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| Q | Significance | CCA | Identified by | Title | Description | |
|---|---|---|---|---|---|---|
| 05000255/FIN-2018411-01 | 2018Q3 | Green | Licensee-identified | Licensee-Identified Violation | ||
| 05000255/FIN-2018411-02 | 2018Q3 | Green | Licensee-identified | Licensee-Identified Violation | ||
| 05000255/FIN-2018003-01 | 2018Q3 | Green | H.12 | NRC identified | Wire Not Landed on Safety Injection Initiation Relay Circuit | The inspectors identified a Green finding and an associated non-cited violation (NCV)of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to accomplish an activity affecting quality in accordance with the implementing procedure. Specifically, only one of two required wires was landed on terminal 13 of relay SIS2 in the right channel of the safety injection system (SIS) actuation logic following surveillance testing that was performed on May 8, 2017. As a result, the right channel of the safety injection system actuation logic was inoperable until the problem was discovered during troubleshooting and the wire was subsequently re-landed onMay 3, 2018 |
| 05000255/FIN-2018415-01 | 2018Q2 | Green | NRC identified | Security | ||
| 05000255/FIN-2018011-03 | 2018Q2 | Green | Licensee-identified | Licensee-Identified Violation | License condition 2.C(3)requires the licensee to implement and maintain in effect all provisions of the approved Fire Protection Program that complies with Title 10of the Code of Federal Regulations(CFR), Part50.48(a) and 10 CFR 50.48(c), NFPA Standard NFPA 805, as approved in the Safety Evaluation Report (SER)dated February 27, 2015. Section 2.4.3.3 of NFPA 805 states, in part, that the Probabilistic Safety Assessment (PSA)(Probabilistic RiskAssessment (PRA))approach, methods, and data shall be based on the as-built and as-operated and maintained plant, and reflect the operating experience at the plant.Contrary to the above, from February 27, 2015, until May 14, 2018, the licensee failed to base the PSA (PRA) approach, methods, and data on the as-built and as-operated and maintained plant.Specifically, the licensees PSA (PRA) model/analysis credited the suppression system located in the cable spreading room to suppress a type 2 fire scenarios, whereas the actual room contained numerous obstructions by the stacked cable trays located near the ceiling that interfered with the water spray pattern discharged from the sprinklers from providing adequate water density pattern to suppress a fire in areas below the cable trays which contained electrical panels.Significance/Severity Level: The performance deficiency was determined to be more-than-minor, and therefore, a finding because the performance deficiency, if left uncorrected, would have the potential to lead to a more significant safety concern. Specifically, the licensees failure to correctly model/analyze the as-built condition of the suppression system located in the cable spreading room in the PRA could potentially affect the risk associated with a fire in the room and could result in inappropriately screening out the effects of otherchanges associated with the fire area.The finding was of very-low safety significance (Green). While there may be a change to the plants baseline risk as a result of this issue, this is a fire modeling issue only; no physical plant fire protection feature was altered by the fire PRA model. Therefore, there was no increase in actual core damage risk to the physical plant. | |
| 05000255/FIN-2018011-02 | 2018Q2 | NRC identified | Failure to Set Action Levels to Ensure that the Assumptions in the Engineering Analysis Remain Valid | The inspectors reviewed a sample of equipment located in the fire areas selectedfor inspection to determine if the licensee had established a proper method of monitoring that equipment as required by NFPA 805, Section 2.6. Section 2.6 of NFPA 805 required that, A monitoring program shall be established to ensure that the availability and reliability of the fire protection systems and features are maintained and to assess the performance of the fire protection program in meeting the performance criteria. Monitoring shall ensure that the assumptions in the engineering analysis remain valid. The licensee utilized Procedure EN-DC-357, NFPA 805 Monitoring Program, Revision 2,to ensure that, the assumptions in the NFPA 805 engineering analyses remain valid by executing an effective and ongoing monitoring program.The inspectors selected the high pressure air compressor (C-6B) and high pressure safety injection pump (P-66B), both of which were located in the West Safeguards Room. The licensee considered these components to be high-safety significant (HSS) structures, systems, or components (SSCs). The licensee chose to monitor the unavailability of these components utilizing the Maintenance Rule (10 CFR 50.65).The licensee set the Maintenance Rule allowable unavailability action level threshold for the high pressure air compressorat 5E-2 (5percent)whereas they assumed in their fire PRA an unavailability of 9.86E-3 (approximately 1percent). For the high pressure safety injection pump the licensee set the Maintenance Rule allowable unavailability at 1.5E-2 (1.5percent) whereas they assumed in their fire PRA an unavailability of 6.32E-3 (approximately 0.6percent). The inspectors believed that by relying on the less conservative action level thresholds in the Maintenance Rule the licensee failed to ensure that the assumptions in the engineering analysis (fire PRA) remained valid.The licensee stated in Procedure EN-DC-357, Section 1.0, Purpose, that, The NFPA 805 Monitoring Program ensures that the assumptions in the NFPA 805 engineering analyses remain valid by executing aneffective and ongoing monitoring program. Under Section 3.0, Definitions, the licensee defined, Action Level Threshold, as, When establishing the action level threshold for reliability and availability, the action level should be no lower than the Fire Probabilistic Safety Analysis (also called fire PRA) assumptions. The licensee stated in Section 5.3.3(c) that, If HSS SSCs have been identified in using the Maintenance Rule guidelines, the associated SSC specific performance criteria may be established as in the Maintenance Rule, provided the criteria are consistent with the Fire Probabilistic safety Analysisassumptions... The inspectors believed that Procedure EN-DC-357 required the licensee set the action level thresholds no lower than the fire PRA assumptions. Procedure section 5.3.4(b)(1) required that HSS equipment that is not sufficiently tracked in the Maintenance Rule be added to the NFPA 805 Monitoring Database. The licensee did not add the high pressure air compressor and the high pressure safety injection pump into the NFPA 805 Monitoring Database. In the SER 2015-2-27 dated February 27, 2015, in which the staff approved the licensee NFPA805 License Amendment Request, the staff noted that the licensee will develop an NFPA 805 Monitoring Program consistent with Frequently Asked Question (FAQ)10-0059. The staff also noted that the stated development of the Monitoring Program would include a review of existing surveillance, inspection, testing, compensatory measures, and oversight 8processes for adequacy. The staff concluded in SER 2015-2-27 that since the final values for availability and reliability, as well as the performance criteria for the SSCs being monitored, have not been established for the Monitoring Program as of the date of this SER, completion of the licensee's NFPA 805 Monitoring Program is an implementation item. Furthermore, the staff concluded that there is reasonable assurance that the licensee will develop a Monitoring Program that meets the requirements specified in Sections 2.6.1, 2.6.2, and 2.6.3 of NFPA 805Section 2.6 of NFPA 805 stated in part that, Monitoring shall ensure that the assumptions in the engineering analysis remain valid. The licensee interpreted this statement to mean that utilizing the existing Maintenance Rule unavailability values is consistent with its commitment in SER 2015-2-27 and would allow the site to appropriately monitor the availability and reliability of fire protection systems and features. The licensee also performed sensitivity studies on the differences in the unavailability values of fire protection systems and features between the Maintenance Rule criteria and the fire PRA values and determined that they were not risk-significant. The inspectors questioned the appropriateness of the licensees interpretation of assumptions as described in Section 2.6 of NFPA 805 above. The inspectors believed that the licensee should monitor the unavailability of fire protection systems and features utilizing the same values as thosedocumented in the fire PRA associated with the NFPA 805 License Amendment Request. The licensee further stated that they were waiting for guidance from the NRCs Office of Nuclear Reactor Regulation and the industry who were working on revising guidance in FAQ10-0059, NFPA 805 Monitoring, to determine if they needed to change their approach. That guidance document was in the process of being revised during the inspection. The inspectors needed to determine if the licensees approach to monitoring the availability and reliability of the fire protection systems and features using the Maintenance Rule monitoring values in order to ensure that the assumptions in the engineering analysis remained valid was an acceptable approach.Planned Closure Action(s): The inspectors will await clarification from the Office of Nuclear Reactor Regulation in order to determine if a performance deficiency exists.Licensee Action(s): The licensee plans to follow the resolution of FAQ 10-0059, Revision 6, and take the appropriate corrective actions based on the guidance provided in that FAQ. | ||
| 05000255/FIN-2018011-01 | 2018Q2 | Green | NRC identified | Failure to Maintain Adequate Fire Protection System Functional Test Procedure | The inspectors identified a finding of very-low safety significance and associated violation of Technical Specification 5.4.1, Procedures,for the licensees failure to maintain fire protection system functional test procedure. Specifically, the licensee failed to maintain Procedure RO-52, Fire Suppression Water System Functional Test and Fire Pump Capacity Test, by failing to include appropriate acceptance criteria in the procedure to demonstrate fire protection system functionality. | |
| 05000255/FIN-2018010-01 | 2018Q1 | Green | Licensee-identified | 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 | Green | Licensee-identified | 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 | Severity level Enforcement Discretion | NRC identified | 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 | Green | Self-revealing | 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 | Green | H.2 | Self-revealing | 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-2017007-03 | 2017Q4 | NRC identified | Containment Dome Truss Analysis | Thedome truss system was originally designed to support the containment liner plate and wet concrete during the construction of the containment dome (i.e., the liner plate initially acted as a form and the truss supported the form). After the concrete cured, the dome truss system was lowered away from the liner and was used to support the safety injection tanks(SITs) and CS system piping and their associated supports. The CS and SIT systemsare both safety-related which were required to be evaluated for seismic loads (self-weight and externally applied loads). The dome truss system would have alsobeen required to be evaluated for seismic loads. The UFSAR,Section 6.1,described the safety-related design function of the SITsystemwas to prevent fuel and cladding damage that could interfere with adequate emergency core cooling, and to limit the cladding-water reaction to less than approximately 1percentfor all break sizes in the primary system piping up to and including the double-ended rupture of the largest primary coolant pipe, for any break location, and for the applicable break time. Also,the SITsystem also functions to provide rapid injection of large quantities of borated water for added shutdown capability during rapid cooldown of the primary system caused by a rupture of a main steam line. UFSAR Section 6.2.1 described the safety-related design function of the CSsystem was to limit the containment building pressure rise and reduce the airborne radioactivity in containment by providing a means for spraying the containment atmosphere after occurrence of a LOCAor a main steam line break.The inspectors requested the design basis analysis of the dome truss system that considers the LOCA loading on the dome truss system as well as the seismic loading due to the applied design loads from the CS and SITsystem. During the time of the inspection, the licensee was unable to locate the dome truss analysis. In response to the inspectors concern, the licensee entered the issue into their CAP asCR 2017-05016, Dome Trusses, dated November 1, 2017. The licensee is investigating the containment dome truss analysis further with the vendor of the dome truss system.This issue is a URI pending additional inspector review of the design basis analysis for the containment dome truss system. (URI 05000255/2017007-03; Containment Dome Truss Analysis) | ||
| 05000255/FIN-2017007-02 | 2017Q4 | Green | NRC identified | Containment Spray Pipe Support Strap Deficiencies | The inspectors identified a finding of low safety significance (Green) and an associated potential NCV of Title 10of theCode of Federal Regulations,Part 50, Appendix B, Criterion III, Design Control, for failure to meet Updated Final Safety Analysis Reportrequirements for containment spraypiping supports, specifically straps. Specifically, the inspectors identified that Calculation No. EA-SP-03369-02, Revision 0, used inelastic acceptance limits for the pipe straps which connect the pipe to the pipe support, in order to demonstrate Class I compliance which was not in accordance with the design and licensing basis specification. The license entered the issue into their Corrective Action Programas CR-PLP-2017-05246, Spray Pipe Support,dated November 14, 2017. The licensee performed an analysis to establish reasonable assurance of operability and the inspectors with support from the Office from the Nuclear Reactor Regulation reviewed this operability and no performance deficiencies were identified.The performance deficiency was determined to be more-than-minor because it was associated with the Barrier Integrity Cornerstone attribute of design control and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public fromradionuclide releases caused by accidents or events. This finding is of very-low safety significance (Green) because there was no actual reactor containment barrier degradation. The inspectors did not identify a cross-cutting aspect associated with thisfinding because this was a legacy design issue; and therefore, was not reflective of current performance. | |
| 05000255/FIN-2017007-01 | 2017Q4 | Green | NRC identified | Failure to Periodically Test the Emergency Diesel Generators Capacity to Start and Accelerate Design Basis Sequenced Loads | The team identified a finding of very-low safety significance (Green) and an associated NCV of Title 10 of the Code of Federal Regulations, Part50, Appendix B, Criterion XI, Test Control, for the failureto periodically test the emergency diesel generators(EDGs) capability to start and accelerate all of the sequenced loads within the applicable design voltage and frequency transient and recovery limits.Specifically, EDG testingactivities did not demonstrate that all of the EDG auto-sequenced loads started and accelerated within the applicable voltage and frequency limits during start-up and recovery. In addition, the licensee did not perform adequate post-modification testing after replacing the EDG governor controller system or voltage regulators. Thelicensee captured theseissuesin their Corrective Action Programas Condition Report (CR)2017-05265 and CR 2017-05283, and performed an operability evaluation which reasonably determined the affected structures, systems, and componentswere operable.The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems thatrespond to initiating events to prevent undesirable consequences. The finding screened as of very-low safety significance (Green) becauseit did not result in the loss ofoperability or functionality of mitigating systems. Specifically, the licensee evaluated the most recent voltage and frequency data from the last EDG output breaker testsin which the data recorder was left running after the output breaker shut and reasonably determined that the EDGs and the affected loads were operable. The team did not identify a cross-cutting aspect associated with this finding because it was not confirmed to reflect current performance due to the age of the performance deficiency. Specifically, the associated testingprocedures were established more than 3years ago. | |
| 05000255/FIN-2017201-02 | 2017Q3 | H.14 | NRC identified | Security | ||
| 05000255/FIN-2017201-01 | 2017Q3 | H.4 | NRC identified | Security | ||
| 05000255/FIN-2017003-01 | 2017Q3 | NRC identified | 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 | NRC identified | 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 | Green | Licensee-identified | 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-2017003-03 | 2017Q3 | Green | H.5 | NRC identified | 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-2017002-01 | 2017Q2 | Green | NRC identified | 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. | |
| 05000255/FIN-2017001-01 | 2017Q1 | Green | Licensee-identified | 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-2016004-03 | 2016Q4 | Green | H.2 | NRC identified | 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-2016404-01 | 2016Q4 | Green | P.3 | NRC identified | Security | |
| 05000255/FIN-2016004-01 | 2016Q4 | Green | H.9 | NRC identified | 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 | Green | Self-revealing | 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-2016404-02 | 2016Q4 | Green | Licensee-identified | Licensee-Identified Violation | ||
| 05000255/FIN-2016009-01 | 2016Q3 | Green | H.9 | NRC identified | Failure to Document 50.59 Evaluation for Removal of Eight Hour Operator Rounds from the FSAR | The inspectors identified a Severity Level IV, Non-Cited Violation of Title 10 of the Code of Federal Regulations (CFR), Part 50.59, Changes, Tests, and Experiments, and an associated finding of very low safety significance (Green) for the licensees failure to maintain records of a change in the facility which included a written evaluation that provided the bases for the determination that the change did not require a license amendment. Specifically, the licensee failed to have a written evaluation that provided the bases for why removal of the 8-hour operator rounds credited to detect a Spent Fuel Pool (SFP) dilution event from the Final Safety Analysis Report did not require a license amendment. The licensee entered this issue into their Corrective Action Program (CAP) as CR-PLP-2016-03055 and issued a standing order to log SFP level every eight hours as an immediate corrective action. The licensees planned corrective actions include preparation of a 10 CFR 50.59 evaluation for the change. The inspectors determined that the failure to perform a 10 CFR 50.59 evaluation for the change to the Final Safety Analysis Report which removed the eight hour operator rounds credited to detect a SFP dilution event was contrary to 10 CFR 50.59(d)(1), and was a performance deficiency. The inspectors determined the performance deficiency was more than minor, and a finding, because it was associated with the barrier integrity cornerstone attribute of Configuration Control and adversely affected the associated Cornerstone Objective of ensuring that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the removal of the 8-hour operator rounds is associated with the boron concentration reactivity control in the SFP and could adversely affect the fuel claddings function to protect the public from radionuclide releases. In addition, the associated 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 inspectors evaluated the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings at Power, dated June 19, 2012, Exhibit 3, for the Barrier Integrity cornerstone and were directed to further evaluate the significance of the finding using IMC 0609 Appendix M, Significance Determination Process Using Qualitative Criteria, dated April 12, 2012. The inspectors performed the qualitative evaluation described in IMC 0609, Appendix M, and determined the significance of the finding to be of very low safety significance (Green) by considering the availability of other measures the licensee had in place to detect a SFP dilution event. In accordance with Section 6.1.d of the NRC Enforcement Policy this violation is categorized as Severity Level IV because the resulting changes were evaluated by the SDP as having very-low safety significance (i.e., Green finding). The inspectors determined the associated finding had a cross-cutting aspect in the area of Human Performance because the licensee did not ensure their staff were adequately trained in the implementation of the 10 CFR 50.59 rule. Specifically, the licensee staff did not realize that a change which fundamentally alters the existing means of performing or controlling design functions (removal of the 8-hour operator rounds for detecting a SFP dilution event in lieu of an automatic alarm) is adverse and requires an evaluation. |
| 05000255/FIN-2016407-01 | 2016Q3 | Severity level Enforcement Discretion | Licensee-identified | Licensee-Identified Violation | ||
| 05000255/FIN-2016003-02 | 2016Q3 | NRC identified | 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-2016003-01 | 2016Q3 | Green | P.2 | Self-revealing | 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-2016007-01 | 2016Q2 | Green | P.2 | NRC identified | Failure to Correct Containment Spray Pump Non-conformance | The team identified a finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to promptly correct a condition adverse to quality. Specifically, the licensee failed to correct a non-conforming condition for containment spray pump P54A, which was discovered in October 2014, during an NRC component design bases inspection (CDBI). The licensee entered this issue into their CAP as CRPLP201601646 with an assigned action to resolve the non-conforming condition of the containment spray pump The team determined that the performance deficiency was more than minor because it was associated with the Mitigating Systems cornerstone attribute of Design Control and adversely affected the cornerstones 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 performance deficiency identified that the licensee failed to correct a non-conformance between their current as-built configuration, the current licensing bases (i.e., Final Safety Analysis Report (FSAR) Section 6.2.3.1), and the design basis (i.e., Design Basis Calculation EAELECLDTAB005) which was identified by the NRC in the 2014 CDBI. In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, issued June 19, 2012, and Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, issued June 19, 2012, the team answered No to all of the questions. Therefore, this finding was of very low safety significance (Green). The team identified a cross-cutting aspect in the Evaluation component of the Problem Identification and Resolution cross-cutting area because the licensee failed to fully evaluate the original issue identified in the 2014 CDBI to ensure that the corrective actions performed adequately addressed the non-conformance. Specifically, the licensee evaluated the effect of the non-conformance, but failed to correct the underlying non-conformance between the licensing basis, the as-built configuration, and the design basis. |
| 05000255/FIN-2016001-01 | 2016Q1 | NRC identified | 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 | Green | H.4 | Self-revealing | 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-2016001-04 | 2016Q1 | Green | Licensee-identified | 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-2016001-03 | 2016Q1 | Green | H.7 | NRC identified | 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-2016001-05 | 2016Q1 | Green | Licensee-identified | 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). | |
| 05000255/FIN-2016403-01 | 2016Q1 | Green | H.5 | NRC identified | Security | |
| 05000255/FIN-2015004-05 | 2015Q4 | Green | Licensee-identified | 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-2015004-04 | 2015Q4 | Green | NRC identified | 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-2015004-02 | 2015Q4 | Green | NRC identified | 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 | Green | H.2 | NRC identified | 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 | Green | NRC identified | 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-2015012-01 | 2015Q3 | Severity level III | Licensee-identified | Inaccurate/Incomplete Information Submitted For Relief Request 4-18 | An apparent violation (AV) of Title 10 of the Code of Federal Regulations (CFR) 50.9 was identified by the licensee, related to a failure to provide information that was complete and accurate in all material respects to the NRC in letter PNP 2014-015, Relief Request (RR) Number 4-18 - Proposed Alternative Use of Alternate ASME (American Society of Mechanical Engineers) Code Case N-770-1 Baseline Examination. Specifically, in this document the licensee stated, In the unlikely case that crack initiation were to occur, crack growth calculations considering primary water stress corrosion cracking (PWSCC) as the failure mechanism demonstrate that the hot leg drain nozzle weldment satisfies ASME Code acceptance criteria for 60 effective full power years (EFPY) for a circumferential flaw, and more than 34 years for an axial flaw. However, this statement was not correct or accurate in that, the ASME Code acceptance criteria were not satisfied for 60 EFPY for a circumferential flaw and 34 years for an axial flaw, where correct information was 20 EFPY for a circumferential flaw, and 11.3 years for an axial flaw. This AV was not an immediate safety concern because the licensee demonstrated an adequate basis for continued operability of the nine affected primary coolant system (PCS) welds. The licensee corrective actions for this AV included completion of an operability evaluation, submittal of a corrected analysis to the NRC, and entering this issue into the Corrective Action Program (CAP) (CR-PLP-2015-03441). If the NRC was provided with the correct information in letter PNP 2014-015, where the affected welds satisfied ASME Code acceptance criteria (i.e., 75 percent through-wall) for only 20 effective full power years for a circumferential flaw, and 11.3 years for an axial flaw, the NRC would not likely have approved RR 4-18 and, as a minimum, would have requested additional supporting analysis (e.g., required substantial further inquiry). Further, the need for substantial further inquiry was illustrated by the licensees subsequent decision in RR 4-21 to abandon the prior analytical approach used in RR 4-18. The inspectors evaluated the underlying technical issue in accordance with the SDP to determine the risk significance of this AV. The issue of concern was of more than minor significance because it was similar to the not minor if aspect of Example 3j in IMC 0612, Appendix E, Example of Minor Issues. Specifically, the erroneous information provided in letter PNP 2014-015 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 error for the operability of nine PCS welds 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). No cross-cutting aspect was assigned because this Green finding was identified by the licensee. | |
| 05000255/FIN-2015003-01 | 2015Q3 | Green | P.5 | NRC identified | 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-2015012-02 | 2015Q3 | Green | P.2 | NRC identified | 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. |
| 05000255/FIN-2015407-02 | 2015Q3 | Green | Licensee-identified | Licensee-Identified Violation | ||
| 05000255/FIN-2015003-02 | 2015Q3 | Green | H.9 | NRC identified | 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-2015407-03 | 2015Q3 | Green | Licensee-identified | Licensee-Identified Violation |