiency was determined to be more than minor because it was associated with the Initiating Event Cornerstone attribute of Equipment Performance and adversely affected the Cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions. The inspectors evaluated the finding in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings, Table 3, for the Initiating Events Cornerstone and IMC 0609, Appendix A, The SDP for Findings At-Power. Because the licensee was able to demonstrate operability of the nine PCS welds susceptible to PWSCC, the inspectors answered No to questions A.1 and A.2, of Exhibit 1, Initiating Events Screening Questions, identified in Appendix A of IMC 609 and, as a result, the finding screened as having very low safety significance (Green). This finding has a crosscutting aspect in Evaluation for the Problem Identification and Resolution cross-cutting area since the licensee failed to thoroughly evaluate the impact on operability of a nonconforming condition associated with nine PCS welds susceptible to PWSCC. |