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| Q | Significance | CCA | Identified by | Title | Description | |
|---|---|---|---|---|---|---|
| 05000333/FIN-2018412-01 | 2018Q3 | Green | NRC identified | Security | ||
| 05000333/FIN-2018411-02 | 2018Q2 | Severity level IV | NRC identified | Security | ||
| 05000333/FIN-2018411-01 | 2018Q2 | Green | H.3 | NRC identified | Security | |
| 05000333/FIN-2018002-01 | 2018Q2 | Green | Licensee-identified | Licensee-Identified Violation | This violation of very low safety significance was identified by Exelon and has been entered into Exelons CAP and is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy Violation: 10 CFR 71.5 requires that licensees who transport licensed material comply with the applicable requirements of the Department of Transportation (49 CFR). 49 CFR 172.202(a)(1) and (a)(2) require that the shipping description on the shipping paper include the proper shipping name and identification number for the material. 49 CFR 172.302(a) requires that shipments in bulk packages be marked with the identification number. Contrary to the above, on July 12, 2016, the shipping description on the shipping paper for shipment JAF-2016-1613 from FitzPatrick to Tennessee did not include the proper shipping name and identification number for the material. Exelon identified the error during a subsequent review of the shipping paperwork. Significance/Severity Level: No examples of transportation issues are presented in IMC 0612, Appendix E (Examples of Minor Issues). IMC 0609, Appendix D, Section VII.C.e.1 lists examples of Green findings that include documentation deficiencies including failure to properly document compliance with 49 CFR requirements such as shipping papers. Corrective Action Reference: Exelon placed this issue into its CAP as CR-JAF-2016-02857. Corrective actions included providing a corrected shipping paper to the facility in Tennessee that had received the package. | |
| 05000333/FIN-2017004-03 | 2017Q4 | Green | Licensee-identified | Licensee-Identified Violation | The following violation of very low safety significance (Green) was identified by Exelon and is a violation of NRC requirements, which meets the criteria of the NRC Enforcement Policy for being dispositioned as a NCV. 10 CFR 50.65(a)(4) states, in part, that before performing maintenance activities (including but not limited to surveillance, post-maintenance testing, and corrective and preventive maintenance), the licensee shall assess and manage the increase in risk that may result from the proposed maintenance activities. The scope of the assessment may be limited to structures, systems, and components that a risk-informed evaluation process has shown to be significant to public health and safety. Contrary to the above, on March 28, 2011, and April 16, 2015, before performing maintenance activities on the electric bay unit coolers, as discussed in Section 4OA2.4, Exelon did not assess and manage the increase in risk that resulted from the maintenance activities. This issue was documented in Condition Report JAF-2016-0838. In accordance with IMC 0609.04, Initial Characterization of Findings, and IMC 0609 Appendix K, Flowchart 2, Assessment of RMAs. The inspectors determined that the violation was of very low safety significance (Green) because the incremental core damage probability was less than 1E-5, with three risk management actions taken during the maintenance activities. | |
| 05000333/FIN-2017004-02 | 2017Q4 | Green | H.12 | Self-revealing | Human Error Resulting in Unplanned HPCI Isolation | A self-revealing NCV of very low safety significance (Green) of Technical Specification (TS) 5.4, Procedures, was identified for a procedural error which resulted in the inadvertent isolation of the high pressure coolant injection (HPCI) system. Specifically, on April 4, 2017, an instrumentation and controls (I&C) technician did not correctly perform procedure ISP-175B1, Reactor and Containment Cooling Instrument Functional Test/Calibration, which caused the HPCI system to isolate. Exelons immediate response to the event included stopping the surveillance test, and developing and implementing a plan to restore the HPCI system to an operable status. The HPCI system was subsequently restored to service approximately five hours after the inadvertent isolation. Additional corrective actions included increased observations of peer checks and validation of I&C activities. This issue was entered into the CAP as IR 03993791. This performance defficiency is more than minor because it is associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to correctly implement procedure ISP-175B1 caused an isolation of the HPCI system and rendered it unavailable to respond to an initiating event. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2012, the inspectors determined that this finding required a detailed risk evaluation since the HPCI isolation resulted in a loss of safety function. Using the Standardized Plant Assessment Risk Model (SPAR), the Region I senior reactor analyst (SRA) determined this finding was of low safety significance (Green). The inspectors determined that this finding had a cross-cutting aspect in the area of Human Performance, Avoid Complacency, because the I&C technician did not correctly implement error reduction tools and verify that the direct current voltage source was installed on the correct trip unit prior to performing the surveillance procedure. (H.12) |
| 05000333/FIN-2017004-01 | 2017Q4 | Green | NRC identified | Inadequate Design Control for Battery Sizing Calculation | The inspectors identified a Green NCV of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion III, Design Control, because Exelon did not verify the adequacy of the low pressure coolant injection (LPCI) motor operated valve (MOV) independent power supply (IPS) with respect to the 419 volt direct current (VDC) battery sizing calculation. Specifically, non-conservative design inputs were used for the safety-related battery sizing calculation which reduced the battery capacity margin. On November 22, 2017, Exelon performed an operability determination for the identified issue and determined that the batteries had sufficient capacity. This issue was entered into the corrective action program (CAP) as issue report (IR) 4079452. The performance deficiency was determined to be more than minor because it was associated with the Design Control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, based on the quantity and magnitude of the errors, there was reasonable doubt that the LPCI MOV batteries would have adequate capacity under all design conditions. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2012, the finding was of very low safety significance (Green) because it was a design deficiency confirmed not to result in a loss of operability. This finding does not have a cross-cutting aspect because the calculation was last revised in 2003 so the finding is not indicative of current performance. | |
| 05000333/FIN-2017003-01 | 2017Q3 | Green | NRC identified | Vent Line Socket Weld Failure | On January 14, 2017, during the initial drywell walkdown following shutdown for a refueling outage, Entergy personnel identified a through-wall leak on the vent line off of the bonnet of the motor operated gate valve on the suction side of the A reactor water recirculation pump. A three- to four-foot steam plume was observed. Entergy determined this constituted a violation of TS 3.4.4, RCS Operational Leakage, that requires RCS leakage to be limited to no pressure boundary leakage. Based on the unidentified leakage rate of 0.06 gallons per minute measured during plant operation and visual inspection of the leak area, the leak likely existed while the plant was online. The condition was reported in Event Notification 52490 as required by 10 CFR 50.72(b)(3)(ii)(A) because it represented a degradation of a principal safety barrier. The inspectors reviewed LER 05000333/2017-001, CR-JAF-2017-00245, and the associated apparent cause evaluation. Entergy determined that this leak was caused by the existing pipe support allowing for excessive lateral movement which led to higher stresses in the socket weld connection. Additionally, the recirculation pumps were operated at a reduced flow condition for an extended period during the previous cycle, which likely resulted in an increased number of vibration cycles. The inspectors also reviewed the leakage data over the previ ous cycle and Entergys operational decision making IR and determined that the existence of RCS pressure boundary leakage was not within Entergys ability to foresee and correct and therefore was not a performance deficiency. The inspectors screened the significance of the condition using IMC 0609, Appendix A, The Significance Determination Process (SDP) For Findings At-Power, and determined that the condition represented very low safety significance (Green) because it would not have resulted in exceeding the RCS leak rate for a small loss of coolant accident and would not have likely affected other systems used to mitigate a loss of cooling accident. Enforcement. TS 3.4.4 requires, in part, that RCS operational leakage shall be limited to no pressure boundary leakage. If pressure boundary leakage exists, the TS 3.4.4 limiting condition for operation action statement requires the unit be in at least hot shutdown within 12 hours and in cold shutdown within 36 hours. Contrary to the above, for a period that began on an unknown date that was likely more than 36 hours before January 14, 2017, and ending on January 14, 2017, RCS pressure boundary leakage existed, and the licensee did not place FitzPatrick in at least hot shutdown within 12 hours and in cold shutdown within 36 hours. This issue is considered within the traditional enforcement process because there was no performance deficiency associated with the violation of NRC requirements. IMC 0612, Power Reactor Inspection Reports, Section 03.22 states, in part, that traditional enforcement is used to disposition violations receiving enforcement discretion or violations without a performance deficiency. The NRC Enforcement Policy, Section 2.2.1 states, in part, that, whenever possible, the NRC uses risk information in assessing the safety significance of violations. Accordingly, after considering that the condition represented very low safety significance, the inspectors concluded that the violation would be best characterized as Severity Level IV under the traditional enforcement process. However, the NRC is exercising enforcement discretion (EA-17-121) in accordance with Section 3.10 of the NRC Enforcement Policy, which states that the NRC may exercise discretion for violations of NRC requirements by reactor licensees for which there are no associated performance deficiencies. In reaching this decision, the 19 NRC determined that the issue was not with in the licensees ability to foresee and correct, the licensees actions did not contribute to the degraded condition, and the actions taken were reasonable to identify and address the condition. Furthermore, because the licensees actions did not contribute to this violation, it will not be considered in the assessment process or the NRCs Action Matrix. This LER is closed. | |
| 05000333/FIN-2017404-01 | 2017Q3 | Green | P.3 | NRC identified | Security | |
| 05000333/FIN-2017002-01 | 2017Q2 | Green | P.2 | Self-revealing | A Control Room Ventilation Subsystems Inoperable Longer than Allowed by Technical Specifications | Green. A self-revealing Green NCV of Technical Specification (TS) 3.7.3, Control Room Emergency Ventilation Air Supply (CREVAS) System, and TS 3.7.4, Control Room Air Conditioning (AC) System, was identified for the failure to declare one subsystem of the control room AC and CREVAS systems inoperable. Specifically, on August 16, 2016, control room operators failed to declare the A CREVAS and A control room AC subsystems inoperable due to a degraded damper actuator. As a result, the A CREVAS and A control room AC subsystems were inoperable from August 16, 2016, until a compensatory measure to assist the dam per linkage by hand as needed was implemented on September 19, 2016, which exceeded the TS allowed outage time. On October 4, 2016, FitzPatrick personnel replaced the actuator. This issue was entered into the corrective action program (CAP) as JAF-CR-2016-3593. The performance deficiency is more than minor because it is associated with the structure, system, and component (SSC) and barrier performance 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, this resulted in the A control room AC and A CREVAS subsystems being inoperable from August 16, 2016, to September 19, 2016, and the exceedance of the allowable TS out-of-service times. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 3 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2012, the inspectors determined that this finding is of very low safety significance (Green) because the performance deficiency did not represent a degradation of the radiological barrier function provided for the control room, and the finding did not represent a degradation of the barrier function of the control room against smoke or a toxic atmosphere (i.e. the B train of both subsystems remained operable). This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, because FitzPatrick personnel failed to thoroughly evaluate the problem such that resolution addressed the cause. Specifically, FitzPatrick failed to fully evaluate the degraded condition during troubleshooting following the failed post-maintenance test (PMT) on August 16, 2016. Thorough testing and evaluation of the degraded actuator would have led to the identification of the need for replacement to restore the damper and its actuator to fully operable status. (P.2) |
| 05000333/FIN-2016004-01 | 2016Q4 | Green | H.11 | NRC identified | Failure to Ensure Proper Configuration Control of a PCIV During Planned Maintenance | Green. The inspectors identified a Green NCV of Technical Specification (TS) 5.4, Procedures, because Entergy staff did not implement procedure AP-12.06, Equipment Status Control, as required. Specifically, Entergy personnel did not recognize the impact of a change associated with the tagout of a C residual heat removal (RHR) system primary containment isolation valve (PCIV). This resulted in motor operated valve 10MOV-13C being electrically isolated in the open position without being recognized as a PCIV and without proper entry into TS 3.6.1.3. Entergy restored the valve to operable status, entered this issue into their corrective action program (CAP) as condition report (CR)-JAF-2016-4419, and conducted meetings with each operating crew to discuss the event and reinforce standards for equipment status control and maintaining a questioning attitude. Training was also provided to operators to review the scenario and discuss requirements associated with PCIVs. This finding is more than minor because it was associated with the configuration control attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers (containment) protect the public from radionuclide releases caused by accidents or events. Specifically, Entergy staff did not recognize the impact of a change associated with the tagout of a containment isolation valve. The change in the tagout resulted in a failure to isolate the containment isolation valve and enter TS 3.6.3.1 prior to maintenance. The finding was similar to Example 3.j in Appendix E of IMC 0612, Examples of Minor Issues, issued August 11, 2009. Since the PCIV was in an open position with power removed, a reasonable doubt of operability existed due to the valves inability to close to perform its containment isolation function. The inspectors evaluated this finding using IMC 0609.04, Initial Characterization of Findings, issued October 7, 2016; Exhibit 3 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2012; and Appendix H of IMC 0609, Containment Integrity Significance Determination Process, issued May 6, 2004. Using Exhibit 3 of IMC 0609, Appendix A, Section B, Reactor Containment, the finding directed the use of IMC 0609, Appendix H because it represented an actual open pathway in the physical integrity of reactor containment (i.e. valve). Using IMC 0609, Appendix H, the finding was classified as a Type B finding because it was related to a degraded condition that had potentially important implications for the integrity of containment, without affecting the likelihood of core damage (i.e. containment isolation was precluded by the isolation valve being failed in the open position, however the low pressure coolant injection function remained 4 available). Using Table 6.1, Phase 1 Screening-Type B Findings at Full Power, for a boiling water reactor, Mark 1 Containment, the inspectors were directed to perform a Phase 2 Assessment because the structure, system, and component (SSC) affected by the finding was a containment isolation valve. Using Table 6.2, Phase 2 Risk Significance-Type B Findings at Full Power, the inspectors determined that the failure of the containment isolation valve critical to suppression pool integrity/scrubbing was less than 3 days, and therefore was of very low safety significance (Green). This finding has a cross-cutting aspect in the area of Human Performance, Challenge the Unknown, because Entergy failed to maintain a questioning attitude to identify an improper configuration associated with a PCIV tagout during maintenance planning and execution. Specifically, a tagout writer modified the configuration for a containment isolation valve, which was not challenged or questioned during subsequent reviews. This resulted in the PCIV being tagged out in the open position, a condition that rendered the valve inoperable. (H.11) |
| 05000333/FIN-2016403-01 | 2016Q3 | Green | H.8 | NRC identified | Security | |
| 05000333/FIN-2016003-01 | 2016Q3 | Green | Self-revealing | Inadequate Preventive Maintenance Results In Transformer Failure and Reactor Scram | A self-revealing Green finding (FIN) was identified for Entergy staffs failure to properly implement the requirements of EN-DC-324, Preventive Maintenance Program, Revision 16, to ensure proper preventive maintenance (PM) was implemented for non-safetyrelated 4KV transformer 71T-5. Specifically, Action Request (AR) 127566, PM change request to perform inspection, cleaning, and electrical testing of 4KV transformer 71T-5 was retired without a review by engineering as required by the PM program. As a result, transformer 71T-5 remained in service beyond its effective life without proper condition monitoring and maintenance, leading to its failure and a reactor scram on June 24, 2016. Entergy staff developed corrective actions to address the failure which included replacement of the transformer and re-establishing the condition monitoring and PM task. Entergy also performed an extent of condition review that confirmed the PM to clean, inspect, and test similar non-safety-related dry-type transformers was still active and performed within its required frequency. This finding is more than minor because it is associated with the Equipment Performance attribute of the Initiating Events cornerstone and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. Specifically, Entergy staff failed to ensure an adequate PM was in place for transformer 71T-5. The PM to ensure adequate cleaning and testing was cancelled in 2011, and transformer 71T-5 ultimately failed on June 24, 2016, resulting in a manual reactor scram. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 1 of IMC 0609, Appendix A, The Significance Determination Process for Findings AtPower, issued June 19, 2012, the inspectors determined that this finding is of very low safety significance (Green) because although the performance deficiency caused a reactor scram, it did not result in the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The inspectors did not assign a cross-cutting aspect to this finding because it is not indicative of current licensee performance. Specifically, the performance deficiency was determined to have occurred in 2011, the guidance in ENDC-324 is clear regarding the PM change process, and no additional failures to follow the process have resulted in significant reactor transients. | |
| 05000333/FIN-2016007-01 | 2016Q2 | Green | NRC identified | Failure to ensure design basis of EDG LO storage facility | The team identified a Green NCV of Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Appendix B, Criterion III, Design Control, because Entergy did not ensure that FitzPatricks emergency diesel generator (EDG) lubrication oil (LO) supply storage facility was designed to withstand the effects of natural phenomena. Specifically, additional LO, maintained and inventoried monthly to ensure an adequate LO supply to meet the EDGs seven day mission time, was stored in a non-Class I structure that was not designed to withstand the effects of natural phenomena induced plant events that the EDGs were designed to mitigate. Entergy entered the issue into the corrective action program (CAP) as condition report (CR) 2016-1471 and promptly relocated the LO reserve inventory from warehouse No. 2 to the EDG building, which is constructed to Class I seismic and tornado protection design criteria. The finding was more than minor because it was associated with the protection against external factors attribute of the Mitigating Systems cornerstone and adversely affected the cornerstones objective of ensuring reliability and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). The team evaluated the significance of this finding using IMC 0609, Appendix A, The Significance Determination Process for Findings at Power, Exhibit 2 Mitigating Systems Screening Questions. The team determined the finding screened as very low safety significance (Green) because the finding was a design deficiency which did not result in an actual loss of functionality of the EDGs. This finding did not have a cross-cutting aspect because the underlying cause occurred in 1988 during warehouse No. 2 construction and was not indicative of current Entergy performance. | |
| 05000333/FIN-2016007-02 | 2016Q2 | Green | P.2 | NRC identified | Failure to adequately evaluate a procedure change impacting a PRA-credited time critical operator action | The team identified a Green finding involving Entergys inability to complete a time critical operator action within the assumed probabilistic risk assessment (PRA) credited accident mitigation time limit to prevent undesirable consequences (i.e., core damage) under a postulated scenario (i.e., using the residual heat removal service water (RHRSW) system as an alternate injection source into the reactor pressure vessel (RPV) via the residual heat removal (RHR) system during a loss of coolant accident (LOCA)). Specifically, in response to a known degraded condition impacting an RHRSW valve, Entergy did not adequately evaluate an associated temporary procedure change to EP-8, Alternate Injection Systems, to ensure operator actions could be accomplished to initiate RHRSW injection to the RPV within the PRA-credited time. Entergy entered the issue into their CAP as CR 2016-1396 and CR 2016-1429 and completed corrective actions to pre-stage a ladder for operator use and provide additional guidance to plant operators. The finding was more than minor because it was associated with the design control (plant modifications) attribute of the Mitigating Systems cornerstone and adversely affected the cornerstones objective of ensuring reliability, availability, and capability of systems and operators that respond to initiating events to prevent undesirable consequences (i.e., core damage). The team evaluated the finding in accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings at Power, Exhibit 2 Mitigating Systems Screening Questions, and concluded it required a detailed risk evaluation (DRE). A Region I Senior Reactor Analyst performed the DRE and concluded that the failure of an operator action to align RHRSW for RPV alternate injection within the assumed PRA accident mitigation time limit results in an estimated increase in core damage frequency in the mid E-8/year range, or very low safety significance (Green). The finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, because Entergy did not thoroughly evaluate issues to ensure that resolutions address causes and extent-of-conditions commensurate with their safety significance. Specifically, Entergy did not thoroughly evaluate the effect of an alternate injection procedure change on PRA-credited time critical operator actions. (PI.2) |
| 05000333/FIN-2016002-01 | 2016Q2 | Green | H.4 | Self-revealing | Failure to Determine Dose Rates Prior to Entering a High Radiation Area | The inspectors identified a self-revealing Green NCV of Technical Specification (TS) 5.7.1, High Radiation Area. Specifically, on January 24 and 25, 2016, operations personnel failed to notify the Radiation Protection (RP) department and non-licensed operators in the field when operating plant equipment that created high radiation areas (HRAs). These areas therefore were not surveyed by RP to determine dose rates prior to non-licensed operators entering the areas. Personnel entry into HRAs without knowledge of the current dose rates is a performance deficiency. In both instances, RP evaluated the operators dose, validated the dosimeter alarms, surveyed both areas in response to the dose rate alarms, and reposted the areas as HRAs. Entergy documented the events in condition reports (CR)-JAF-2016-00269 and CR-JAF-2016-00369 The finding was more than minor because it resulted in the unintended exposure of two workers and affected the Occupational Radiation Safety cornerstone attribute of program and process associated with exposure/contamination controls and if left uncorrected could result in more significant exposures. The finding was determined to be of very low safety significance (Green) because it was not related to as low as is reasonably achievable (ALARA), did not result in an overexposure or a substantial potential for overexposure, and did not compromise the licensee's ability to assess dose. A cross-cutting aspect of Human Performance, Teamwork, was associated with this finding. Specifically, licensed operators did not communicate to RP or non-licensed operators in the field when operating plant equipment that caused plant radiological conditions to change. (H.4) |
| 05000333/FIN-2016002-03 | 2016Q2 | Green | Licensee-identified | Licensee-Identified Violation | TS 3.6.3.2, Containment Atmosphere Dilution System, requires that, if one CAD subsystem is inoperable, then restore the subsystem to operable within 30 days or be in mode 3 within 12 hours. Contrary to the above, from June 17, 2015, to July 31, 2015, a period of 34 days, A CAD subsystem was inoperable without the plant being placed in mode 3 within 30 days and 12 hours of becoming inoperable. Also, contrary to the above, from August 5, 2015, to November 11, 2015, a period of 99 days, B CAD subsystem was inoperable without the plant being placed in mode 3 within 30 days and 12 hours of becoming inoperable. FitzPatrick staff entered this issue into their CAP as CR-JAF-2015-05453. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, the inspectors determined that this finding was of very low safety significance (Green) because the performance deficiency was not a design or qualification deficiency, did not involve an actual loss of a safety function of a single train for greater than its TS allowed outage time (because operator action could be taken to restore system function if the subject temperature transmitter failed), and did not screen as potentially risk-significant due to a seismic, flooding, or severe weather initiating event. | |
| 05000333/FIN-2016002-02 | 2016Q2 | Green | P.3 | NRC identified | Failure to Conduct Operations to Minimize the Introduction of Residual Radioactivity to the Site | The inspectors identified a Green NCV of Title 10 of the Code of Federal Regulations (10 CFR) 20.1406(c) due to Entergy not conducting operations to minimize the introduction of residual radioactivity into the site. For at least the past four years, Entergy allowed leakage of the solid radwaste processing system to occur, resulting in spilled radioactive waste that accumulated and remained on the floor of the filter sludge tank room in the radwaste building. The failure to control spilled radioactive wastes is a performance deficiency. Entergy entered this issue into their corrective action program (CAP) as CR-JAF-2016-01784 with actions to characterize the introduction of residual radioactivity and evaluate cleanup actions. This issue is more than minor because it is associated with the program and process attribute of the Public Radiation Safety cornerstone and affected the cornerstone objective to ensure the licensees ability to prevent inadvertent release and/or loss of control of licensed material. The finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Resolution, in that the condition was known to exist for over four years, impacted the radwaste system effectiveness to process solid radwaste, and had not been corrected. (P.3) |
| 05000333/FIN-2016001-04 | 2016Q1 | Severity level IV | NRC identified | Untimely 10 CFR 50.72 Notification of Inoperable Secondary Containment | The inspectors identified a SL IV NCV of 10 CFR 50.72, Immediate Notification Requirements for Operating Nuclear Power Reactors, because unplanned inoperability of the secondary containment system was not reported to the NRC within eight hours of the occurrence, as required by 10 CFR 50.72(b)(3)(v), Event or Condition That Could Have Prevented Fulfillment of a Safety Function. Specifically, following reasonable resolution of questions regarding the reliability of secondary containment differential pressure (d/p) instrumentation indications, FitzPatrick staff did not promptly report that, during a transfer from normal reactor building ventilation in service to the reactor building being isolated with the SBGTS in service, reactor building d/p briefly dropped below the TS required minimum value of 0.25 inches of vacuum water gauge and therefore caused the secondary containment system to be inoperable. As immediate corrective action, the event was reported to the NRC in accordance with 10 CFR 50.72(b)(3)(v). The issue was entered into the CAP as CR-JAF-2015-05244 and CR-JAF-2015-05265. The inspectors determined that the failure to inform the NRC of the secondary containment system inoperability within eight hours in accordance with 10 CFR 50.72(b)(3)(v) was a performance deficiency that was reasonably within Entergys ability to foresee and correct. The inspectors evaluated this performance deficiency in accordance with the traditional enforcement process because the issue impacted the regulatory process, in that a safety system functional failure was not reported to the NRC within the required timeframe, thereby delaying the NRCs opportunity to review the matter. Using Example 6.9.d.9 from the NRC Enforcement Policy, the inspectors determined that the violation was a SL IV (more than minor concern that resulted in no or relatively inappreciable potential safety or security consequence) violation, because Entergy personnel failed to make a report required by 10 CFR 50.72 when information that the report was required had been reasonably within their ability to have identified. In accordance with IMC 0612, Power Reactor Inspection Reports, traditional enforcement issues are not assigned cross-cutting aspects. | |
| 05000333/FIN-2016001-02 | 2016Q1 | Green | H.11 | NRC identified | Uncontrolled RPV Level Increase after Initiation of RHR Shutdown Cooling | The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to take actions specified in the procedure for initiation of shutdown cooling. Specifically, prior to placing the A loop of the residual heat removal (RHR) system into shutdown cooling, an operator was not stationed to close the condensate transfer system cross-connect valve to the A RHR loop (10RHR-274), nor was the valve immediately closed after initiation of shutdown cooling, as specified by the operating procedure. This resulted in a significant loss of operational control, in that RPV level increased to the point of putting water down the main steam lines. As immediate corrective action, operators closed 10RHR-274, thus stopping the RPV inventory increase. The issue was entered into the CAP as CR-JAF-2016-00273. The finding was more than minor because it was associated with the human performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the resultant loss of RPV level control represented a significant loss of operational control that could have affected the operability of the HPCI and reactor core isolation cooling (RCIC) systems, as well as the S/RVs, had their use again been required in the near term. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, the inspectors determined that this finding was of very low safety significance (Green) because the performance deficiency was not a design or qualification deficiency, did not involve an actual loss of a safety function of a single train for greater than its TS allowed outage time, and did not screen as potentially risk-significant due to a seismic, flooding, or severe weather initiating event. The finding had a cross-cutting aspect in the area of Human Performance, Challenge the Unknown, because operators did not stop when faced with uncertain conditions. Specifically, without otherwise having maintained status control on the condensate transfer system cross-connect valve to the A RHR loop, operators did not stop to positively establish the condition of the valve when it appeared in a conditional step in the procedure (that is, if 10RHR-274 is open, then station an operator at 10RHR-274) (H.11). |
| 05000333/FIN-2016001-03 | 2016Q1 | Green | H.1 | Self-revealing | Inadequate Post-Maintenance Testing of the Reactor Building Ventilation System Resulted in Short-Term Inoperability of Secondary Containment | The inspectors identified a self-revealing NCV of TS 5.4, Procedures, for FitzPatrick staffs failure to perform adequate post-maintenance testing (PMT) following maintenance on a limit switch in the reactor building ventilation system in August 2014, that, along with another unrelated component failure in the reactor building ventilation system, resulted in secondary containment pressure, relative to the outside pressure, exceeding the TS limit of 0.25 inches of vacuum water gauge. As immediate corrective action, operators started both trains of the standby gas treatment system (SBGTS), which restored secondary containment pressure to within the TS limit. Operators subsequently secured the A refuel floor exhaust train and placed the B train in service. The issue was entered into the CAP as CR-JAF-2015-04166. The finding was more than minor because it was associated with the configuration control attribute of the Barrier Integrity cornerstone and affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, as a result of this event, secondary containment was not preserved, in that secondary containment pressure exceeded the limit of TS surveillance requirement (SR) 3.6.4.1.1. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 3 of IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, the inspectors determined that this finding was of very low safety significance (Green) because the performance deficiency was not a pressurized thermal shock issue, did not represent an actual open pathway in the physical integrity of the reactor containment, did not involve an actual reduction in function of hydrogen igniters in the reactor containment, and only represented a degradation of the radiological barrier function provided by the reactor building and SBGTS. The finding had a cross-cutting aspect in the area of Human Performance, Resources, because FitzPatrick staff did not ensure that procedures for PMT of the reactor building refuel floor exhaust damper limit switch following maintenance performed in August 2014, were adequate to support the nuclear safety function of the secondary containment (H.1). |
| 05000333/FIN-2016001-01 | 2016Q1 | Green | H.8 | NRC identified | Unintended HPCI Pump Suction Transfer during Pressure Control Mode Operation | The inspectors identified a Green NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to maintain a condition specified in an emergency operating procedure. Specifically, while operating the high pressure coolant injection (HPCI) system in the pressure control mode, operators failed to override automatic transfer of the HPCI pump suction from the condensate storage tank (CST) to the suppression pool prior to the transfer actually occurring. As a result, operators had to revert to using the safety/relief valves (S/RVs) for pressure control, which introduced additional, unnecessary plant challenges. As immediate corrective action, operators secured HPCI, overrode the automatic HPCI pump suction transfer, realigned the pump suction to the CST, and restarted HPCI in the pressure control mode. The issue was entered into the corrective action program (CAP) as condition report (CR)-JAF-2016-00765. The finding was more than minor because it was associated with the human performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the operators failure to timely override automatic transfer of the HPCI suction to the suppression pool resulted in an additional, avoidable post-scram pressure and level transient being placed on the reactor pressure vessel (RPV) and unnecessarily reduced the thermal capacity of the suppression pool. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At- Power, the inspectors determined that this finding was of very low safety significance (Green) because the performance deficiency was not a design or qualification deficiency, did not involve an actual loss of a safety function of a single train for greater than its technical specification (TS) allowed outage time, and did not screen as potentially risksignificant due to a seismic, flooding, or severe weather initiating event. The finding had a cross-cutting aspect in the area of Human Performance, Procedure Adherence, because operators did not follow guidance of EOP-2 for the HPCI pump suction to be aligned to the CST by bypassing the HPCI pump suction swap to the suppression pool in a timely manner, such that the swap actually occurred (H.8). |
| 05000333/FIN-2015004-03 | 2015Q4 | Green | Licensee-identified | Licensee-Identified Violation | TS 3.4.3 requires that at least nine S/RVs shall be operable in operating modes 1, 2, and 3. Contrary to this, on June 1, 2015, FitzPatrick personnel identified that the plant had operated in these modes during cycle 21 with less than nine operable S/RVs. FitzPatrick personnel documented this condition in CR-JAF-2015-02493. The inspectors determined this TS violation was of very low safety significance (Green) because it did not result in the loss of the overpressure relief safety function based on operability of the electric lift system. | |
| 05000333/FIN-2015004-01 | 2015Q4 | Green | H.5 | NRC identified | Unintended Elevated Plant Risk During EDG Maintenance | The inspectors identified a Green NCV of Title10 of the Code of Federal Regulations (10 CFR) 50.65, Requirements for monitoring the effectiveness of maintenance at nuclear power plants, for failure to adequately manage the increase in risk during planned maintenance on the A emergency diesel generator (EDG). Specifically, Entergy staff action to make the C EDG unavailable while the A EDG was already unavailable resulted in an unplanned increase in overall plant risk and deviation from the approved EDG outage risk management plan from a risk category of Green to the next higher risk category of Yellow. As immediate corrective action, the issue was entered into the corrective action program (CAP) as condition report (CR)-JAF-2015-05242. The finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the C EDG was not available when it should have been, in accordance with the approved risk management plan, which resulted in an unplanned escalation of risk from Green to Yellow. Additionally, this finding was similar to example 7.e in IMC 0612, Appendix E, Examples of Minor Issues. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At- Power, the inspectors determined that this finding was of very low safety significance (Green) because the performance deficiency was not a design or qualification deficiency, did not involve an actual loss of safety function, did not represent actual loss of a safety function of a single train for greater than its Technical Specification (TS) allowed outage time, and did not screen as potentially risk-significant due to a seismic, flooding, or severe weather initiating event. The finding had a cross-cutting aspect in the area of Human Performance, Work Management, because FitzPatrick did not execute the A EDG maintenance outage work activities as planned, and after deviating from that plan, did not identify and manage the risk of barring the C EDG while the A EDG was unavailable (H.5). |
| 05000333/FIN-2015004-02 | 2015Q4 | Severity level IV | NRC identified | Untimely 10 CFR 50.72 Notification of Inoperable Secondary Containment | The inspectors identified a Severity Level (SL) IV NCV of 10 CFR 50.72, Immediate Notification Requirements for Operating Nuclear Power Reactors, because inoperability of the secondary containment system was not reported to the NRC within eight hours of when the need to do so should reasonably have been recognized, as required by 10 CFR 50.72(b)(3)(v), Event or Condition that Could Have Prevented Fulfillment of a Safety Function. Specifically, positive pressure in the secondary containment due to a previously unidentified equipment malfunction that occurred during transition between the reactor building being isolated and normal reactor building ventilation being in service was not promptly recognized as a condition that caused the single train secondary containment system to be inoperable and therefore to be reportable under 10 CFR 50.72. This issue was entered into the CAP as CR-JAF-2015-05244 and CR-JAF-2015-05265. The inspectors determined that the failure to inform the NRC of the secondary containment system inoperability within eight hours in accordance with 10 CFR 50.72(b)(3)(v) was a performance deficiency that was reasonably within Entergys ability to foresee and correct. The inspectors evaluated this performance deficiency in accordance with the traditional enforcement process because the issue impacted the regulatory process, in that a safety system functional failure was not reported to the NRC within the required timeframe, thereby delaying the NRCs opportunity to review the matter. Using Example 6.9.d.9 from the NRC Enforcement Policy, the inspectors determined that the violation was an SL IV (more than minor concern that resulted in no or relatively inappreciable potential safety or security consequence) violation, because Entergy personnel failed to make a report required by 10 CFR 50.72 when information that the report was required had been reasonably within their ability to have identified. In accordance with IMC 0612, Power Reactor Inspection Reports, traditional enforcement issues are not assigned cross-cutting aspects. | |
| 05000333/FIN-2015003-01 | 2015Q3 | Green | Self-revealing | Inadequate Corrective Actions Result in Control Rod Drift and Reactor Power Reduction | A self-revealing NCV of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion XVI, Corrective Action, was identified because FitzPatrick staff failed to correct a condition adverse to quality. Specifically, Entergy failed to take effective corrective actions for condition report (CR)-JAF-2010-00287 to replace the control rod drive (CRD) hydraulic control unit (HCU) directional control valve (DCV) bolting material which had signs of corrosion after the same material was identified through operational experience as the cause of a control rod drift. As a result, on July 19, 2015, FitzPatrick control rod 10-07 drifted from the fully withdrawn to the fully inserted position in the reactor core leading to an immediate power reduction from 100 to 99 percent followed by a manual rapid power reduction to 56 percent. Entergys subsequent corrective actions included an extent of condition review and completed or planned replacement of all susceptible directional control valve bolting. The performance deficiency was determined to be more than minor because it was associated with the equipment performance attribute of the Initiating Events cornerstone, and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors determined that this finding was of very low safety significance (Green) using Exhibit 1 of IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, because the finding did not cause both a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition (e.g. loss of condenser, loss of feed water). The inspectors determined that there was no crosscutting aspect associated with this finding because the cause of the performance deficiency occurred more than three years ago, and was not representative of current plant performance. | |
| 05000333/FIN-2015003-02 | 2015Q3 | Green | H.12 | Self-revealing | Inadequate Instructions for Reactor Building Roof Replacement Result in Inadvertent Loss of Secondary Containment | The inspectors identified a self-revealing violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because FitzPatrick staff failed to provide instructions appropriate to the reactor building roof replacement project. Specifically, inadequate instructions were provided to ensure that roofing material removal would be performed in slow, deliberate manner, such that its effect on secondary containment could be assessed and operability maintained. As a result, this activity caused secondary containment to be inoperable for a period in excess of its four hour technical specification (TS) allowed outage time. As immediate corrective action, roofing material removal was stopped and the new roofing materials were installed to reseal the affected area of the reactor building roof. Secondary containment vacuum was restored to greater than the TS-required minimum after a period of 92 minutes and secondary containment was declared operable after a period of five hours and 26 minutes. The issue was entered into the corrective action program (CAP) as CRJAF- 2015-03260. The finding was more than minor because it is associated with the procedure quality attribute of the Barrier Integrity cornerstone, and affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system (RCS), and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the work order (WO) did not provide adequate instruction to ensure that roofing material removal would be performed in slow, deliberate manner, coordinated between operations and maintenance personnel, and allowing adequate time after actions that could impact secondary containment such that their effect on secondary containment could be assessed and operability maintained. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 3 of IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At- Power, the inspectors determined that this finding was of very low safety significance (Green) because the performance deficiency was not a pressurized thermal shock issue, did not represent an actual open pathway in the physical integrity of the reactor containment, did not involve an actual reduction in function of hydrogen igniters in the reactor containment, and only represented a degradation of the radiological barrier function provided by the reactor building and standby gas treatment system. The finding had a cross-cutting aspect in the area of Human Performance, Avoid Complacency, because FitzPatrick staff did not adequately plan for the possibility of latent issues and inherent risk associated with the reactor building roof replacement project, such that the commencement of work resulted in a loss of secondary containment. |
| 05000333/FIN-2015007-01 | 2015Q2 | Green | H.14 | NRC identified | Failure to Adequately Assess the Impact of SRV Leakage on Operability | The inspectors identified a Green, non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion III, Design Control, associated with FitzPatricks failure to adequately assess and control the acceptance criteria specified in engineering analysis in EC-JAF-56258, Operability Input for CR-JAF-2015-01271 SRV G Tailpipe Temperature Increase, which referenced JAF-RPT-03-0056 Operational Leakage Action Levels for Target Rock Two-Stage Safety/Relief Valves. Specifically, FitzPatrick concluded that a 2-stage Target Rock Safety Relief Valve (SRV) was operable with pilot valve leakage provided the leak rate was less than 1000 lbm/hr. This conclusion was not adequately supported by the available industry and plant data on setpoint drift and the references provided. As a result, FitzPatrick did not declare 2-stage Target Rock Pilot valves inoperable when the leak rate exceeded 600 lbm/hr in 2007 and 2009. FitzPatrick entered this issue into the corrective action system (CR-JAF-2015-02850) and is reassessing the appropriate operability criteria. This performance deficiency is more than minor because it adversely affects the equipment performance attribute of the initiating events cornerstone in IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, to limit the likelihood of events that upset plant stability and challenge critical safety functions during power operations by ensuring reactor coolant system (RCS) barrier integrity. This finding screens to Green using IMC 0609, Significance Determination Process, Attachment 4, Initial Characterization of Findings, and IMC 0609, Appendix A, Exhibit 1, Initiating Events Screening Questions, Section A, LOCA Initiators, as the finding could not result in leakage exceeding that of a small break loss-of-coolant accident (LOCA) nor could it have resulted in an interfacing system LOCA. The inspectors determined that this performance deficiency had a crosscutting aspect in human performance, conservative bias, where individuals use decision making-practices that emphasize prudent choices over those that are simply allowable. (H.14) Section 1R17 |
| 05000333/FIN-2015001-01 | 2015Q1 | Green | H.8 | Self-revealing | Incomplete Fuel Support Piece Seating Not Identified During Post-Refueling Core Verification | A self-revealing, Green NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50 Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified because the existence of a partially seated fuel support piece at reactor cell location 38-39 was not identified when FitzPatrick staff performed the procedure for reactor core verification at the conclusion of refueling operations during the 2014 refueling outage (RO21). Specifically, the fact that the four fuel assemblies associated with cell 38-39 were elevated by an estimated 1.5 inches above the top of the rest of the fuel assemblies in the reactor core was not identified during visual verification of fuel assembly seating performed after the conclusion of core alterations in accordance with procedure EN-RE-210, BWR (boiling water reactor) Reactor Core and MPC (multi-purpose canister) Cask Fuel Verification. As immediate corrective action, FitzPatrick staff engaged the fuel vendor, who provided an interim thermal limit penalty to be applied to the four affected fuel assemblies pending completion of a formal analysis. The issue was entered into FitzPatrick s CAP as CR-JAF- 2015-00789. The finding was more than minor because it was associated with the Configuration Control attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the fuel support piece not being completely fitted into the top of the control rod guide tube resulted in increased bypass flow around the cell 38-39 fuel assemblies, which reduced the margin to thermal limits for these assemblies during normal, transient, and accident conditions. Since the performance deficiency associated with the finding occurred during shutdown operations and also had potential safety significance during normal at-power operations, the inspectors screened the finding for significance using both IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, and IMC 0609 Appendix G, Attachment 1, Shutdown Operations Significance Determination Process. The inspectors determined that the finding was of very low safety significance (Green) because the displaced fuel bundles did not have any negative impact on safety during shutdown conditions, and through application of a thermal limit penalty, did not negatively impact the safe operation of the reactor at power. This finding had a cross-cutting aspect in the area of Human Performance, Procedure Adherence, because FitzPatrick staff did not follow the procedure requirement for reactor core verification to verify that the tops of the fuel channels and bail handles were all at approximately the same height. |
| 05000333/FIN-2015001-02 | 2015Q1 | Green | H.1 | Self-revealing | Inadequate Preventive Maintenance Strategy and Test Procedure for RWR MG Resulted in Multiple Plant Transients | A self-revealing, Green NCV of Technical Specification (TS) 5.4, Procedures, was identified for failure to institute appropriate processes and procedures for periodic maintenance activities of the reactor water recirculation motor generators (RWR MGs). During startup from refueling outage 21, degraded material conditions led to tripping of an RWR MG, with the resultant loss of the associated RWR pump and down power transient, on three occasions. Specifically, one trip was due to carbon dust buildup within the A RWR MG exciter, and two trips were due to a high resistance connection between the B RWR MG generator field winding and a slip ring. Additionally, a fourth trip occurred during performance of an inadequately prepared RWR MG test procedure. As corrective action, the high resistance connection associated with the B RWR MG was eliminated, voltage regulator tuning for the B RWR MG was successfully completed, and temporary instrumentation was connected to both RWR MGs to monitor various key parameters pending the implementation of long term corrective actions. The RWR MG trips were entered into the corrective action program (CAP) through individual condition reports (CRs) that were subsequently consolidated under CR-JAF-2014-06258 for root cause evaluation (RCE). The finding was more than minor because it was associated with the Equipment Performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the transient initiation of single RWR loop operations challenges the reactor feedwater and vessel level control systems such that a more significant plant transient could result, and challenges plant operators in establishing allowable single RWR loop operating conditions. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 1 of IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, the inspectors determined that the finding was of very low safety significance (Green) because the performance deficiency was a transient initiator that did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The finding had a cross-cutting aspect in the area of Human Performance, Resources, because FitzPatrick staff did not ensure that procedures for RWR preventive maintenance (PM) and voltage regulator tuning were adequate to support nuclear safety. |
| 05000333/FIN-2015405-01 | 2015Q1 | Severity level Enforcement Discretion | NRC identified | Security | ||
| 05000333/FIN-2014005-01 | 2014Q4 | Severity level IV | NRC identified | Untimely 10 CFR 50.72 Notification of a Secondary Containment System Functional Failure | Severity Level IV. The inspectors identified an SL IV NCV of Title10 of the Code of Federal Regulations (10 CFR) 50.72, Immediate Notification Requirements for Operating Nuclear Power Reactors, because unplanned inoperability of the secondary containment system was not reported to the NRC within eight hours of the occurrence, as required by 10 CFR 50.72(b)(3)(v), Event or Condition that Could Have Prevented Fulfillment of a Safety Function. Specifically, while restoring the normal reactor building ventilation (RBV) syste to service following maintenance, reactor building-to-ambient differential pressure droppe below the Technical Specification (TS) required minimum value of 0.25 inches of vacuum water gauge and therefore caused the secondary containment system to be inoperable. However, FitzPatrick staff did not promptly recognize this as a condition reportable under 10 CFR 50.72. As corrective action, FitzPatrick staff reported the condition to the NRC in accordance with 10 CFR 50.72(b)(3)(v) and entered it into the corrective action progra (CAP) as condition report (CR)-JAF-2014-06498. The inspectors determined that the failure to inform the NRC of the secondary containment system inoperability within eight hours in accordance with 10 CFR 50.72(b)(3)(v) was a performance deficiency that was reasonably within Entergys ability to foresee and correct. The inspectors evaluated this performance deficiency in accordance with the traditional enforcement process because the issue impacted the regulatory process, in that a safety system functional failure was not reported to the NRC within the required timeframe, thereby delaying the NRCs opportunity to review the matter. Using Example 6.9.d.9 from the NRC Enforcement Policy, the inspectors determined that the violation was a SL IV (more than minor concern that resulted in no or relatively inappreciable potential safety or security consequence) violation, because Entergy personnel failed to make a report required by 10 CFR 50.72 when information that the report was required had been reasonably within their ability to have identified. In accordance with IMC 0612, Power Reactor Inspection Reports, traditional enforcement issues are not assigned cross-cutting aspects. | |
| 05000333/FIN-2014005-02 | 2014Q4 | Green | P.3 | NRC identified | TS Actions for Inoperable ECCS Not Performed Within the TS Allowed Completion Time | The inspectors identified a Green NCV for two violations of TS 3.5.1, ECCS (emergency core cooling systems) - Operating, associated with the non-functionality of east crescent area ventilation and cooling (CAVC) subsystem unit cooler 66UC-22H. Specifically, during the periods May 5 through May 21, 2010, and March 15 through March 25, 2011, the Technical Requirements Manual (TRM) requirements for east crescent unit cooler operability were not satisfied for longer than the allowed outage time (AOT), which caused the ECCS in the east crescent to become inoperable and remain so for longer than the TS AOT without completion of the required plant mode changes. As immediate corrective action, Entergy personnel reconditioned the fan motor contactor for the affected unit cooler to obtain satisfactory low voltage pickup response. The issue was entered into Entergys CAP as CR-JAF-2012-00584 and CR-JAF-2012-02288. The finding was more than minor because it is associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the unsatisfactory low voltage response of the 66UC-22H fan motor contactor, along with the unavailability of another east CAVC unit cooler due to maintenance, could have degraded the capability of ECCS systems in the east crescent area during an accident concurrent with degraded voltage conditions. In light of FitzPatrick staffs determination that there was reasonable assurance that the remaining three operable unit coolers would have been capable o removing required post-accident heat loads, the inspectors determined that the finding wa of very low safety significance (Green) in accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The Significanc Determination Process for Findings At-Power, because the performance deficiency was no a design or qualification deficiency, did not involve an actual loss of safety function, did not represent the actual loss of a safety function of a single train for greater than its TS AOT, and did not screen as potentially risk-significant due to a seismic, flooding, or sever weather initiating event. This finding had a cross-cutting aspect in the area of Proble Identification and Resolution, Resolution, because FitzPatrick staff did not take effectiv corrective actions to address the low voltage pickup issue in a timely manner commensurat with its safety significance (P.3). |
| 05000333/FIN-2014201-02 | 2014Q4 | Green | NRC identified | Security | ||
| 05000333/FIN-2014005-03 | 2014Q4 | Green | Licensee-identified | Licensee-Identified Violation | TS 3.3.5.2, Reactor Core Isolation Cooling (RCIC) System Instrumentation, requires that the RCIC system instrumentation for all 4 channels of low CST water level be operable while in Modes 1, 2, or 3 with reactor steam dome pressure greater than 150 psig. With any level switch inoperable, Condition D requires that the channel be placed in trip. When this condition is not met, Condition E requires that RCIC be declared inoperable. LCO 3.5.3, RCIC System, further requires that RCIC be restored to operable status within 14 days or be placed in Mode 3. Contrary to TS 3.3.5.2, with two RCIC CST level switches, 13LS-76B and -77B, inoperable from July 16, 2013 to August 19, 2013, Entergy did not place the channels in trip or declare RCIC inoperable, or place the reactor in Mode 3 per TS 3.5.3. The cause o the inoperability was corrosion buildup on the level switches caused by water intrusion through a junction box common to the switches. Entergy entered this issue into the CAP as CR-JAF-2013-04311. The inspectors determined, through a review of IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, Exhibit 2, that the finding was of very low safety significance (Green) because the finding was not related to a design or qualification deficiency, did not represent a loss of a mitigating system safety function, and did not screen as potentially risk significant due to external initiating events. The Senior Reactor Analyst (SRA) used the Systems Analysis Programs for Hands-On Evaluatio (SAPHIRE), Revision 8.1.2, and the Standardized Plant Analysis Risk (SPAR) Model for Fitzpatrick, Model Version 8.1.17, to confirm that no loss of safety function occurred. The SRA determined that the RCIC pump suction is assumed to remain on the CST for the duration of operation to complete its safety function and therefore this issue was determined to be of very low safety significance (Green). The CST inventory is modeled to be sufficient because the function of RCIC is to respond to transient events to provide makeup coolant to the reactor. | |
| 05000333/FIN-2014008-01 | 2014Q4 | Green | Licensee-identified | Licensee-Identified Violation | The team determined that the condition identified in LER 50-333/2013-003 constituted a Green violation of Condition 2.C(3) of Renewed Facility Operating License DPR-59, in that Entergy did not ensure that equipment required for post-fire safe shutdown was protected from the effects of a postulated fire. Specifically, Entergy did not provide overcurrent protection for wiring associated with DC ammeter indication in the control room to prevent the wires from overheating due to fire-induced faults. The issue was identified by Entergy during review of industry operating experience. Description: During JAFNPPs review of operating experience information from the DavisBesse and Cooper stations, plant staff determined that the condition pertains to JAFNPP. The stations review determined that the plant wiring design for the station battery ammeter circuits contains a shunt in the current flow from each DC battery, with leads to an ammeter in the main control room. The ammeter wiring attached to the shunts does not have overcurrent protection devices, and if one of the ammeter wires shorts to ground during a fire at the same time another DC wire from the opposite polarity on the same battery also shorts to ground (as a result of the fire), a ground loop through the unprotected ammeter cable could occur. With enough current flowing through the cable, the potential exists that the overloaded ammeter wiring could damage safe shutdown wiring in physical contact with the cable resulting in a loss of the associated safe shutdown component or a secondary fire in another fire area. The cause of the condition is that the original design criteria did not specify overcurrent protection for shunt fed ammete circuits. Renewed Facility Operating License DPR-59, condition 2.C(3) requires Entergy t implement and maintain the fire protection program as documented in the Updated Final Safety Analysis Report and approved in various safety evaluation reports (SERs) including the SER dated August 1, 1979. Section 5.11 of the SER dated August 1, 1979 evaluates the Battery Charger Rooms, and concluded that with the proposed modifications (installation of fire detection and signaling systems), the areas will meet the objectives set out in Section 2.2 of the SER, and are therefore acceptable. The objectives set out in Section 2.2 include (3) maintain the capability to safely shut down the plant if fires occur. Entergy did not meet this requirement, and did not protect other safe shutdown equipment from the effects of postulated fires. Entergy initiated condition report CR-JAF-2013-05546 for evaluation and resolution of the condition. Entergy reported the issue to NRC on October 31, 2013 via event notification 49491 in accordance with 10 CFR 50.72, and initiated LER 50-333/2013-003 to report the issue to NRC in accordance with 10 CFR 50.73. Entergy implemented compensatory measures in accordance with their technical requirements manual for degraded barriers to prevent the spread of a fire to other areas. Entergy subsequently initiated design change EC 48868 to install fuses in the DC ammeter circuits. The team concluded that Entergys interim compensatory measures were commensurate with the risk significance of the issue. Analysis: Entergys failure to protect safe shutdown cables from the effect of postulated fires was a performance deficiency. This issue is more than minor because it is associated with the External Factors attribute (fire) of the Mitigating Systems Cornerstone and adversely affects the cornerstone objective to ensure the availability, reliability, and capability of systems to prevent undesirable consequences. Specifically, the availability of safe shutdown cables was not ensured for fires in the Battery Charger Rooms, Battery Room Corridor, Cable Spreading Room, Relay Room, and Control Room. As stated previously, Entergy entered this issue into the corrective action program and promptly initiated compensatory measures in accordance with their fire protection program. Inspection Manual Chapter (IMC) 0609, Appendix F, Fire Protection Significance Determination Process, was used to evaluate the issue. Because the spread of fire to other areas if the cable ignited, the condition was considered a high degradation of the fire confinement category. As a result, the issue did not screen in Phase 1. Phase 2 requires the development of credible fire scenarios. An NRC inspector reviewed cable routing information for the ammeter cabling and walked down the cable routing in the plant. The inspector determined that there were no in-situ or transient combustible materials which could result in a credible exposure fire to cause damage to the ammeter cables. This meets the screening criteria of IMC 0609, Appendix F, Step 2.3.5. Therefore, the inspector determined that this licensee-identified violation would be of very low safety significance (Green). In addition, the team concluded that Entergys interim compensatory measures for degraded fire barriers until final resolution of the issue is completed wa commensurate with the risk significance. Cross-cutting aspects are not applicable to ol design issues which do not represent current performance. Enforcement: Condition 2.C(3) of Renewed Facility Operating License DPR-59 requires that Entergy implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report for the facility and as approved in various SERS, including the SER dated August 1, 1979. The August 1, 1979, SER indicates that with the proposed modifications to the battery charger rooms, the ability to safely shut down the plant will be maintained. Contrary to the above, on October 31, 2013, Entergy identified that they did not meet this requirement for the Battery Charger Rooms, Battery Room Corridor, Cable Spreading Room, Relay Room and Main Control Room fire zones and failed to maintain post-fire shutdown cables free of the effects of fire induced cable faults during postulated fires. This issue existed since the plant was constructed in the early 1970s. In addition to the above licensee identified violation the inspector noted that Entergy identified the issue as a result of its operating experience review program. Interim compensatory measures were implemented immediately on discovery, and long-term actions are in progress to correct the deficiency. Entergy is performing an extent of condition review to determine if other DC circuits are susceptible to similar overcurrent conditions as a result of fire-induced circuit faults. The condition was not likely to be identified by other efforts such as normal testing or routinely scheduled reviews such as Quality Assurance or fire protection program reviews. This issue is not linked to current Entergy performance, but rather, is tied to design standards and construction activities in the 1960s and 1970s and there was no prior notice so that Entergy would not have reasonably identified the condition earlier. | |
| 05000333/FIN-2014201-01 | 2014Q4 | Green | NRC identified | Security | ||
| 05000333/FIN-2014004-02 | 2014Q3 | Green | Licensee-identified | Licensee-Identified Violation | TS 3.3.5.2, Reactor Core Isolation Cooling System Instrumentation, requires that the RCIC system instrumentation for all four channels of low CST water level be operable while in Modes 1, 2, or 3 with reactor steam dome pressure greater than 150 psig. With one level switch inoperable, Condition D requires that the channel be placed in trip. When this condition is not met, Condition E requires that RCIC be declared inoperable. TS 3.5.3, RCIC System, further requires that RCIC be restored to operable status within 14 days or be in Mode 3. Contrary to TS 3.3.5.2, with one RCIC CST level switch, 13LS-76B, inoperable from September 17, 2013 until November 4, 2013, Entergy did not place the channel in trip or declare RCIC inoperable, or place the reactor in Mode 3 per TS 3.5.3. The cause of the inoperability was the failure to align the microswitch in accordance with vendor manual instructions when the switch was replaced in September. Entergy entered this issue into the CAP as CR-JAF-2013-5576. The inspectors determined that the finding was of very low safety significance (Green) in accordance with IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, because the finding was not a design or qualification deficiency, did not involve the actual loss of safety function, did not represent the actual loss of a safety function of a single train for greater than its TS allowed outage time, and did not screen to potentially risk significant due to a seismic, flooding, or severe weather initiating event. | |
| 05000333/FIN-2014004-01 | 2014Q3 | Severity level IV | NRC identified | Failure to Notify NRC Within 30 Days of Medical Changes for Licensed Operators | The inspectors identified a Severity Level (SL) IV NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50.74, Notification of Change in Operator or Senior Operator Status. Specifically, on three occasions, Entergy staff did not notify the NRC of a change in the medical status of a licensed operator within 30 days of learning of the diagnosis. These issues were entered into the corrective action program (CAP) as condition report (CR)-JAF-2014-02227 and CR-JAF-2014-02304. The inspectors determined that Entergys failure to notify the NRC of licensed operator medical status changes as described above within 30 days was a performance deficiency that was within Entergys ability to foresee and correct and should have been prevented. Because the issue had the potential to affect the NRCs ability to perform its regulatory function, the inspectors evaluated this performance deficiency in accordance with the traditional enforcement process. Using example 6.4.d.1(b) from the NRC Enforcement Policy, the inspectors determined that the violation was a Severity Level IV (more than minor concern that resulted in no or relatively inappreciable potential safety or security consequence) violation because Entergy staff did not communicate licensed operator permanent medical status changes within the 30 day reporting requirement for three licensed operators. In accordance with IMC 0612, Power Reactor Inspection Reports, traditional enforcement issues are not assigned cross-cutting aspects. | |
| 05000333/FIN-2014003-01 | 2014Q2 | Green | H.7 | NRC identified | Failure to Properly Ship RAM-QC | The inspectors identified a Green NCV of 10 CFR 71.5, Transportation of Licensed Material, and 49 CFR 172, Subpart I, Safety and Security Plans. Specifically, Entergy personnel shipped a radioactive quantity of category 2 Radioactive Material in Quantities of Concern (RAM-QC) on the public highways to a waste processor without adhering to its transportation security plan. Prior to shipment, Entergy staff failed to recognize that the quantity of radioactive material met the definition of RAM-QC. Entergy staff entered this issue into their corrective action program (CAP) as condition report (CR)-JAF-2014-02337. The issue was more than minor because it was associated with the Program and Process attribute of the Public Radiation Safety cornerstone and affected the cornerstone objective to ensure adequate protection of public health and safety from exposure to radioactive materials released into the public domain. In accordance with IMC 0609, Appendix D, "Public Radiation Safety Significance Determination Process," the finding was determined to be of very low safety significance (Green) because Entergy had an issue involving transportation of radioactive material, but it did not involve: (1) a radiation limit exceeded; (2) a breach of package during transport; (3) a certificate of compliance issue; (4) a low level burial ground nonconformance; or (5) a failure to make notifications or provide emergency information. This finding had a cross-cutting aspect in the area of Human Performance, Work Processes, in that the documentation (procedures) to support this activity was inadequate (H.7). |
| 05000333/FIN-2014403-01 | 2014Q1 | Green | H.3 | NRC identified | Security | |
| 05000333/FIN-2014002-02 | 2014Q1 | Green | Licensee-identified | Licensee-Identified Violation | With a reactor building ventilation radiation monitor inoperable, TS 3.3.6.2 Condition A requires placing the associated secondary containment isolation instrumentation channel in trip within 24 hours. If Condition A is not met, Condition C requires that the reactor building ventilation system be isolated or the associated secondary containment isolation valves be declared inoperable, and that the SGT system be placed in operation within one hour. Contrary to the above, at 7:40 a.m. on November 6, 2013, the A reactor building ventilation radiation monitor was declared inoperable for maintenance, but the associated secondary containment isolation instrumentation channel was not placed in trip within 24 hours, the reactor building ventilation system was not isolated, the associated secondary containment isolation valves were not declared inoperable, and the SGT system was not placed in operation within one hour. The cause of this TS violation was human error. Entergy staff entered this issue into the CAP as CR-JAF-2013-05676. The inspectors determined that the finding was of very low safety significance (Green) in accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, because the finding only represents a degradation of the radiological barrier function provided for the secondary containment. | |
| 05000333/FIN-2014002-01 | 2014Q1 | Green | NRC identified | HPCI System Inoperable for Longer than Allowed by TS | The inspectors identified a Green NCV of Technical Specification (TS) 3.5.1, ECCS (emergency core cooling system) - Operating, because filling the high pressure coolant injection (HPCI) system with low quality water from the suppression pool following maintenance caused the HPCI booster pump recirculation pressure control valve, 23PCV-50, to fail, thereby making the HPCI system inoperable, and this condition existed for greater than the TS allowed outage time of 14 days. Although the HPCI system was inoperable, it still maintained its safety function to provide emergency core coolant flow in the event of an accident. As corrective action, Entergy staff changed the procedure to indicate that the HPCI system should be filled using the CSTs, and submitted revision 1 to the associated licensee event report (LER) to report the TS violation. This issue was entered into the corrective action program (CAP) as condition report (CR)-JAF-2014-00961. The inspectors determined that Entergy staffs actions to refill the HPCI system with water from the suppression pool following maintenance, thereby causing the failure of 23PCV-50 to control pressure the next time that the HPCI system was operated, was a performance deficiency that was reasonably within Entergys ability to foresee and correct. The finding was more than minor because it was associated with the human performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the issue resulted in failure of 23PCV-50 to control pressure, which caused the HPCI system to be inoperable for greater than its TS allowed outage time. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, the inspectors determined that this finding was of very low safety significance (Green) because the performance deficiency was not a design or qualification deficiency, did not involve an actual loss of safety function, did not represent the actual loss of a safety function of a single train for greater than its TS allowed outage time, and did not screen as potentially risk-significant due to a seismic, flooding, or severe weather initiating event. This finding had a cross-cutting aspect in the area of Problem Identification and Resolution, because FitzPatrick staff did not implement internal and external operating experience concerning the inadvisability of using suppression pool water to refill the HPCI system following maintenance (P.5) | |
| 05000333/FIN-2013005-01 | 2013Q4 | Severity level IV | NRC identified | Untimely 10 CFR 50.72 Notification of a HPCI System Functional Failure | The inspectors identified a Severity Level IV (SL IV) NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50.72, Immediate Notification Requirements for Operating Nuclear Power Reactors, because unplanned inoperability of the high pressure coolant injection (HPCI) system was not reported to the NRC within eight hours of when it should reasonably have been discovered, as required by 10 CFR 50.72(b)(3)(v), Event or Condition that Could Have Prevented Fulfillment of a Safety Function. Specifically, identification that issues with two of the condensate storage tank (CST) level detectors that provide automatic transfer of the HPCI suction from the CSTs to the suppression pool would have caused this transfer to occur at less than the minimum CST level allowed by Technical Specifications (TSs) and therefore caused the HPCI system to be inoperable, was not promptly recognized as a condition reportable under 10 CFR 50.72. This issue was entered into the corrective action program (CAP) as condition report (CR)-JAF-2013-06344. The inspectors determined that the failure to inform the NRC of the HPCI system inoperability within eight hours in accordance with 10 CFR 50.72(b)(3)(v) was a performance deficiency that was reasonably within Entergys ability to foresee and correct. Because the issue impacted the regulatory process; in that, a safety system functional failure was not reported to the NRC within the required timeframe thereby delaying the NRCs opportunity to review the matter, the inspectors evaluated this performance deficiency in accordance with the traditional enforcement process. Using example 6.9.d.9 from the NRC Enforcement Policy, the inspectors determined that the violation was a SL IV (more than minor concern that resulted in no or relatively inappreciable potential safety or security consequence) violation, because Entergy personnel failed to make a report required by 10 CFR 50.72 when information that the report was required had been reasonably within their ability to have identified. In accordance with IMC 0612, Power Reactor Inspection Reports, traditional enforcement issues are not assigned cross-cutting aspects. | |
| 05000333/FIN-2013004-01 | 2013Q3 | Green | H.7 | Self-revealing | Inadequate Reactor Water Recirculation Digital Flow Control Modification Post Maintenance Test Procedure Results in Unexpected Power Increase | The inspectors identified a Green self-revealing NCV of Technical Specification (TS) 5.4, Procedures, because Entergy staff did not adequately preplan the implementation of a plant modification to install a digital reactor water recirculation (RWR) flow control system during the 2012 refueling outage. Specifically, post-maintenance testing (PMT) failed to identify that a portion of the runback logic was incorrectly programmed. As a result, the RWR system was restored to operation without identifying the error. On November 8, 2012, during power ascension activities following a subsequent forced outage, the A RWR pump demand signal increased from minimum flow (approximately 30 percent) to approximately 44 percent with no operator action when feedwater flow increased above 20 percent. This resulted in an unexpected power increase of approximately 1.4 percent (37 megawatts thermal (MWth)). As immediate corrective action, control room operators reduced flow in the A RWR loop to restore it to pre-transient conditions, locked the scoop tubes for both RWR motor-generators, and placed the power ascension on hold pending further evaluation of the event. The issue was entered into the corrective action program (CAP) as condition report (CR)-JAF-2012-08042. The issue of inadequate PMT was subsequently entered into the CAP as CR-JAF-2013-05326. The finding was more than minor because it was similar to Example 4.b in IMC 0612, Appendix E, Examples of Minor Issues, in that it resulted in a plant transient. In addition, the finding adversely affected the Initiating Events cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 1 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2012, the inspectors determined this finding was of very low significance (Green) because the performance deficiency did not cause a reactor trip or the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The finding had a cross-cutting aspect in the area of Human Performance, Resources, because FitzPatrick did not ensure that the PMT acceptance criteria specified in the engineering change (EC) package were clearly translated into PMT testing work packages to verify successful implementation of the digital RWR flow control modification. |
| 05000333/FIN-2013007-02 | 2013Q3 | Green | NRC identified | Failure to verify adequacy of the FOTP NPSH | The team identified a finding of very low safety significance (Green) involving a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, because Entergy had not verified the adequacy of the existing design analyses for the available net positive suction head (NPSH) to the EDG fuel oil transfer pumps. Specifically, the team identified several non-conservative design assumptions indicating that Entergy did not adequately account for NPSH in their calculation for the 7-day onsite supply of fuel oil to the EDGs. Entergy performed an operability evaluation, implemented appropriate compensatory measures, and entered the issue into their corrective action program to evaluate and resolve the design deficiency. The performance deficiency was determined to be more than minor because it was similar to Example 3.j of NRC IMC 0612, Appendix E, and was associated with the Design Control 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. The team evaluated the finding in accordance with IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, Exhibit 2 Mitigating Systems screening questions. The finding was determined to be of very low safety significance because it was a design deficiency confirmed not to result in a loss of operability. This finding was not assigned a cross-cutting aspect because it was a historical design issue not indicative of current performance. Specifically, the performance deficiency had occurred outside of the nominal three year period for evaluating present performance as defined in IMC 0612. | |
| 05000333/FIN-2013007-01 | 2013Q3 | Green | H.2 | NRC identified | Failure to correctly position EDG room ventilation temperature controllers in automatic | The team identified a finding of very low safety significance (Green) involving a non-cited violation of Technical Specification (TS) 5.4, Procedures. Specifically, following EDG maintenance, operators did not restore the A and C EDG ventilation systems in accordance with operating procedure OP-60, Diesel Generator Room Ventilation. In particular, operators failed to correctly position the A and C EDG room ventilation temperature controllers to automatic as required by Entergy procedure OP-60. Following discovery, operators promptly restored controllers to automatic, performed additional extentof- condition control panel walkdowns throughout the plant, and entered the issue into their corrective action program to evaluate and address causal factors. The performance deficiency was determined to be more than minor because it was associated with the Configuration Control 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. The team evaluated the finding in accordance with IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, Exhibit 2 Mitigating Systems Screening Questions. The team determined that the finding was of very low safety significance because it was not a design qualification deficiency resulting in a loss of functionality or operability and did not represent an actual loss of safety function of a system or train of equipment. The team determined that this finding has a cross-cutting aspect in the area of Human Performance, Work Practices Component, because Entergy did not adequately ensure supervisory and management oversight of EDG ventilation system restoration activities such that nuclear safety was supported. |
| 05000333/FIN-2013003-01 | 2013Q2 | Green | H.10 | NRC identified | TS Actions for Inoperable Control Rod Not Performed Within the TS Allowed Completion Time | The inspectors identified an NCV of technical specification (TS) 3.1.3, Control Rod Operability, because Entergy operators did not take the required actions within the allowed completion time in response to indication that the scram capability of a control rod was indeterminate. Specifically, when available information concerning the scram solenoid pilot valves (SSPVs) required control rod 30-11 to be declared inoperable, operators did not declare the control rod inoperable, did not fully insert the control rod within three hours, and did not disarm the associated control rod drive within four hours, as required by TS 3.1.3.C. Entergys corrective actions included fully inserting and electrically disarming control rod 30-11, replacing the SSPVs, revising the instructions to operators, briefing operators on this issue, and initiating a condition report. This finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, operators did not fully insert and electrically disarm control rod 30-11 within the TS allowed completion time when the scram capability of the control rod was indeterminate, and therefore required to be declared inoperable. In accordance with IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the finding was of very low safety significance (Green) because it did not affect multiple automatic reactor shutdown functions, did not involve an unintentional positive reactivity addition, and did not result in inability to control changes in reactivity during crew operations. The finding had a cross-cutting aspect in the area of Human Performance, Decision Making, because, given industry operating experience (OE) that cessation of the SSPV buzzing sound was a possible indication of a condition that would prevent the SSPV from performing its safety function, Entergy staff did not communicate to on-shift Operations Department personnel the need to promptly declare control rod 30-11 inoperable if this condition were to occur. |
| 05000333/FIN-2013003-02 | 2013Q2 | Green | P.3 | Self-revealing | Inadequate Corrective Action for DHR System Degradation Results in Loss of DHR During R20 | A self-revealing finding (FIN) was identified for a loss of decay heat removal (DHR) during refueling outage 20 (R20) that was the result of inadequately remediated DHR system degradation. Specifically, prior to using the system during R20, Entergy did not clean scale buildup in the DHR secondary cooling loop heat exchangers (HXs) causing low secondary system pressure, and Entergy did not address the resultant reduction in margin to the primary cooling loop pump automatic shutdown on low primary-to-secondary differential pressure. As a result, a spurious automatic DHR system shutdown occurred while it was functioning as the alternate method of DHR in place of residual heat removal (RHR) shutdown cooling. Entergys corrective actions included restarting DHR and initiating condition report CR-JAF-2012-06934. Entergy also initiated actions to evaluate corrective measures such as modifying the differential pressure trip, adding secondary loop water chemistry treatment, and cleaning of the HXs. This finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, there was an unplanned shutdown of the DHR system for about 50 minutes when it was providing the shutdown cooling function. The inspectors determined the significance of the finding using IMC 0609, Appendix G, Shutdown Operations Significance Determination Process. Per Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Operational Checklists for both PWRs (pressurized water reactors) and BWRs (boiling water reactors), Checklist 7, BWR Refueling Operation with RCS Level > 23, this finding impacted checklist item I.C, because at the time of the event the DHR system was functioning as the alternate method of DHR in place of RHR shutdown cooling. The finding was determined to be of very low safety significance (Green) because the finding did not require a quantitative assessment as described in checklist 7 of Attachment 1 to Appendix G, because checklist item I.C. is not listed as requiring phase 2 or 3 analysis, and the finding did not constitute a loss of control event per Appendix G, Table 1. The inspectors determined that the finding had a cross-cutting aspect in the Problem Identification and Resolution area, Corrective Action Program component, because Entergy staff did not take appropriate corrective actions to address the adverse trend in DHR system performance. |
| 05000333/FIN-2013002-03 | 2013Q1 | Severity level IV | NRC identified | Failure to Obtain NRC Staff Review and Approval Prior to Changing the TS Definition of a Core Quadrant | The inspectors identified a Severity Level (SL) IV non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) 50.59, Changes, Tests, and Experiments, because Entergy personnel implemented a change to the technical specification (TS) definition of core quadrant without prior review and approval by the NRC staff in accordance with 10 CFR 50.59(c)(1)(i). Specifically, Entergy staff changed the definition of core quadrant in Revision 5 of Reactor Analyst procedure RAP-7.1.04C, Neutron Instrumentation Monitoring During In-Core Fuel Handling, which allowed operators to interpret what constitute core quadrant boundaries, such that core alterations could be performed anywhere in the core provided any three source range (neutron) monitors (SRMs) were operable. As immediate corrective action to the task interface agreement (TIA) final response, FitzPatrick staff withdrew RAP-7.1.04C pending revision of the core quadrant definition. The inspectors verified that TS 3.3.1.2.2 had been satisfied during all core alterations that were performed during the 2010 and 2012 refueling outages, using the standard definition of a core quadrant. Entergy staff entered this issue into the corrective action program (CAP) as condition report (CR)-HQN-2013-00034. The inspectors determined that Entergy staffs implementation of a redefinition of core quadrant prior to its review and approval by the NRC staff as specified in 10 CFR 50.59(c)(1)(i) was a performance deficiency that was reasonably within Entergy staffs ability to foresee and correct. Because this was a violation of 10 CFR 50.59, it was considered to be a violation that potentially impedes or impacts the regulatory process. Therefore, this violation was characterized using the traditional enforcement process. The violation was determined to be more than minor in accordance with the NRC Enforcement Manual, Section 7.3.E.6, because there was a reasonable likelihood that the change to the definition of what constituted a core quadrant boundary would require Commission review and approval prior to implementation. Additionally, the inspectors noted that, in accordance with Inspection Manual Chapter (IMC) 0612, Appendix B, Issue Screening, the underlying performance deficiency would screen as more than minor because, if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, potentially inadequate SRM coverage during refueling operations could affect the TS bases function to provide early indication of unexpected subcritical multiplication that could be indicative of an approach to criticality. NRC Enforcement Manual Section 7.3 provides guidance to assess 10 CFR 50.59 violations through the significance determination process (SDP). In this case, the inspectors determined the violation could be evaluated using the SDP in accordance with IMC 0609 Appendix G, Shutdown Operations Significance Determination Process, Checklist 7, BWR Refueling Operation with RCS Level Greater Than 23 Feet. The finding affected the Reactivity Guidelines attribute that assumes existing core alteration TS are being met. Since this attribute does not require quantitative assessment, the finding was screened as Green in accordance with Section 3.3, Mitigation Capability. In accordance with the NRC Enforcement Policy, Section 6.1.d.2, this violation was categorized as SL IV because the issue was evaluated by the SDP as having very low safety significance (Green). The finding did not have a cross-cutting aspect because the performance deficiency did not occur within the past three years and therefore was not reflective of present performance. |