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| Q | Significance | CCA | Identified by | Title | Description | |
|---|---|---|---|---|---|---|
| 05000271/FIN-2000003-01 | 2000Q2 | NRC identified | N/A | The team identified that the augmented off-gas building ventilation system failed a surveillance in May 1999. Subsequently, the licensee identified that the shutdown iodine filter for the mechanical vacuum pump for the main condenser failed a surveillance in March 1998. In both cases, a work request was initiated to repair the system; but no ER was written, as required by the ER procedure. The team identified a third example where a work request was initiated to resolve a discrepancy related to an alarm setpoint, but the request was canceled without resolving the problem. Nonetheless, the failure to initiate ERs for the first two issues is a violation of the VY Technical Specifications related to procedure implementation, and is being treated as a Non-Cited Violation. The violation was not assessed using the Significance Determination Process, as it did not impact one of the cornerstones; however, it provides substantive information relative to the cross cutting issue of problem identification and resolution. (Section 4OA2.1) | ||
| 05000271/FIN-2004006-02 | 2004Q4 | Severity level IV | NRC identified | Entergy DID Not Notify the NRC of a Licensed Senior Operator'S Medical Condition | The inspectors identified a Severity Level IV NCV of 10 CFR 50.74(c) because Entergy did not notify the NRC within 30 days of the identification of a medical condition that caused a licensed senior operator to fail to meet the requirements of 10 CFR 55.21. That medical condition ultimately required the NRC to issue a conditional (restricted) license. Specifically, Entergy became aware of a medical condition in March 2004 that caused a licensed senior operator to fail to meet the requirements of 10 CFR 55.21 and for which a conditional (restricted) license was required. However, Entergy did not notify the NRC of the medical condition until five months later, in August 2004 Entergy's failure to report the medical condition to the NRC impacted the regulatory process, in that, between April and August 2004, the NRC was unaware of a medical condition that warranted issuance of a conditional (restricted) license. Because the finding impacted the regulatory process, it was dispositioned using the traditional enforcement process instead of the significance determination process. This issue has been entered into Entergy's corrective actions program. | |
| 05000271/FIN-2004007-01 | 2004Q3 | Severity level III | NRC identified | DID Not Keep Adequate Records, Follow Procedures, and Perform Inventory of Special Nuclear Material | The inspectors identified an apparent violation of 10 CFR 74.19 because Entergy and its predecessor did not keep adequate special nuclear material inventory records of two spent fuel rod pieces, did not follow its written procedures when two spent fuel rod pieces were moved to a fuel storage liner, and did not conduct adequate periodic physical inventories of the two spent fuel rod pieces. Because the two spent fuel rod pieces remained in the Vermont Yankee spent fuel pool, the entire time the apparent violation existed, there was no actual safety consequence of this apparent violation. Nevertheless, the NRC considers this apparent violation a potentially significant failure of Entergys material and control accounting program. This failure could have resulted in these two spent fuel rod pieces being inappropriately included in a shipment of radioactive material to a low-level radioactive waste site. | |
| 05000271/FIN-2004009-01 | 2004Q4 | White | NRC identified | Failure to establish a means to provide early notification and clear instruction to a portion of the populace within the plume exposure pathway emergency planning zone (EPZ), as required by the Vermont Yankee Emergency Plan. | The inspector identified an apparent violation associated with emergency planning standard 10 CFR 50.47(b)(5) which has a low to moderate safety significance because the method of distributing tone alert radios to members of the public outside of siren coverage was not meeting the intent of the design basis for the alert and notification system. The finding is greater than minor because this impacts the EP cornerstone attribute of facilities and equipment and it affects the cornerstone objective of ensuring that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The status of this finding will be finalized pending review of any additional information submitted to the NRC in response to this report. | |
| 05000271/FIN-2007010-01 | 2006Q3 | Severity level IV | NRC identified | Improper Entry into a Locked High Radiation Area (Section 4OA5) | The Licensees technician did not perform a survey to establish radiological conditions in the LHRA prior to allowing access to an AO. Specifically, on August 17, 2006, the RPT did not perform a survey of the reactor water clean up room following the completion of a resin transfer to establish the current radiological conditions, and to ensure occupational dose limits would not be exceeded. The RPT\'s actions resulted in a violation of 10 CFR 20.1501, which requires that each licensee make, or cause to be made, surveys that may be necessary for the licensee to comply with the regulations in Part 20, including Part 20.1201, occupational dose limits. The NRC further determined that the technician\'s actions were willful, in careless disregard for the requirements. Specifically, despite the fact that the RPT was hampered in his assessment of the job task by not being originally assigned to the job, based on his training, 31 years of experience, and procedural knowledge, there was sufficient evidence to indicate he willfully violated the survey requirements and caused Vermont Yankee to be in violation of NRC regulations. Because you are responsible for the actions of your employees, including contract employees, and because the violation was willful, the violation was evaluated under the NRC traditional enforcement process as set forth in Section IV.A.4 of the NRC Enforcement Policy. The NRC concluded that the violation, absent willfulness, would be considered minor, because the exposure from this incident did not result in the individual\'s occupational dose limits being exceeded. However, the NRC increased the severity level to Severity Level IV because the technician\'s actions were willful. The NRC considered issuance of a Notice of Violation for this issue. However, after considering the factors set forth in Section VI.A.1 of the NRC Enforcement Policy, the NRC determined that a non-cited violation (NCV) is appropriate in this case because: (1) you initially identified the violation and promptly informed the NRC of the occurrence; (2) the violation involved the acts of an individual who was not a supervisor in your organization; (3) the violation appeared to be an isolated action of the employee without management involvement and was not caused by a lack of management oversight; and (4) you took significant remedial action commensurate with the significance of the event such that it demonstrated the seriousness of the violation to other employees and contractors. Although the technician received no supervision leading up to and during his task, the NRC concluded that the violation was not attributable to a lack of management oversight, because it was reasonable to expect that an RPT with 31 years of experience would not need significant oversight to perform this task. | |
| 05000271/FIN-2008002-01 | 2008Q1 | Green | H.7 | Self-revealing | Inadequate Work Order Results in Unplanned \"A\" Service Water Pump Inoperability | A self-revealing non-cited violation (NCV) of Technical Specification 6.4, Procedures, was identified for Entergys failure to provide an adequate procedure for setting the A service water (SW) pump lower motor guide bearing during its replacement in May 2007. Specifically, work order (WO) 111249-14, Replace A Service Water Pump Lower Motor Guide Bearing, did not provide adequate guidance to ensure proper verification of shaft lift to prevent loading the lower motor guide bearing. As a result, the bearing was improperly set which caused bearing degradation and unplanned A SW pump inoperability on February 12, 2008. Corrective actions taken or planned include replacement of the A SW pump motor and revisions to applicable procedures. The finding is more than minor because it is associated with the equipment performance attribute of the Initiating Events Cornerstone and affects the associated Cornerstone objectives of limiting the likelihood of those events that upset plant stability. The finding is of very low safety significance because the estimated increase in core damage frequency was less than 1E-06, assuming the reactor was operating at full power and the A SW pump was unavailable for less than 3 days. The performance deficiency has a cross-cutting aspect in the area of Human Performance, Resources component, in that Entergy did not provide an adequate procedure for the installation of the lower motor guide bearing. (H.2(c)) (Section 1R12 |
| 05000271/FIN-2008003-01 | 2008Q2 | Green | P.3 | NRC identified | Ineffective Reactor Building Crane Maintenance | The NRC identified a non-cited violation for Entergys failure to take timely corrective action, as required by 10CFR50.65(a)(1), after the reactor building crane (RBC) exceeded reliability performance goals. Specifically, from April 12, 2007, when the RBC was classified as (a)(1), until May 12, 2008, when the RBC brakes failed to function during the movement of a spent fuel storage cask, Entergy failed to take corrective actions in response to the RBC not meeting established goals. This issue was entered into the licensee\'s corrective action program as Condition Report CR 2008-2043. The issue is greater than minor because the failure to implement timely corrective actions resulted in the failure of the RBC brakes during the movement of a spent fuel storage cask. The finding is not suitable for evaluation under the Significance Determination Process, but has been reviewed by NRC management and was determined to be a finding of very low safety significance (Green) because the spent fuel storage cask was in an approved load path, and the refuel floor allowed the brakes to engage when sufficient load was removed from the hoist. This finding has a crosscutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, in that Entergy failed to take appropriate corrective actions to address safety issues and adverse trends in a timely manner, commensurate with their safety significance and complexity. (P.1(d) |
| 05000271/FIN-2008004-01 | 2008Q3 | Green | Licensee-identified | Licensee-Identified Violation | The following violation of very low safety significance was identified by the licensee and meets the criteria of Section VI of the NRC Enforcement Policy, NUREG-1600, for being dispositioned as a Non-Cited Violation. Two examples were identified in which the licensee failed to report potentially disqualifying licensed operator medical conditions to the NRC as required by 10CFR55.25 and 10CFR50.74. Contrary to these requirements, in the first instance, notification of a potentially disqualifying condition related to prescription medication was delayed for 15 months. The notification of a second operators potentially disqualifying physical condition was delayed for six months. The issues have been entered in the licensees corrective action program as site condition reports CR-2008-02901 and CR-2008-3429. Corrective actions included obtaining a peer review of medical records from another site, and a corporate level condition report CR-HQN-2008-00724 to review medical records at all Entergy sites for two years. The issues were of very low safety significance because the first individual did not stand watch since his diagnosis. The second individual made no operator errors to indicate potential impairment from failure to take prescribed medication, and the ultimate license restriction added was simply the requirement to take the medication as prescribed (EA- 2008-277) | |
| 05000271/FIN-2008006-01 | 2008Q3 | Green | Self-revealing | Inadequate Preventative Maintenance Program for Reactor Building Crane | A self-revealing Finding of very low safety significance was identified forEntergy not fully developing an adequate preventive maintenance (PM) program for thereactor building crane (RBC). As a result, on May 12, 2008, when the first loaded spentfuel storage cask was removed from the spent fuel pool (SFP) and was being lowered toa height of four inches above the refueling floor, the crane brakes did not engage andthe spent fuel storage cask continued to be slowly lowered to the refueling floor. Thisissue was entered into the licensees corrective action program as condition reportCR-VTY-2008-02043.This issue is greater than minor because the finding resulted in the failure of the RBCbrakes to engage during the lowering of a loaded spent fuel storage cask. The findingwas determined to be of very low safety significance (Green) because the spent fuelstorage cask remained under control of the reactor building crane, was in an approvedload path, and the emergency braking system was available | |
| 05000271/FIN-2008008-01 | 2008Q3 | Green | P.1 P.1(a) | NRC identified | Inadequate Testing of Safety Related Batteries (Section 1R21.2.1.1) | The team identified a finding of very low safety significance involving a non-cited violation of 10 CFR 50, Appendix B, Criterion XI, Test Control, in that, Entergy did not properly document and evaluate safety related battery test results. Specifically, the NRC identified three instances involving the rotating uninterruptible power supply system and the alternate shutdown batteries where Entergy did not adequately evaluate test results to calculate battery capacity. In response, Entergy entered these issues into the corrective action program and demonstrated that there was sufficient margin to assure operability of the safety related batteries. The finding is more than minor because it is associated with the human performance attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The team determined the finding was of very low safety significance (Green) because it was not a design or qualification deficiency, did not represent a loss of system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program Component, because Entergy did not identify issues in a timely manner commensurate with their safety significance. (IMC 0305, Aspect P.1(a)) (1R21.2.1.1 |
| 05000271/FIN-2008008-02 | 2008Q3 | Green | NRC identified | Inadequate Design Control for Emergency Diesel Generator Load Testing Criteria (Section 1R21.2.1.19) | The team identified a finding of very low safety significance involving a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, in that, Entergy did not ensure that the design basis, as defined in calculations and the Updated Final Safety Analysis Report for manual emergency diesel generator (EDG) loading, was verified by a suitable testing program. Specifically, Entergy had not performed a suitable test to demonstrate that the 1B EDG was capable of loading to a value that demonstrated its calculated maximum load during a postulated accident scenario, as allowed in operating procedures. Entergy entered the issue into their corrective action program and completed an operability assessment, which demonstrated that the emergency diesel generators were capable of performing their design function. The finding is more than minor because it is associated with the design control attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team determined the finding was of very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in a loss of standby onsite power operability or functionality. (1R21.2.1.19 | |
| 05000271/FIN-2008008-03 | 2008Q3 | Green | NRC identified | Inadequate Procedure for Station Blackout Load Shedding (Section 1R21.2.2.1) | The team identified a finding of very low safety significance (Green) involving a non-cited violation of 10 CFR 50.63, Loss of all Alternating Current Power, in that, Entergy did not ensure that adequate battery capacity would be available during a station blackout (SBO), as assumed in the stations SBO analysis. Specifically, unrecognized delays in performing a credited manual direct current (DC) load shedding operator action, as well as an incorrectly translated minimum battery voltage referenced in the stations SBO procedure, could have resulted in the B station battery capacity being insufficient during an SBO. Entergy entered the issue into the corrective action program. Entergy also recalculated the B station battery capacity and determined that sufficient battery capacity existed when realistic load shedding assumptions were applied (battery remained operable). The finding is more than minor because it is associated with the procedure quality attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This finding was determined to be of very low safety significance (Green) because it was not a design or qualification deficiency, did not represent a loss of system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. (1R21.2.2.1 | |
| 05000271/FIN-2009002-01 | 2009Q1 | Green | P.1 P.1(a) | NRC identified | Failure to write a condition report (CR) for an adverse condition associated with water accumulating in the turbine building supply fan housing plenum area | The inspectors identified a Green non-cited violation (NCV) of Vermont Yankee Technical Specifications Section 6.4, Procedures, for Vermont Yankees failure to take action to correct a specific and foreseen malfunction of a plant component. Specifically, Vermont Yankee failed to initiate a condition report (CR) for an adverse condition associated with water accumulating in the turbine building supply fan housing plenum area, which led to the inoperability of the A emergency diesel generator (EDG) on January 21, 2009 for four hours. Vermont Yankee operations and maintenance personnel stopped the source of the water accumulation and restored the A EDG to operable status. This NCV has since been entered into the Vermont Yankee corrective action program (CAP). The finding was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone; and, it affected the cornerstone objective of ensuring the reliability, availability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the A EDG was rendered inoperable for approximately four hours, but less than the sevenday Technical Specifications (TS) 3.10 allowed outage time. The finding had a crosscutting aspect in the Corrective Action Program component of the Problem Identification and Resolution (PI&R) cross-cutting area because Vermont Yankee did not identify within the CAP the rising water level in the turbine building supply fan housing plenum area in a timely manner commensurate with its safety significance (P.1(a)). (Section 1R12) |
| 05000271/FIN-2009002-02 | 2009Q1 | Green | H.2 | NRC identified | Failure to perform procedurally required engineering evaluations for scaffolding. | The inspectors identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for Vermont Yankees failure to routinely perform procedurally required engineering evaluations for scaffold bracing attached to pipe supports. Specifically, Vermont Yankee failed to perform engineering evaluations on 27 out of 32 scaffolds with horizontal bracing attached to safety related pipe supports. Subsequently, each scaffold was evaluated and documented by Vermont Yankee engineering and no immediate safety issues were found. This NCV has been entered into the Vermont Yankee corrective action program (CAP). The performance deficiency was more than minor because if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern. Specifically, installing scaffold bracing on pipe supports without engineering approval could place a pipe support in an unanalyzed seismic condition, which could lead to failure in a seismic event. The finding had a cross-cutting aspect in the Work Practices component of the Human Performance cross-cutting area because Vermont Yankee did not implement adequate management oversight of contractor work activities regarding scaffold procedural compliance. (H.4(c)). (Section 4OA2) |
| 05000271/FIN-2009002-03 | 2009Q1 | Green | Licensee-identified | Licensee-Identified Violation | The following violation of very low safety significance (Green) was identified by Vermont Yankee and is a violation of NRC requirements which meets the criteria of the NRC Enforcement Policy, for being dispositioned as a Non-Cited Violation. Vermont Yankee Technical Specification 6.5.B requires, in part, that areas in which the intensity of radiation is greater than 1000 mrem/hour at 30 cm, but less than 500 rad/hour at one meter shall follow the requirements of a high radiation area as well as provide for a locked or continuously guarded entryway to prevent unauthorized entry. Contrary to the above, on October 22, 2008, a Quality Assurance (QA) person identified that a posted LHRA over the Control Rod Drive access tube into the Drywell (Reactor Building 252 elevation) was not being continuously guarded by an access control guard in lieu of a locked barricade. There was a posted access control guard in place at the time of discovery but the individual was not aware of his responsibility relative to high radiation area control for that specific potential access point. Technical Specification 6.5.A in part requires that areas in which the intensity of radiation is greater than 100 mrem/hour at 30 cm, but less than 1000 mrem/hour at one meter be barricaded and conspicuously posted and access controlled by requiring issuance of a RWP. Contrary to the above, on October 23, 2008, Vermont Yankee staff discovered a conveyer ramp used for the transport of lead shielding that was not conspicuously posted or barricaded. The ramp extended over the existing high radiation area boundary established at the Drywell equipment hatch, Reactor Building 252 elevation. Radiation Protection established a guard for the area until the ramp could be removed. Additionally, on November 23, 2008, Vermont Yankee identified that an area previously posted and controlled as a high radiation area was found un-posted during a routine survey in the RHR room, 213 elevation southeast corner room. The area was immediately secured until posting and a barricade could be established at the bottom of the RHR heat exchanger. These issues were identified in Vermont Yankees CAP as CR-VTY-2008-04294, CRVTY- 2008-04377, and CR-VTY-2008-05236. These issues constitute a finding of very low safety significance (Green) because no persons had unauthorized access to these areas while these conditions existed and therefore no unplanned exposures occurred | |
| 05000271/FIN-2009004-01 | 2009Q3 | Green | H.8 | NRC identified | Failure to initiate corrective action condition reports for all deficient items identified during cooling tower inspections | The inspectors identified a Green NCVof 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and DraWings, in that Entergy did not initiate corrective action condition reports (CRs)\'for all deficient items identified during Cooling Tower (CT) inspections. Entergy entered this issue into their corrective action program (CAP) and performed an operability assessment which determined that the safety related function of the CTs was always available. The inspectors determined that the finding was more than minor because if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, deficiencies might not be tracked to resolution, management attention or other independent reviews would not be appropriately applied, and the need for operability determinations may be missed. The finding was determined to be of very low safety significance (Green) because the finding did not involve a design or qualification deficiency resulting in loss of operability or functionality, did not result in a loss of system safety function, and did not screen as potentially risk significant due to external initiating events. This finding had a cross-cutting aspect in the Work Practices component of the Human Performance cross-cutting area because Entergy did not follow procedures and initiate CRs to identify cooling tower deficiencies as required by operating procedure (OP) 52114. IH.4(b) |
| 05000271/FIN-2009005-01 | 2009Q4 | Green | H.5 | NRC identified | Inadequate Risk Assessment Associated with the Low Pressure Coolant Injection Subsystem | The inspectors identified a non-cited violation (NCV) of 10 CFR 50.65 paragraph (a)(4), Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, because Entergy did not assess and manage the increase in risk that resulted from maintenance activities that impacted the availability of the low pressure coolant injection subsystem (LPCI). On December 4, 2009, Entergy conducted a test of the high pressure coolant injection (HPCI) system as a retest following maintenance activities. Operations placed both trains of the residual heat removal (RHR) system in the torus cooling mode. This alignment impacted the ability of the LPCI subsystem to automatically perform its function in some design basis accident scenarios. However, the inspectors noted that the LPCI subsystem was not included as part of the risk assessment, and that subsystem was not maintained as available in accordance with Entergy procedures. Entergy entered this issue into the corrective action program (CAP), and initiated a preliminary investigation to review the effectiveness of Maintenance Rule accounting for LPCI unavailability while in the torus cooling mode. The finding is more than minor because Entergy\'s risk assessment did not consider risk significant structures, systems, and components (SSCs) (i.e. LPCI subsystem) that were unavailable during the maintenance activity. The finding is associated with the Configuration Control attribute of the Mitigating Systems cornerstone, and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined that the finding is of very low safety significance because the incremental core damage probability deficit was less than 1.0E-6. This finding has a cross-cutting aspect in the Human Performance cross-cutting area,Work Control component, because Entergy did not appropriately plan and incorporate risk insights. in work activities that impacted the availability of the LPCI subsystem. (H.3(a)) |
| 05000271/FIN-2009005-02 | 2009Q4 | NRC identified | Troubleshooting Activities on Inoperable vacuum Breakers | The inspectors identified an Unresolved Item (URI) associated with troubleshooting activities on inoperable vacuum breakers. On May 14, 2009, and August 14, 2009, Entergy declared both the V1619- 5E and V16-19-5F torus-to-drywell vacuum breakers inoperable, respectively. The vacuum breakers were declared inoperable when it was identified that their breakaway force exceeded the maximum allowable TS value of 0.5 psid. Entergy entered TS 3.7.A.6.b, which stated that up to two out of ten torus-to-drywell vacuum breakers may be determined to be inoperable provided that they are secured, or known to be, in the closed position. On September 29,2009, and October 8,2009, Entergy conducted troubleshooting activities on both inoperable vacuum breakers. The troubleshooting activities involved opening and closing the inoperable vacuum breakers to obtain breakaway force data and to possibly repair the vacuum breakers. The inspectors noted TS 3.7.A.6 did not have a condition that allowed the opening of an inoperable vacuum breaker once it is secured in the closed position in accordance with the requirements of TS 3.7.A.6.b. Furthermore, TS 3.7.A.8 stated that if TS 3.7.A.6 cannot be met, an orderly shutdown shall be initiated immediately, and the reactor shall be in a cold shutdown within 24 hours. The inspectors noted that once the inoperable vacuum breakers are secured in the closed position, TS 3.7.A.6 can be met, and that opening the inoperable vacuum breakers is a potential violation of TS 3.7.A.6.b. Vacuum breakers can be opened for surveillance testing, however, there are no TS requirements to conduct surveillances on inoperable equipment. This issue remains unresolved pending a review by the NRC Office of Nuclear Reactor Regulation to determine if this issue constitutes a violation of Entergy Vermont Yankee TS. (URI 05000271/2009005-02, Troubleshooting Activities on Inoperable Vacuum Breakers | ||
| 05000271/FIN-2009006-01 | 2009Q2 | Green | P.3 | NRC identified | Failure to take adequate corrective actions for a HPCI system functional failure | The team identified a Green non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for Entergys failure to take adequate corrective actions for a condition adverse to quality involving an issue that had the potential to negatively impact the high pressure coolant injection (HPCI) system. Specifically, Entergy failed to take timely and appropriate corrective actions commensurate with the safety significance (potential repeat functional failure of the HPCI system due to degraded direct current (DC) contactors) of the issue. Entergys short-term corrective actions included a visual inspection of several affected DC breaker cubicles, a HPCI system operability evaluation, and interim guidance to plant operators. Entergy entered the condition into their CAP (CR 2009-1489) and performed a root cause evaluation. The finding is more than minor because it affected the equipment performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the reliability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the adverse condition represented a challenge to the reliability of the HPCI system due to the systems vulnerability to a repeat functional failure. The finding was determined to be of very low safety significance (Green) because it: was not a design or qualification deficiency confirmed not to result in loss of operability; did not represent a loss of system safety function; did not represent actual loss of safety function of a single train for greater than its technical specification allowed outage time; did not represent an actual loss of safety function of one or more non-technical specification trains for equipment designated as risk-significant per 10 CFR 50.65 for greater than 24 hours; and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding has a crosscutting aspect in the area of Problem Identification and Resolution, Corrective Action Program Component, because Entergy failed to take appropriate corrective actions to address a safety issue in a timely manner, commensurate with the safety significance and complexity (P.1.d). Specifically, Entergy did not take appropriate corrective actions to adequately address the extent of condition for a HPCI functional failure in June 2007 due to degraded DC contactors prior to April 2009. (Section 4OA2.1.c |
| 05000271/FIN-2009405-01 | 2009Q4 | Green | H.5 | NRC identified | Security | |
| 05000271/FIN-2010002-01 | 2010Q1 | Green | P.1 P.1(a) | NRC identified | High Pressure Coolant Injection Inoperable Due to Spurious Suction Valve Swap and Technical Specification Actions Not Performed | The inspectors identified an NCV of very low safety significance (Green) of technical specification 3.5.E, High Pressure Coolant Injection (HPCI) System, because Entergy staff failed to identify that HPCI was inoperable, enter the required limiting condition for operation, and immediately verify that the reactor core isolation cooling (RCIC) system was operable. Entergy initiated CR-VTY-2010-01420 and CR-VTY-2010-01506 to address the issues, issued standing orders to ensure HPCI and RCIC are considered inoperable when not aligned to the condensate storage and transfer system (CST), and initiated corrective actions to ensure design basis analysis associated with power uprate is properly incorporated into various documents, including technical specifications (TS) and the updated final safety analysis report (UFSAR). This finding is more than minor because it affected the equipment performance attribute of the Mitigating Systems 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, the availability of the CST to provide water for core cooling to HPCI during transient and emergency situations was affected. The inspectors determined the significance of the finding using IMC 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations. The finding was determined to be of very low safety significance (Green) because the exposure time associated with the HPCI suction valves being not properly aligned to the CST was 45 minutes, i.e. less than three days. The inspectors determined this finding had a cross-cutting aspect in the area of problem identification and resolution within the corrective action program (CAP) component because Entergy personnel did not completely and accurately identify the issues associated with HPCI being aligned to the torus instead of to the CST. (P.1(a) |
| 05000271/FIN-2010002-02 | 2010Q1 | Green | H.5 | NRC identified | Emergency Diesel Generator Surveillance Testing Not Risk Assessed in Accordance with 10 CFR 50.65 (a)(4) | The inspectors identified an NCV of very low safety significance (Green) of 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, because Entergy staff did not assess and manage the increase in risk due to surveillance testing activities that impacted the availability of the \'A\' emergency diesel generator (EDG) in accordance with 10 CFR 50.65 (a)(4). Entergy initiated CR-VTY-201 0-01 019 to address the issue, issued a standing order to ensure the EDGs are properly considered unavailable during future surveillance tests, and commenced an extent of condition review to determine the staff\'s effectiveness at properly accounting for unavailability in accordance with 10 CFR 50.65 (a)(4) for the EDGs and other risk significant systems. This finding is more than minor because it affected the equipment performance attribute of the Mitigating Systems 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, the availability of the \'A\' EDG was affected and Entergy\'s risk assessment did not consider risk significant structures, systems and components (SSCs) (i.e., EDGs) that were unavailable during the maintenance activity and did not take risk management actions. The inspectors determined the significance of the finding using IMC 0609 Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process. The finding was determined to be of very low safety significance (Green) because the incremental core damage probability deficit for the time the \'A\' EDG \\ was unavailable was less than 1.0E-6. The inspectors determined this finding had a cross-cutting aspect in the area of human performance within the work control component because Entergy did not appropriately plan and incorporate risk insights in work activities that impacted the availability of the \'A\' EDG. (H.3(a)) |
| 05000271/FIN-2010002-03 | 2010Q1 | Green | P.2 | NRC identified | Inadequate Design Control for Continuously Submerged Underground Cables | The inspectors identified an NCV of very low safety significance (Green) of 10 CFR Part 50, Appendix B, Criterion III, Design Control, because Entergy did not select and review safety-related cables suitable for application in the environment in which they were found. Specifically, Entergy allowed the continuous submergence of safety-related cables that were not qualified for continuous submergence and failed to demonstrate that the cables would remain operable. Entergy initiated CR-VTY-2009-04142 and CR-VTY-2010-01422 to address the issues, commenced dewatering of the affected manholes, and initiated a preventive maintenance plan to ensure proper conditions. This finding is more than minor because if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern. Specifically, the inspectors noted that the insulation of continuously submerged cables would degrade more than dry or periodically wetted cables which would lead to failures. The inspectors determined the significance of the finding using IMC 0609.04, Phase 1 -Initial Screening and Characterization of Findings. The finding was determined to be of very low safety significance (Green) because it was a design or qualification deficiency which was confirmed to have not resulted in a loss of operability or functionality. Specifically, the continuously submerged cables were not designed or qualified for that environment but were still fully capable of performing their design functions. The inspectors determined this finding had a cross-cutting aspect in the area of problem identification and resolution within the CAP component because Entergy personnel did not thoroughly evaluate the problem when submerged cabling was identified. (P.1 (c)) |
| 05000271/FIN-2010003-01 | 2010Q2 | Green | H.7 | Self-revealing | Inadvertent Loss of RCS Inventory During ECCS Testing Due to Inadequate Procedure | A self-revealing, NCV of very low safety significance (Green) of Technical Specification (TS) 6.4, Procedures, was identified when operators inadvertently drained water from the reactor pressure vessel (RPV) during integrated emergency core cooling system (ECCS) testing. Specifically, Entergy failed to establish the initial plant conditions necessary to perform integrated ECCS testing without causing an inadvertent drain down of the vessel through the main steam lines, the RCIC turbine, and into the torus. Entergy restored the RPV inventory, initiated a CR to perform a root cause evaluation of the issue, and assigned a corrective action to revise the procedure in order to preclude recurrence in future outages. The inspectors determined that the finding was more than minor because it was associated with the Procedure Quality attribute of the Initiating Events Cornerstone, and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the implementation of the inadequate procedure guidance resulted in an unexpected loss of RPV water inventory of approximately 2100 gallons. The inspectors determined that this finding had a cross-cutting aspect in the Human Performance cross-cutting area, Resources component, because the test procedure was inadequate. Specifically, the procedure did not provide adequate directions for establishing plant conditions during a test that had the capability of draining RCS inventory (H.2(c)). |
| 05000271/FIN-2010004-01 | 2010Q3 | Green | H.11 H.12 | Self-revealing | Inadvertent isolation of Reactor Core Isolation Cooling (RCIC) During Surveillance Testing | A self-revealing, Green, non-cited violation (NCV) of Technical Specification 6.4, Procedures, was identified in which technicians incorrectly performed reactor core isolation cooling (RCIC) surveillance test operating procedure (OP) 4365, RCIC Steam Line Low Pressure Functional/Calibration, Rev. 25, resulting in the inadvertent isolation of the RCIC system. Entergy entered this issue into their corrective action program, correctly installed the test equipment, and subsequently performed the test satisfactorily. The inspectors determined that the finding was more than minor because it adversely affected the Human Performance attribute for the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. In accordance with IMC 0609, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, the finding was determined to be of very low risk significance (Green) because the finding was not a design or qualification deficiency, did not represent a loss of system safety function or loss of a single train for greater than its allowed technical specification time, and did not screen as potentially risk significant due to seismic, flooding, or severe weather initiating events. The inspectors determined this finding had a cross-cutting aspect in the Human Performance cross-cutting area, Work Practices component, in that Entergy failed to appropriately self-check and peer-check the digital multimeter (DMM) setup prior to connecting it to the RCIC isolation logic. (H.4(a)) |
| 05000271/FIN-2010005-01 | 2010Q4 | Green | H.13 | NRC identified | Failure to Perform Required Quality Control Inspections | The inspectors identified a non-cited violation (NCV) of 10 CFR 50, Appendix B . Criterion X, lnspection, for the failure to ensure that Quality Control verification inspections were consistently included and correctly specified in quality-affecting procedures and work instructions for construction-like work activities as required by the Quality Assurance Program. Entergy initiated prompt fleet-wide corrective actions to ensure proper work order evaluation and proper inclusion of Quality Control verification inspections. This issue was entered into the corrective action program as condition reports (CR) CR-HQN 2009-01184 and CR-HQN-2O10-0013. The failure to ensure that adequate Quality Control verification inspections were included in quality-affecting procedures and work instructions as required by the Quality Assurance Program was a performance deficiency. This issue was more than minor because, if left uncorrected, it could lead to a more significant safety concern; in that, the failure to check quality attributes could involve an actual impact to plant equipment. This issue affected the Design Control attribute of the Mitigating Systems cornerstone because missed or improper quality control inspections during plant modifications could impact the availability, reliability, and capability of systems needed to respond to initiating events. This performance deficiency was determined to be of very low safety significance (Green), since it was confirmed to involve a qualification deficiency that did not result in a loss of operability or functionality. The inspectors determined that this issue had a cross-cutting aspect in the Human Performance cross-cutting area, Decision-Making component, because the licensee did not have an effective systematic process for obtaining interdisciplinary reviews of proposed work instructions to determine whether Quality Control verification inspections were appropriate H.1(a)1. |
| 05000271/FIN-2010005-02 | 2010Q4 | Green | H.13 | NRC identified | Failure to Implement the Experience and Qualification Requirements of the Quality Assurance Program | The inspectors identified an NCV of 10 CFR 50, Appendix B, Criterion ll, Quality Assurance Program, for the failure to implement the experience and qualification requiiements of the Quality Assurance Program. As a result, the licensee failed to ensure that two individuals assigned to the position of Quality Assurance Manager met the qualification and experience requirements of ANSI/ANS 3.1-1978 as required by the Quality Assurance Program. Specifically, the individual assigned to be the responsible person for the licensee\'s overall implementation of the Quality Assurance Program did not have at least one year of nuclear plant experience in the overall implementation of the Quality Assurance Program within the quality assurance organization prior to assuming those responsibilities. This issue was entered into the corrective action program as CR-HQN-201 0-00386. The failure to ensure that an individual assigned to the position of Quality Assurance Manager met the qualification and experience requirements of ANSI/ANS 3.1-1978 as required by the Quality Assurance Program was a performance deficiency. This issue was more than minor because, if left uncorrected, it could create a more significant safety concern. The failure to have a fully qualified individual providing overall oversight to the Quality Assurance Program had the potential to affect all cornerstones, but the inspectors determined that this finding will be tracked under the Mitigating Systems cornerstone as the area most likely to be impacted. The issue was not suitable for quantitative assessment using existing NRC Significance Determination Process (SDP) guidance, so it was determined to be of very low safety significance (Green) using NRC Inspection Manual Chapter (lMC) 0609, Appendix M, Significance Determination Process Using Qualitative Criteria. The inspectors determined that there was no crosscutting aspect associated with this finding because this issue was not indicative of current performance as it occurred more than three years ago. ( |
| 05000271/FIN-2010005-03 | 2010Q4 | Green | Licensee-identified | Licensee-Identified Violation | Procedure, EN-QV-1 1 1, Training and Certification of InspectionA/erification and Examination Personnel, Section 4.0 (4)(i), requires that the Entergy corporate ANSI Level lll inspector shall perform periodic (annual) surveillances of quality control inspection activities to ensure that the program is being adequately implemented and maintained. Contrary to the above, no surveillances of quality control inspection activities were performed for any Entergy site during calendar year 2008. The issue was not suitable for quantitative significance determination, so it was assessed using IMC 0609, Appendix M, and was evaluated using the qualitative criteria listed in Table 4.1. This finding was determined to be of very low safety significance because other quality assurance program functions remained unaffected by this performance deficiency, so defense-in-depth continued to exist. This issue was entered into the licensee\'s CAP as CR-HQN-2009-001 1 1. | |
| 05000271/FIN-2010008-01 | 2010Q4 | Green | NRC identified | Fire Scenario Resulting in Loss of Reactor Core Isolation Cooling System | The team identified a Green, Non-Cited Violation of the Vermont Yankee Nuclear Power Station Facility Operating License, Condition 3.F, in that Entergy failed to implement and maintain in effect all provisions of the approved Fire Protection Program as described in the Final Safety Analysis Report. Specifically, Entergy failed to assure that reactor vessel water level would remain below the reactor core isolation cooling (RCIC) system steam line for postulated alternate shutdown fire scenarios that spuriously started a reactor feedwater pump (RFP). Entergy initiated condition report CR-VTY-2010-04682 and promptly revised the alternate shutdown procedure to additionally trip all running condensate pumps. The additional action prevented a single spurious operation from restarting or precluding a trip of the RFPs. This finding was more than minor because it was 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 that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the availability of the RCIC system was not ensured for postulated fires in alternate shutdown areas. The team used Phase 1 of IMC 0609, Appendix F, Fire Protection Significance Determination Process, to determine that this finding was of very low safety significance (Green) because the Vermont Yankee Nuclear Power Station alternate shutdown system also includes safety relief valves and a residual heat removal train that can be utilized for reactor pressure and water level control. This finding did not have a cross-cutting aspect because the m9st significant contributor of the performance deficiency was not reflective of current licensee performance. | |
| 05000271/FIN-2010403-01 | 2011Q1 | Green | P.2 | NRC identified | Security | |
| 05000271/FIN-2011002-01 | 2011Q1 | Green | H.8 | Self-revealing | Failure to Follow Foreign Material Exclusion Procedure | A self-revealing, non-cited violation (NCV) of very low safety significance (Green) of Technical Specifications 6.4, Procedures, was identified for inadequate implementation of Entergy procedure EN-MA-118, Foreign Material Exclusion, Revision 6, which resulted in foreign material intrusion into the Residual Heat Removal Service Water (RHRSW) system. Specifically, Entergy did not establish a Foreign Material Exclusion (FME) Zone 1 around the open RHRSW system between completing the closeout inspection and system closure following pump replacement. Entergy\'s immediate corrective actions included conducting a stand down, reinforcing the standards and requirements for FME controls and general procedural compliance, as well as reinforcing expectations for the attention to detail of work practices. Entergy entered the issue into their corrective action program to evaluate for additional corrective measures. The inspectors determined that 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 of systems that respond to initiating events to prevent undesirable consequences, (ie., core damage). Specifically, foreign material made its way into the \'A\' Residual Heat Removal Heat Exchanger (RHR HX) and rendered the \'A\' RHRSW train inoperable for several days. A review of NRC Inspection Manual Chapter (IMC) 0612, Appendix E, Minor Examples, revealed that no minor examples were applicable to this finding. The inspectors used IMC 0609.04, Phase 1 - Initial Screening and Characterization of Findings, and determined that the finding required a Phase 2 review because the \'A\' RHRSW train had an actual loss of safety function for greater than its allowed outage time (7 days). This finding was assessed using IMC 0609 and was determined to be of very low safety significance (Green) based on a Phase 2 analysis. The finding had a cross-cutting aspect in the Human Performance crosscutting area, Work Practices component, because Entergy personnel did not follow EN-MA- 118. Specifically, they did not establish a FME Zone 1 after the system closeout inspection. (H.4(b) |
| 05000271/FIN-2011002-02 | 2011Q1 | Green | H.13 | Self-revealing | Steam Leak on High Pressure Coolant Injection (HPCI) During Surveillance Testing | A self-revealing, Green NCV of Technical Specification 6.4, Procedures, was identified in which maintenance and planning personnel did not involve engineering personnel as required by Entergy procedure EN-MA-1 01, Fundamentals of Maintenance, Revision 9, and EN-WM-105, Planning, Revision 8, resulting in the incorrect material being used to replace the gasket on the flange of High Pressure Coolant Injection System (HPCI) steam trap 23T-3. Entergy ultimately replaced the gasket with the correct material and entered this issue into their corrective action program. The inspectors determined that 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 of systems that respond to initiating events to prevent undesirable consequences. The finding was determined to be of very low safety significance (Green) in accordance with IMC 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations, using Significance Determination Process (SOP) Phases 1, 2 and 3. A Region I Senior Reactor Analyst (SRA) conducted a Phase 3 analysis because the Phase 2 analysis indicated that the finding had the potential to be greater than very low safety significance (Greater than Green). This finding had a cross-cutting aspect in the Human Performance cross-cutting area, Decision Making component. because Vermont Yankee personnel did not obtain interdisciplinary input on the decision to use a different, incorrect gasket material in a steam trap in the HPCI system. (H.1 (a) |
| 05000271/FIN-2011002-03 | 2011Q1 | Green | Licensee-identified | None | Technical Specification 3.5.F, Automatic Depressurization System, allows up to one of. four SRVs in the automatic depressurization system to be inoperable for up to seven days at any time the reactor steam pressure is above 150 psig with irradiated fuel within the vessel, or an orderly shutdown of the reactor shall be initiated and the reactor pressure shall be reduced to less than 150 psig within 24 hours. Contrary to the above, Entergy determined that two (2) of the four (4) SRVs were inoperable for a period of time greater than allowed by Technical Specifications. This determination was based on pneumatic actuator thread seal leakage that was identified during testing of the pneumatic SRV actuators in the 2010 refueling outage. Entergy determined the leakage to be in excess of design requirements. This condition has been entered in the licensee\'s corrective action program (CR-VTY-2010-2187) and corrective actions have been developed. The inspectors determined that this finding was more than minor because it adversely affected the Mitigation Systems cornerstone objective of ensuring the reliability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined that the function for core decay removal was affected, since the safety function of the ADS valves is to depressurize the reactor to allow for low pressure coolant injection. The inspectors determined that this finding was not greater than Green, because subsequent laboratory analysis and engineering evaluation documented in Entergy Operability Recommendation VTY 2011-0631 concluded that sufficient margin was available in the safety-class backup supply to the pneumatic actuation system. The inspectors reviewed Entergy\'s laboratory results and Operability Recommendation, and concluded that the ADS function would have been met under the worst case leakage for all design basis conditions | |
| 05000271/FIN-2011002-04 | 2011Q1 | Green | Licensee-identified | None | Technical Specification 3.6.0, Safety and Relief Valves, requires the reactor to be shut. down and pressure brought below 150 psig within 24 hours with two (2) or more SRVs inoperable. Contrary to the above, Entergy determined that two (2) of the four (4) SRVs were inoperable for a period of time greater than allowed by Technical Specifications. This determination was based on pneumatic actuator thread seal leakage that was identified during testing of the pneumatic SRV actuators in the 2010 refueling outage. Entergy determined the leakage was in excess of design requirements, thereby rendering the SRV manual depressurization function inoperable. This condition has been entered in the licensee\'s corrective action program (CR-VTY-2010-2187) and corrective actions have been developed. The inspectors determined that this finding was more than minor because it adversely affected the Mitigation Systems cornerstone objective of ensuring the reliability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined that the function for core decay heat removal was affected, since the ability to manually discharge steam from core decay heat to the suppression pool was degraded by the thread seal leakage. The inspectors determined that this finding is not greater than Green, because subsequent laboratory analysis and engineering evaluation documented in Entergy Operability Recommendation VTY 2011-0631 concluded that sufficient margin was available in the safety-class backup supply to the pneumatic actuation system. The inspectors reviewed Entergy\'s laboratory results and Operability Recommendation, and concluded that the SRV manual depressurization function would have been met under the worst case leakage for all design basis condition | |
| 05000271/FIN-2011002-05 | 2011Q1 | Green | Licensee-identified | None | 10 CFR 50.65(a)(4) requires, in part, that before performing maintenance activities, the. licensee shall assess and manage the increase in risk that may result from proposed maintenance activities. Contrary to the above, on January 3, 2011, Entergy did not adequately assess and manage the increase in risk due to proposed emergent maintenance activities. This resulted in a non-conservative risk assessment and failure to take all of the appropriate risk management actions for the actual plant conditions. Entergy identified this after the emergent maintenance activities had been completed, and entered the issue into their corrective action program (CR-VTY-2011-00028) to evaluate for appropriate corrective actions. The finding is more than minor because it is similar to IMC 0612, Appendix E, Example 7.e; in that, the overall elevated plant risk put the plant in a higher licensee-established risk category. The finding was evaluated using IMC 0609 Appendix K, \"Maintenance Risk Assessment and Risk Management Significance Determination Process,\" and was determined to be of very low safety significance (Green) because the Incremental Core Damage Probability Deficit between the actual plant conditions and the incorrect risk assessment for the duration of the activity was less than 1.0 E-6 (approximately 3.3 E-9) | |
| 05000271/FIN-2011005-01 | 2011Q4 | Green | H.2 | Self-revealing | Inadvertent Trip of the \\\"A\\\" Emergency Diesel Generator Fuel Rack | A self-revealing NCV of very low safety significance of 10 CFR 50 Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified because Entergy personnel used instructions that were not appropriate to the circumstances, resulting in an inadvertent trip of the A emergency diesel generator (EDG) fuel rack. Entergy\\\'s corrective actions included promptly restoring the A EDG to an operable state, removing the qualifications for the auxiliary operator and field support supervisor involved in the event, and initiating CR-VTY-2011-05483. The inspectors determined that the inadvertent trip of the A EDG fuel rack by Entergy personnel was a performance deficiency that was reasonably within Entergy\\\'s ability to foresee and prevent. This finding is more than minor because it is associated with the Human 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 (i.e. core damage). Specifically, the inadvertent trip of the A EDG fuel rack resulted in the unplanned unavailability of the A EDG for approximately two minutes. The inspectors determined the significance of the finding using IMC 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The finding was determined to be of very low safety significance (Green) because it did not represent a loss of system safety function, a loss of safety function of a single train for greater than its technical specification allowed outage time, and did not screen as potentially risk significant due to external initiating events. The inspectors determined that this finding had a crosscutting aspect in the Human Performance cross-cutting area, Work Practices component, because Entergy did not ensure supervisory oversight of work activity such that nuclear safety was supported |
| 05000271/FIN-2011005-02 | 2011Q4 | Green | H.7 | Self-revealing | Loss of Shutdown Cooling due to Tag-Out Error | A self-revealing NCV of very low safety significance of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified because drawing B191301, Sheet 576, Control Wiring Diagram - Emergency Heater Drain Valve Diagram was not of the appropriate quality to allow tagging activities to be accomplished in accordance with the drawing. As a result of the inadequate drawing, the wrong breaker was selected to be tagged out, which resulted in an unexpected loss of shutdown cooling for 12 minutes. Entergy took immediate corrective action to restore shutdown cooling and entered this issue into their corrective action program (CR-VTY-2011-04203). The inspectors determined that Entergy\\\'s tag-out of the distribution breaker to Vital AC subpanel A due to a drawing error was a performance deficiency that was reasonably within Entergy\\\'s ability to foresee and correct. This finding is more than minor because it is similar to the more than minor statement in example 4.b. of IMC 0612, Appendix E, Examples of Minor Issues, where an operator inadvertently operated the wrong component and caused a transient. Additionally, the finding is more than minor because it affects the objective of the Initiating Events cornerstone to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors determined that this finding was of very low safety significance (Green), using IMC 0609, Appendix G, Checklist 7, BWR Refueling Operation with RCS Level >23\\\'. This determination was based on the fact that the finding did not degrade Entergy\\\'s ability to recover decay heat removal once lost, and that the temperature increase was small enough that it did not represent a loss of control. The inspectors determined that this finding had a cross-cutting aspect in the Human Performance cross-cutting area, Resources component, because components in the tagging database were not labeled correctly |
| 05000271/FIN-2011005-03 | 2011Q4 | Green | H.5 | Self-revealing | Incomplete Inventory for Spent Resin Shipment | A self-revealing NCV of very low safety significance of 10 CFR 20.1501 and 10 CFR 20.2006(b) was identified because Entergy personnel failed to indicate an accurate total of radionuclide activity on the manifest for a radioactive waste shipment on September 19, 2011. Radiation surveys by the receiving personnel at the radioactive waste processing facility identified radiation levels exceeding those indicated on the shipping manifest. Subsequently, Entergy personnel determined that the total radionuclide activity for the shipment was 17 curies instead of 13.4 curies as originally documented. Entergy staff initiated CR-VTY-2011-03902, revised the NRC Form 541, and sent the revision to the radioactive waste processor to correct this error. The inspectors determined that the failure to indicate an accurate total of radionuclide activity on the manifest for a radioactive waste shipment was a performance deficiency that was reasonably within Entergy\\\'s ability to foresee and correct. This finding is more than minor because it affects the Public Radiation Safety cornerstone objective to ensure adequate protection of public health and safety from exposure to radioactive materials released into the public domain as a result of routine civilian nuclear reactor operation. Specifically, the failure to accurately account for all of the radioactive wastes in shipment No. 2011-85 had the potential for misclassifying wastes non-conservatively in subsequent radioactive waste processing and final shipment activities to a low level burial ground facility. The inspectors evaluated the finding using IMC 0609, Appendix D, Public Radiation Safety Significance Determination Process. The inspectors determined the finding to be of very low safety significance (Green) because the error was corrected at the waste processor rather than after shipment to a waste disposal facility, and did not affect low level burial ground nonconformance as evaluated under 10 CFR 61, Licensing Requirements for Land Disposal of Radioactive Wastes. Additionally, there were no radiological consequences (dose) to the public as a result of the shipping manifest error. The inspectors determined that this finding had a cross-cutting aspect in the Human Performance cross-cutting area, Work Control component, because Entergy did not appropriately coordinate work activities by incorporating actions to address the need for interdepartmental coordination and communication. Specifically, the impact of flushing a reactor water cleanup resin transfer line was not sufficiently communicated or coordinated by all groups to ensure all solid radioactive wastes discharged from the plant into the waste container were accounted for in a subsequent radioactive waste shipment |
| 05000271/FIN-2011403-01 | 2011Q1 | Green | H.9 | NRC identified | Security | |
| 05000271/FIN-2011404-01 | 2011Q2 | Green | H.12 | Self-revealing | Security | |
| 05000271/FIN-2012002-01 | 2012Q1 | Green | H.14 | Self-revealing | Failure of the B UPS Techometer Coupling Due to Age and Inadequate Corrective Actions | A self-revealing, Green, NCV of 10 CFR 50 Appendix B, Criterion XVI, Corrective Action, was identified because Entergy did not promptly correct an adverse condition resulting in the failure of the B uninterruptible power supply (UPS) motor generator (MG) set direct current (DC) tachometer coupling. Specifically, Entergy personnel did not promptly replace or verify the physical condition of the B tachometer coupling when it was known that it was aged and susceptible to age-related failure. Entergys corrective actions included replacing the B tachometer coupling, establishing a 12 year preventive maintenance replacement frequency, and initiating CR-VTY-2011-03686, CR-VTY-2011- 03744, CR-VTY-2011-05335, CR-VTY-2011-05337, and CR-VTY-2012-01096. The inspectors determined that the issue was 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 and reliability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Specifically, the B UPS MG set failed in service, affecting the overall system redundancy and reliability, and resulted in 22 hours of unplanned unavailability. The inspectors determined the significance of the finding using IMC 0609.04, Phase 1 Initial Screening and Characterization of Findings. The finding was determined to be of very low safety significance (Green) because it did not represent a loss of system safety function, a loss of safety function of a single train for greater than its technical specification (TS) allowed outage time (UPS-1B), and did not screen as potentially risk significant due to external initiating events. The inspectors determined that the finding had a cross-cutting aspect in the Human Performance cross-cutting area, Decision-Making component, because Entergy personnel did not use conservative assumptions in decision making and did not adopt a requirement to demonstrate that the proposed action to delay the coupling replacement until June 2012 was safe. |
| 05000271/FIN-2012002-02 | 2012Q1 | Green | P.1 | Self-revealing | Failure to the D Service Water Pump Due to LOW Oil and Inadequate Corrective Actions | A self-revealing, Green, NCV of 10 CFR 50 Appendix B, Criterion XVI, Corrective Action, was identified because Entergy personnel did not promptly correct an adverse condition resulting in the unplanned unavailability of the D service water pump. Specifically, Entergy personnel did not maintain a clear oil sight glass and did not identify a low oil level for the upper motor bearing prior to damage to the bearing. Entergys corrective actions included initiating CR-VTY-2012-00483, performing an apparent cause evaluation (ACE), and replacing the motor and sight glass. Entergy staff completed the D service water pump work and restored it to service. The inspectors determined that the issue was 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 and reliability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Specifically, the D service water pump failed in service affecting overall safety system redundancy and reliability, and resulted in three days of unplanned unavailability. The inspectors determined the significance of the finding using IMC 0609.04, Phase 1 Initial Screening and Characterization of Findings. The finding was determined to be of very low safety significance (Green) because it did not represent a loss of system safety function, a loss of safety function of a single train for greater than its TS allowed outage time, and did not screen as potentially risk significant due to external initiating events. The inspectors determined that this finding had a cross-cutting aspect in the Problem Identification and Resolution cross-cutting area, Corrective Action Program component, because Entergy personnel did not implement a corrective action program with a low threshold for identifying issues and as a result, the stained sight glass was not recognized as an adverse condition |
| 05000271/FIN-2012003-01 | 2012Q2 | Green | H.1 | NRC identified | Inadequate Risk Assessment for Isolating the Condensate Pumps\' Minimum Flow Line\'S Automatic Flow Control Valve | The inspectors identified an NCV of 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, paragraph (a)(4), for Entergys failure to conduct an adequate risk assessment prior to isolating the condensate pumps minimum flow automatic control valve. Specifically, the inspectors identified that Entergy personnel had not analyzed the impact to plant risk with the condensate pumps minimum flow line to the main condenser isolated. Entergys corrective actions included declaring and announcing to site personnel the plant risk to be Orange, protecting further equipment, and initiating CR-VTY-2012-02074. The inspectors determined that the issue was more than minor because it is similar to IMC 0612, Power Reactor Inspection Reports, Appendix E, Examples of Minor Issues, example 7.e in that the overall elevated plant risk put the plant into a higher risk category established by Entergy. The inspectors determined the significance of the finding using IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process. The finding was determined to be of very low safety significance (Green) because the Incremental Core Damage Probability Deficit for the timeframe that the condensate pumps were unavailable was less than 1E-6 (approximately 2E-7). The inspectors determined that the finding had a cross-cutting aspect in the Human Performance cross-cutting area, Resources component, because the equipment relied upon to perform the risk assessment, the equipment out of service software program (EOOS), did not include the condensate system automatic minimum flow control valve, which was not adequate to ensure nuclear safety. |
| 05000271/FIN-2012003-02 | 2012Q2 | Green | H.7 | NRC identified | Inadequate Risk Assessment Due to Not Considering the Increased Risk of a Plant Transient When Securing a Feedwater Pump | The inspectors identified an NCV of 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, paragraph (a)(4) for Entergys failure to conduct an adequate risk assessment prior to securing the C feedwater pump. Specifically, the inspectors identified that Entergy personnel had not analyzed the impact to plant risk of securing the C feedwater pump. Entergys corrective actions included briefing operators that securing a feedwater pump was a HRE-TRAN, i.e. an activity considered to raise the likelihood of an initiating event that is likely to result in a plant trip, and initiating CR-VTY-2012-02160 and CR-VTY-2012-02894. The inspectors determined that the issue was more than minor because it is similar to IMC 0612, Power Reactor Inspection Reports, Appendix E, Examples of Minor Issues, example 7.e in that the overall elevated plant risk put the plant into a higher risk category established by Entergy. The inspectors determined the significance of the finding using IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process. The finding was determined to be of very low safety significance (Green) because the Incremental Core Damage Probability Deficit for the timeframe that the C feedwater pump was being secured was less than 1E-6 (approximately 4E-9). The inspectors determined that the finding had a cross-cutting aspect in the Human Performance cross-cutting area, Resources component, because the procedure describing HRE-TRAN was not sufficiently clear and complete in its description to ensure nuclear safety. |
| 05000271/FIN-2012004-01 | 2012Q3 | Green | H.8 | Self-revealing | Incorrect Assessment of Equipment Condition Resulted in Single Recirculating LOOP Operation | A self-revealing, Green finding (FIN) was identified because Entergy failed to implement a preventive maintenance procedure. Specifically, Entergy personnel classified the discovery status code for the minor motor inspection on the A recirculation pump motor generator set drive motor incorrectly, as B satisfactory or normal wear, instead of D abnormal wear, which resulted in a missed opportunity to replace degraded components that caused the A recirculation pump to trip and an unplanned entry into single recirculation loop operation. Entergys corrective actions included cleaning the motor and the junction box, replacing components that had been damaged by an arc flash, and testing the circuit to verify no other components were degraded prior to restarting the motor. In addition, Entergy initiated condition report CR-VTY-2012-02811 and issued a corrective action to reinforce the requirements of Entergy Procedure EN-DC-324 among maintenance staff. Entergy also plans to add all large motor and generator junction boxes to the predictive maintenance program and to perform thermography on a six month frequency. The inspectors determined that the issue was more than minor because it resulted in a transient, i.e. an event that upset plant stability (an unplanned entry into single recirculation loop operation). In particular, the issue is associated with the Equipment Performance attribute of the Initiating Events cornerstone and affected the cornerstone objective to limit the likelihood of events that upset plant stability during power operations. The inspectors determined the significance of the finding using IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power. The finding was determined to be of very low safety significance (Green) because the finding was a transient initiator that did not cause a reactor trip. The inspectors determined that the finding had a cross-cutting aspect in the Human Performance cross-cutting area, Work Practices component, because Entergy did not sufficiently define and effectively communicate expectations regarding procedural compliance for the selecting of the discovery status code and personnel did not follow procedures. |
| 05000271/FIN-2012004-02 | 2012Q3 | Green | H.9 | NRC identified | Dedicated Operators Required for Operability Under Applied Administrative Controls Left Immediate Vicinity of Open Valves | The inspectors identified an NCV of technical specification (TS) 6.4, Procedures, for Entergys failure to implement a surveillance activity in accordance with the written procedure. Specifically, the inspectors identified that during a surveillance test, dedicated operators required to maintain operability of primary containment left the immediate vicinity of open manual containment isolation valves. Entergys corrective actions included restoring the administrative controls required to maintain primary containment operability during the subject surveillance test, initiating condition report CR-VTY-2012-03561, sending a memorandum to and discussing the issue with all operating crew shift managers explaining the error and the requirements of a dedicated operator, and issuing a temporary night order further explaining these requirements. Additional corrective actions included implementing and tracking training for all operators on these requirements, and revising licensed operator training on primary containment to specifically describe these requirements. The inspectors determined that the issue was more than minor because it is associated with the Human Performance 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, the dedicated operators were required to be stationed in the immediate vicinity of the valve controls to rapidly close the valves when primary containment isolation is required during accident conditions, but the operators were significantly beyond the required immediate vicinity when they left the reactor building. The inspectors determined the significance of the finding using IMC 0609, Appendix H, Containment Integrity Significance Determination Process. The finding was determined to be of very low safety significance (Green) using Appendix H, Table 6.2, Phase 2 Risk Significance Type B Findings at Full Power, because primary containment was inoperable for 37 minutes, i.e. less than 3 days. The inspectors determined that this finding had a cross-cutting aspect in the Human Performance cross-cutting area, Resources component, because the training of personnel did not describe specific requirements of dedicated operators, including the definition of immediate vicinity. |
| 05000271/FIN-2012005-01 | 2012Q4 | H.14 | Self-revealing | Failure of the B Emergency Diesel Generator from Jacket Water Leakage Due to Inadequate Corrective Action | A self-revealing apparent violation (AV) of 10 CFR 50 Appendix B, Criterion XVI, Corrective Action, was identified because Entergy did not promptly correct an adverse condition resulting in the failure of the B emergency diesel generator. Specifically, Entergy personnel did not promptly replace a degraded jacket water flange gasket prior to its subsequent failure. Entergys corrective actions included replacing the gasket, visually inspecting the other jacket water connections, and initiating condition report CR-VTY-2012- 05044. The finding 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 and reliability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the B emergency diesel generator failed in service due to a known degraded condition that affected the overall system redundancy and reliability and resulted in 37 days of unplanned unavailability. The significance of the finding is designated as To Be Determined (TBD) until a Phase 3 analysis can be completed. The finding had a cross-cutting aspect in the Human Performance, Decision-Making because Entergy personnel did not use conservative assumptions in decision making in that the chosen action was to monitor the leak for a prolonged period of time. | |
| 05000271/FIN-2012005-02 | 2012Q4 | Green | Licensee-identified | Licensee-Identified Violation | 10 CFR 50, Appendix B, Criterion III, Design Control, requires, in part, that the design basis is correctly translated into specifications. Contrary to the above, the design basis was not correctly translated into specifications in that the specification for the mechanical flood seal used in spare four inch conduit was not adequate such that a design basis flood could have penetrated the conduit and allowed water intrusion into the switchgear rooms. Entergy entered this issue into the corrective action program as CR-VTY-2012-02391. The inspectors determined that the finding was of very low safety significance (Green) because the missing conduit seal would not cause a plant trip or an initiating event, degrade two or more trains of a multi-train system, degrade one or more trains of a system that supports a risk significant system, or involve the total loss of any safety function. Specifically, Entergy procedures direct a plant shutdown and staging of portable pumps to remove water from the manholes within the switchgear rooms during a design basis flood. The calculated flow rate of water through the conduit was bounded by the capacity of the two portable pumps. | |
| 05000271/FIN-2013002-01 | 2013Q1 | Green | P.1 | NRC identified | Appendix R Fire Door Not Latching Closed Due to Misalignment | The inspectors identified an NCV of operating license condition 3.F, fire protection program, because Entergy did not correct a degraded latch on a three-hour rated fire door on the entrance to the B emergency diesel generator (EDG) room, and as a result the three-hour fire barrier was non-functional and the required compensatory measure of an hourly fire watch was not in effect. Entergys corrective actions included restoring vertical alignment of the latching mechanism, further inspection by a locksmith to ensure reliable operation, planning a preventive replacement of the latch due to existing excessive wear, and initiating a condition report. The finding is more than minor because it is associated with the protection against external factors attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the reliability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the fire door being degraded with unreliable latching without an assigned hourly fire watch from January 20 to January 22 resulted in a barrier to fire propagation that was less robust than required by the approved fire protection program. In accordance with IMC 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix F, Fire Protection Significance Determination Process, the inspectors determined that this finding is of very low safety significance (Green) per Task 1.3.2, Task 1.3.2: Supplemental Screening for Fire Confinement Findings. The inspectors determined the degradation rating associated with the deficiency to be Moderate B since a closure mechanism held the door against the door jamb, the door swings out from the EDG room, no combustibles were stored in the adjacent hallway, and no equipment important to safety exists in the turbine building hallway. Therefore, the degraded fire door provided a minimum of 20 minutes of fire endurance protection and the fixed or in situ fire ignition sources and combustible or flammable materials were positioned such that, even considering fire spread to secondary combustibles, the degraded fire door would not have been subject to direct flame impingement since no combustible material was located near the door during the time of concern. The inspectors determined that the finding had a crosscutting aspect in the Problem Identification and Resolution area, Corrective Action Program component, because Entergy personnel did not completely identify the issue with the alignment of the striker plate when the degradation was first identified and did not identify that the latching deficiency still existed during subsequent transits through the door. |
| 05000271/FIN-2013002-02 | 2013Q1 | Green | H.5 | NRC identified | Failure to Implement Compensatory Measures Associated with a Temporary Modification | The inspectors identified an NCV of Technical Specification 6.4, Procedures, because procedure OPOP-SW-2181, Service Water/ Alternate Cooling System, was inadequate. Specifically, the step in the procedure to identify and isolate sources of water lost from the cooling tower basin would not have been implemented in a timely manner while a temporary fire water system was drawing on the basin. Entergys corrective actions included writing a night order describing the fire fighting strategy for a fire in the intake and directing the temporary fire pumps to be stopped if they started automatically while the alternate cooling system (ACS) was in service, implementing temporary procedure changes, and initiating a condition report. The finding is more than minor because it impacted the design control attribute of the Mitigating Systems cornerstone. Specifically, the temporary modification added another potential path for loss of water from the cooling tower deep basin and the appropriate compensatory measures to address that loss path were not implemented, impacting the capability and reliability of ACS. Additionally, the finding is similar to IMC 0612, Appendix E, Examples of Minor Issues, example 3.j more than minor description, because the added draw on the cooling tower basin water had the potential to affect the accident analysis calculation assumption of the amount of water available for running ACS. 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 is of very low safety significance (Green) because the performance deficiency did not involve the total loss of a safety function that contributes to external event initiated core damage accident sequences. This condition existed for less than the technical specification allowed outage time of seven days. This finding had a cross-cutting aspect in the area of human performance, Work Control, because Entergy did not appropriately coordinate work activities by incorporating actions to address the need to keep personnel apprised of the operational impact of work activities. Specifically, Entergy identified the need for compensatory measures for the temporary modification for the fire water system work, but the necessary actions were not coordinated to ensure operations and maintenance understood the operational impact of the work. |