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| Q | Significance | CCA | Identified by | Title | Description | |
|---|---|---|---|---|---|---|
| 05000271/FIN-2014005-03 | 2014Q4 | Green | Self-revealing | Failure to Follow Procedure Results in Inoperable Emergency Diesel Generator | The inspectors identified 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, because Entergy did not properly implement the prescribed maintenance instructions during the installation of the copper o-ring to the Kiene valve adapter of cylinder number six of the A emergency diesel generator (EDG) on June 25, 1992. Entergys corrective actions included initiating condition report CR-VTY-2014-3503, performing a root cause evaluation, and removing and reinstalling the number six cylinder adapters using new copper o-ring gaskets with the correct applied torque. Entergy restored the A EDG to operable status on October 2, 2014. Entergy also performed cylinder air testing of all cylinders on both A and B EDGs to ensure no other leaks existed. This finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to properly install the Kiene valve adapter on June 25, 1992, resulted in the A EDG failing to start on September 29, 2014, during a quarterly surveillance. Using IMC 0609, Significance Determination Process, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that this finding required a detailed risk evaluation (DRE) because the failure of the A EDG to start on September 29, 2014, represented an actual loss of function of a single train of emergency alternating current power for greater than the technical specification allowed outage time. The Region I senior reactor analyst (SRA) used the Systems Analysis Programs for Hands-On Evaluation (SAPHIRE) Revision 8.1.0 and the Standardized Plant Analysis Risk (SPAR) Model for VY, Version 8.19, to conduct the internal events DRE and VYs Individual Plant Examination (IPE) for Severe Accident Vulnerabilities and the Individual Plant Evaluation External Events (IPEEE) to assess the external events risk contribution for this performance deficiency. The SRA made the following assumptions and SPAR model changes to best represent the condition of the A EDG: The exposure time was estimated using T/2 because the actual date and time the EDG became inoperable is indeterminate based upon the uncertainty of gasket coolant leak rate into the cylinder. The time between the last successful operation of the EDG and the observed failure was 36 days. Therefore, T/2 most accurately approximates the postulated exposure time, T/2 = 36/2= 18 days. Adding the unavailability time due to corrective maintenance, the total exposure time for this finding is 21 days. Basic event EPS-DGN-FS-DGA, Diesel Generator A Fails to Start, was set to True, consistent with the Risk Assessment Standardization Project guidance, to reflect the potential for a common cause failure mechanism of both the A and B EDGs. Based upon the observed success of plant staff to recover the EDG following the failure to start event on September 29, 2014, and the available procedure guidance and training, the SRA assumed recovery of the A EDG could reasonably be achieved under accident conditions. Accordingly, the EDG recovery basis event, EPS-XHE-XL-NR01H, nominal probability (8.71E-1) was revised to 0.1 (1 in 10 chance of not successfully recovering the EDG). Based upon the above stated assumptions, the increase in internal risk core damage frequency (delta CDF) associated with this performance deficiency is in the mid E-7 range, or very low safety significance (Green). The SRA examined the IPEEE and associated fire safe shutdown analysis/procedures to determine the external event risk contribution due to this finding. The SRA determined that the A EDG is credited for safe shutdown in the event a fire compromises the West Switchgear Room. Assuming worst case fire conditions, without suppression, the external fire contribution per IPEEE, Table 4.10.1 would be approximately 3.4E-7 (5.9E-6 X 21/365). The SRA also determined that the unavailability of the A EDG has no appreciable impact on seismic, flooding, and high winds associated mitigation capability. Combining the internal and external risk contributions yields a total delta CDF of high E-7 (Green). Based upon a review of the dominant cutsets (loss of offsite power initiating events with subsequent failures of high pressure injection and depressurization), the unavailability of the A EDG may result in an increased risk associated with a Large Early Release Frequency (LERF). The SRA used a 0.1 LERF factor to account for the probability that operators would procedurally take action to mitigate the consequences of a potential containment breach due to these postulated high pressure accident sequences. Consequently, the delta LERF for this finding is high E-8 (Green). The inspectors determined that the finding did not have a cross-cutting aspect because the performance deficiency did not occur within the last three years and would not likely occur today under similar circumstances. | |
| 05000271/FIN-2014005-02 | 2014Q4 | Green | H.9 | NRC identified | Misclassification of Emergency Conditions Based on Radiological Criteria | The inspectors identified a Green NCV of 10 CFR 50.54(q), Emergency Plans, because Entergys emergency action level (EAL) classification process could result in a misclassification, a deficiency related to a risk significant planning standard. Specifically, Entergy personnel failed to correctly classify emergency conditions during an emergency preparedness drill using the applicable EAL criteria. Entergy initiated CR-VTY-2014-3990 for not recognizing the proper application of Note 1 contained in EAL AG1.1, and subsequently trained operating crews and emergency response organization (ERO) decision makers on the basis and intent of Note 1. Entergy reviewed the implementing procedures and associated basis documents and determined that applicable procedures and guidance properly described and implemented the required hierarchy of EAL AG1.1 and EAL AG1.2. This finding is more than minor because it is associated with the ERO attribute of the Emergency Preparedness cornerstone and adversely affected the cornerstone objective of ensuring that Entergy is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Specifically, the failure to establish an adequate EAL classification process could result in a general emergency (GE) declaration when a GE is not warranted. The inspectors determined the significance of the finding using IMC 0609, Appendix B, "Emergency Preparedness Significance Determination Process." The finding was determined to be of very low safety significance (Green) because of the following: the drill scenario and applicable sequence of EALs applied solely to damage to spent fuel which has a lower volatile radiological inventory than fuel in the reactor, the language in EAL AG 1.1 Note 1 uses the word "should" rather than "shall," and the EAL scheme provides for the use of judgment when meeting an EAL threshold is believed to be imminent, and the likelihood of actual emergency conditions reaching EAL AG1.1 and not reaching EAL AG1.2 is considered very low. Specifically, the inspectors considered the above extenuating circumstances and evaluated the finding in accordance with IMC 0609, Appendix B, Section 5.4, "10 CFR 50.47(b)(4), Emergency Classification System," and concluded that the EAL issue could result in an early GE declaration, but the finding more closely fit the "would result in unnecessary classification" significance category rather than the "would result in unnecessary (protective action recommendations) PARS for the public" significance category. The inspectors determined that the finding has a cross-cutting aspect in the area of Human Performance, Training, because Entergy did not ensure the knowledge and training of the trainers and ERO personnel sufficiently conveyed realistic application of radiological based EAL criteria which properly balanced the risks of radiological dose consequences with the risks associated with unnecessary protective action recommendations. |
| 05000271/FIN-2014005-01 | 2014Q4 | Green | P.3 | NRC identified | Compensatory Continuous Fire Watches Not Implemented as Required | The inspectors identified a Green NCV of operating license condition 3.F, Fire Protection Program, because Entergy did not implement and maintain in effect all provisions of the NRC approved fire protection program. Specifically, Entergy did not implement and maintain the required compensatory continuous fire watch in the east and west switchgear rooms when the fire detection and suppression systems were not functional for planned maintenance. Entergys corrective actions included stationing separate continuous fire watches in the east and west switchgear rooms, initiating condition report CR-VTY-2014- 4019, communicating and reinforcing the fire watch requirements with all operating crews and maintenance personnel, and initiating additional training. This 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 east and west switchgear rooms are separate fire areas each containing separated trains of Appendix R safe shutdown equipment which are required to respond to initiating events to prevent undesirable consequences and the implemented compensatory fire watches were less than required by the fire protection program. The inspectors used IMC 0609, Appendix F, Fire Protection Significance Determination Process, to analyze this finding because the condition had an adverse effect on the Fire Prevention and Administrative Controls Program element in accordance with the degradation rating guidance. The inspectors determined that the finding screened as Green because the impact of the fire finding was limited to no more than one train of equipment important to safety at any given time. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Resolution, because Entergy did not take effective corrective actions when this issue was self-identified in a condition report on January 5, 2013. |
| 05000271/FIN-2014004-04 | 2014Q3 | 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 specified current rating for electrical switches supplying electric heaters in the standby gas treatment system were less than the designed circuit amperage from original plant construction, February 28, 1973, until November 13, 2013. Entergy identified that the standby gas treatment auxiliary switches supplying the charcoal bed heaters in both A and B subsystems were rated for a continuous current of 3 A when the continuous current was approximately 8 A. Entergy entered this issue into the corrective action program as CR-VTY-2013-06257. 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 only represented a degradation of the radiological barrier function provided for the standby gas treatment system. | |
| 05000271/FIN-2014004-03 | 2014Q3 | Green | H.8 | Self-revealing | Failure to Follow Procedure Results in Inoperable Containment Isolation Valve | The inspectors identified a self-revealing Green NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because Entergy staff did not implement the prescribed maintenance instructions during the refurbishment of the airoperated valve (AOV) actuator for a drywell floor drain containment isolation valve. Specifically, Entergy staff used a lubricant other than the type specified per the equipment manual, which was incompatible with the seals in the valve. Entergys immediate corrective actions included entering the issue into their corrective action program as CR-VTY-2013- 05763, performing a rebuild of the valve, and troubleshooting the as-found condition. This finding is more than minor because it is associated with the SSC and barrier performance attribute of the Barrier Integrity cornerstone and affected the cornerstone objective to provide reasonable assurance that physical design barriers (e.g., containment) protect the public from radionuclide releases caused by accidents or events. Specifically, when tested, the valve exceeded the maximum allowable stroke time for closure and was declared inoperable. 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, issued June 19, 2012, the inspectors determined that this finding is of very low safety significance (Green) because it was associated with the functionality of the reactor containment but did not represent an actual open pathway in the physical integrity of containment, containment isolation system, and heat removal components. This finding has a cross-cutting aspect in the area of Human Performance, Procedure Adherence, because Entergy personnel did not properly implement the requirements prescribed in the maintenance instructions. Specifically, during the refurbishment of the valves actuator, Entergy staff did not use the lubricant specified in the equipment manual referenced in the work order. |
| 05000271/FIN-2014004-01 | 2014Q3 | Green | H.1 | NRC identified | Failures to Promptly Identify Through-Wall Leakage from Service Water Piping to the Emergency Diesel Generators | The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, because Entergy did not promptly identify conditions adverse to quality related to the service water system. Specifically, observable through-wall leaks that were reasonably able to be identified existed in service water piping supplying the emergency diesel generators (EDGs) cooling system for an extended period of time without being identified. In addition, the affected service water piping was not appropriately scheduled for treatment and replacement given known conditions favorable to microbiologically induced corrosion (MIC). Entergys corrective actions to restore compliance consisted of performing complete walkdowns of all accessible safety-related service water piping, performing ultrasonic inspections of the three leak locations and fifteen extent of condition locations, conducting structural analyses to determine structural integrity of the piping with the measured thinning, and performing daily leak rate monitoring and frequent periodic ultrasonic inspections of no more than 30 day intervals. This finding is more than minor because if left uncorrected it has the potential to lead to a more significant safety concern. Specifically, the through-wall leaks were unmonitored degraded conditions with reasonable doubt on the operability of the service water and alternate cooling systems before the results of ultrasonic inspections and new structural analyses were obtained. The inspectors determined the significance of the finding using IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At- Power. The finding was determined to be of very low safety significance (Green) because the finding was a deficiency affecting the design and qualification of the service water and alternate cooling systems and the systems maintained their operability. The inspectors determined that the finding has a cross-cutting aspect in the area of Human Performance, Resources, because Entergy did not ensure that the combination of piping replacements, chemical treatments, guidance and procedures for walkdowns, and camera coverage were adequate to support nuclear safety. |
| 05000271/FIN-2014004-02 | 2014Q3 | Green | H.12 | NRC identified | Failure to Submit Reactor Building Crane Digital Control System Modification for Approval | The inspectors identified a finding of very low safety significance (Green) and an associated Severity Level IV NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50.59, Changes, Tests and Experiments, when Entergy made changes to the reactor building crane that resulted in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component (SSC) important to safety previously evaluated in the Updated Final Safety Analysis Report (UFSAR). Specifically, Entergy did not recognize that they had removed redundancy from the control system needed to qualify the crane as single-failure proof. Entergy entered this issue into their corrective action program as condition report (CR)-VTY-2014-03028 and completed modifications to the crane that restored the independence of the redundant upper travel limits. The inspectors determined that the finding was more than minor because the change would have required NRC review and approval in order to qualify the crane as single-failure proof. Additionally, this finding was associated with the design control attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers (e.g. fuel cladding) protect the public from radionuclide releases caused by accidents or events. Specifically, the design change increased the likelihood of a heavy load drop, which could have impacted the fuel in the spent fuel pool. This issue impeded the ability of the NRC to perform its regulatory oversight function because the failure to follow the requirements in 10 CFR 50.59 resulted in Entergy not submitting the change to the NRC for approval. Therefore, the enforcement aspects of this finding were processed using the Traditional Enforcement process. This violation is associated with a finding that has been evaluated by the SDP and communicated with an SDP color reflective of the safety impact of the deficient licensee performance. The SDP, however, does not specifically consider the regulatory process impact. Thus, although related to a common regulatory concern, it is necessary to address the violation and finding using different processes to correctly reflect both the regulatory importance of the violation and the safety significance of the associated finding. The inspectors evaluated this finding using IMC 0609, Attachment 4, Initial Characterization of Findings. The inspectors determined that the finding affected the Barrier Integrity cornerstone and evaluated the finding using Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 3, Barrier Integrity Screening Questions. The inspectors determined the finding was of very low safety significance (Green) because the crane was not operated over the spent fuel pool, nor was there an actual load drop. Per Subsection d.2 of Section 6.1, Reactor Operations, of the NRC Enforcement Policy, this is a Severity Level IV violation, because it is a 10 CFR 50.59 violation that results in conditions evaluated as having very low safety significance by the SDP. This finding has a cross-cutting aspect in the area of Human Performance, Avoid Complacency, because Entergy did not avoid complacency on the review of this design by recognizing and planning for the possibility of latent issues. The 50.59 screening was not reviewed to ensure it fully captured the final design from the vendor, and as a result, the vulnerability introduced by the digital controller was not considered. |
| 05000271/FIN-2014403-01 | 2014Q2 | Green | H.14 | NRC identified | Security | |
| 05000271/FIN-2014403-02 | 2014Q2 | Green | H.7 | NRC identified | Security | |
| 05000271/FIN-2014007-01 | 2014Q2 | Green | H.8 | NRC identified | Inadequate Design Control of SBO Loading Calculation | The team identified a finding of very low safety significance (Green), in that Entergy did not ensure correct implementation of their design control process when establishing the capacity requirement for the new Station Blackout (SBO) alternate alternating current (AAC) power source. Specifically, Entergy did not use the latest revision of the SBO load capacity analysis as a design input to the load capacity requirement when verifying the adequacy of the sizing of the new SBO diesel generator (DG). Entergy entered the issue into their corrective action system to evaluate the capability of the SBO DG to support the expected SBO loads and initiated actions to ensure the design analysis assumptions for loading are consistent with the established operational procedures for SBO response. 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 capability of systems that respond to initiating events to prevent undesirable consequences. In addition, inspectors reviewed IMC 0612, Appendix E, Examples of Minor Issues, and found that example 3.j was similar, in that, the team had reasonable doubt of the capability of the SBO DG to operate within its analyzed load rating. Specifically, the most limiting condition with residual heat removal service water (RHRSW) pumps in service had not been accounted for in the SBO DG load rating evaluation. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Mitigating Systems Screening Questions, Section A, Mitigating SSCs and Functionality, the team concluded that this finding was a design deficiency that did not result in the SBO DG losing its functionality. Specifically, the team evaluated decay heat level requirements and determined there was reasonable assurance the SBO DG load would have remained within its design rating. The team determined that this finding had a cross-cutting aspect in the area of Human Performance, Procedure Adherence, because the design control engineering change process procedure was not adequately followed, in that, the increased SBO load associated with a second RHRSW pump was not evaluated and resolved through the design review process. |
| 05000271/FIN-2014002-01 | 2014Q1 | Green | H.11 | NRC identified | Failure to Monitor the Unavailability of the Fire Water to Service Water Crosstie | The inspectors identified an NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, paragraph a(1), because Entergy did not evaluate the fire protection system for (a)(1) classification even though the unavailability performance criterion had been exceeded. Specifically, Entergy did not recognize that the fire water system to service water system crosstie function was risk-significant and that its unavailability (nine days in 2013 and 34 days in 2014) was required to be monitored. Entergy entered this issue into their corrective action program as condition report CR-VTY-2014-01064. The inspectors determined that the failure to recognize that the fire water system to service water system crosstie function was risk-significant, to monitor the crosstie function?s unavailability (nine days in 2013 and 34 days in 2014), and to evaluate the fire protection system for 10 CFR 50.65 (a)(1) classification was a performance deficiency that was reasonably within Entergy?s ability to foresee and correct, and should have been prevented. This finding is more than minor because it is associated with the configuration control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability of systems that respond to initiating events to prevent undesirable consequences. Specifically, since Entergy personnel did not recognize that the risk significant function was not being tracked against the unavailability performance criterion no actions were taken to address exceeding that criterion and no changes were made to the temporary pump design to reduce additional unavailability. 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 represent an actual loss of function of a non-technical specifications train of equipment designated as high safety-significant for greater than 24 hours. Specifically, the performance deficiency was not the underlying cause of the unavailability in 2013 or 2014. This finding has a cross-cutting aspect in the area of Human Performance because Entergy did not challenge the unknown reason why no system was accruing maintenance rule unavailability while the station was in an elevated risk condition, i.e. Yellow, with the fire water pumps out of service. |
| 05000271/FIN-2013005-02 | 2013Q4 | Green | H.13 | NRC identified | Inadequate Risk Assessment for Isolating All Nitrogen Supply to the Containment Instrument Air System | The inspectors identified a 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 nitrogen supply to the containment instrument air system. Specifically, the inspectors identified that Entergy personnel had not correctly analyzed the impact to plant risk with the liquid nitrogen supply, containment air compressor, and safety relief valve (SRV) nitrogen bottle backup supply removed from service. Entergys corrective actions included establishing a contingency to restore nitrogen supply, protecting further equipment, initiating a condition report, and revising the procedures for drywell entry to maintain the SRV nitrogen backup bottle supply in service until the reactor is shutdown. This finding is more than minor because it is associated with the configuration control 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. The inspectors reviewed IMC 0612, Appendix E, Examples of Minor Issues, and found that example 7.e was similar to the issue. Specifically, the inspectors determined that the issue was more than minor because 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 nitrogen supply system was unavailable was less than 1E-6 (approximately 1E-7). The inspectors determined that the finding had a cross-cutting aspect in Human Performance, Decision- Making component, because Entergy failed to use a systematic process using available risk assessment guidance and did not obtain interdisciplinary input to make a risk-significant decision (H.1(a)). |
| 05000271/FIN-2013005-01 | 2013Q4 | Green | H.7 | Self-revealing | Inadequate Corrective Actions to Restore Switchgear Room Flood Boundary | A self-revealing cited violation of Title 10 Code of Federal Regulations (10 CFR) 50, Appendix B, Criterion XVI, Corrective Action, was identified because Entergy did not promptly correct two separate conditions adverse to quality related to flood protection of the switchgear rooms. Specifically, within one conduit a mechanical screw-type flood seal that rotated in place was removed and not promptly replaced with a reliable foam seal and within a second conduit a mechanical screw-type flood seal was left installed and not promptly replaced with a reliable foam seal, allowing for two flooding pathways into the switchgear rooms. The inadequate seals were identified on March 23, 2013, following water intrusion into the switchgear room manholes, and the NRC documented a Green NCV in Inspection Report (IR) 05000271/2013003, ML13224A068; however, the intended corrective actions were not implemented. This violation is cited because Entergy failed to restore compliance within a reasonable period of time after the initial NCV was identified. On November 7, 2013, Entergy restored compliance by installing a SYLGARD foam seal in both the MH-S2 Spare-4 conduit and MH-S2 40805B conduit. This finding is more than minor because it is associated with the protection against external events attribute of the Mitigating Systems cornerstone and affected the objective to ensure the availability and reliability of systems that respond to external events to prevent undesirable consequences. Specifically, the failed flood barriers provided an external flooding pathway that could impact the reliability and availability of both electrical switchgear rooms during a design basis flood event. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 4 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2012, the inspectors determined that this finding was of very low safety significance (Green) because, in spite of the failed flood barriers, sufficient water removal capability was available to ensure there was no loss of electrical switchgear safety function. The switchgear would still have been able to perform its function because the water level would have been maintained below floor level using the additional sump pump capacity available on site. The inspectors determined that the finding has a cross-cutting aspect in the area of Human Performance, Resources component, because Entergy did not have complete, accurate, and up-to-date design documentation, drawings, and procedures for the switchgear room manhole conduit seals. Specifically, Entergy did not establish a flood seals program and program document, procedure, or drawing that tracked which conduits had mechanical screw-type flood seals and which had SYLGARD foam seals (H.2(c)). |
| 05000271/FIN-2013004-03 | 2013Q3 | Green | H.12 | Self-revealing | Operator Error Results in Diesel Generator Overload | A self-revealing NCV of Technical Specification 6.4, Procedures, was identified because Entergy overloaded the B emergency diesel generator to 130 percent of its sustained load rating. Specifically, an auxiliary operator (AO) took the speed droop switch to zero before the output breaker was opened, contrary to procedure, which resulted in the overload condition. Entergys immediate corrective actions included initiating a condition report, conducting a root cause evaluation, and performing management assessment of control room communications. 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. Specifically, the B emergency diesel generator was unavailable for an additional 24 hours in order to perform required inspections and testing to verify it was not damaged by the overload condition. 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 represent a loss of system safety function or a loss of safety function of a single train for greater than its Technical Specification allowed outage time. The inspectors determined that this finding has a cross-cutting aspect in the Human Performance area, Work Practices component, because Entergy personnel did not use human performance error prevention techniques commensurate with the risk of the assigned task such that work activities were performed safely. Specifically, self-checking, peer checking, and three-part communications were not used effectively to prevent performing procedure steps out of order. |
| 05000271/FIN-2013004-02 | 2013Q3 | Green | H.5 | Self-revealing | Failure to Maintain Radiation Exposure ALARA During Refueling Activities | A self-revealing finding was identified because Entergy inadequately planned and controlled work while performing reactor reassembly and reactor cavity decontamination activities during refueling outage (RFO) 30 resulting in excessive unintended occupational collective exposure that exceeded the planned dose exposure established by Radiation Work Permit (RWP) 2013-702. Inadequate work planning and control resulted in unplanned, unintended collective exposure due to conditions that were reasonably within Entergys ability to control. The work activity performance deficiencies resulted in the collective exposure for these activities increasing from the original estimate of 9.950 person-rem to an actual dose of 18.940 person-rem. Entergy entered the issues into their corrective action program. This finding is more than minor because it is associated with the program and process attribute of the Occupational Radiation Safety cornerstone and affected the cornerstone objective to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Additionally, the performance deficiency was determined to be more than minor based on a similar example (6.i) in Appendix E of IMC 0612, in that the actual collective dose exceeded 5 person-rem and exceeded the planned, intended dose by more than 50 percent. In accordance with IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, the inspectors determined that this finding is of very low safety significance (Green) because the plant\'s current three year rolling average collective dose (142.6 person-rem/reactor years for 2010 through 2012) is less than the criteria of 240 person-rem per boiling water reactor unit. The inspectors determined that this finding has a cross-cutting aspect in the Human Performance area, Work Control component, because Entergy did not implement the planned work as intended, which involved job site activities, and impacted radiological safety. |
| 05000271/FIN-2013004-01 | 2013Q3 | Green | H.8 | NRC identified | Failure to Monitor the Unavailability of the B Control Rod Drive Equipment Train | The inspectors identified a NCV of Title 10 Code of Federal Regulations (10 CFR) 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, because Entergy did not monitor the performance of the B control rod drive (CRD) equipment train. Specifically, Entergy did not include seven days of unavailability for the B CRD flow control valve in the tracking database, and therefore did not initiate corrective actions when the train exceeded its unavailability criterion. Entergy initiated a condition report to document exceeding the performance criterion, entered the unavailability into the tracking database, and initiated a condition report to document the oversight in unavailability tracking. 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 of ensuring the availability of systems that respond to initiating events to prevent undesirable consequences. Specifically, since Entergy personnel did not recognize that this unavailability put the plant into a higher integrated risk category and did not recognize the plant risk impact of the flow control valves extended unavailability, no corrective actions were taken to address the maintenance practices which caused the unavailability performance criterion to be exceeded unnecessarily. 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 represent a loss of system safety function or a loss of safety function of a single train for greater than its Technical Specification allowed outage time. In addition, the failure to recognize and manage the plant risk associated with the 169 hours of unavailability of the B CRD flow control valve resulted in an incremental core damage probability of approximately 2E-10, which is less than 1E-6, and therefore also of very low safety significance. The inspectors determined that this finding has a cross-cutting aspect in the Human Performance area, Work Practices component, because Entergy personnel did not follow the maintenance rule program procedures. Specifically, operations did not log the unavailability in the maintenance rule out-of-service log and the system engineer did not review the scoping document to verify which components counted toward the train unavailability. |
| 05000271/FIN-2013003-01 | 2013Q2 | Green | P.1 | Self-revealing | Inadequate Corrective Actions for Sealing Flood Pathways into the Electrical Switchgear Rooms | A self-revealing NCV of Title 10 Code of Federal Regulation (10 CFR) 50, Appendix B, Criterion XVI, Corrective Action, was identified because Entergy did not promptly identify and correct two separate conditions adverse to quality related to flood protection of the switchgear rooms. Specifically, mechanical screw-type flood seals were not promptly replaced with reliable foam seals and an open drain line was not promptly identified and corrected allowing for water intrusion pathways into the switchgear rooms. Entergys corrective action to restore compliance consisted of placing the issue into the corrective action program and sealing all the potential pathways with Sylguard by April 8, 2013. The inspectors determined that the failure to identify the flood pathways was a performance deficiency that was within Entergys ability to foresee and correct and should have been prevented. This finding is more than minor because it is associated with the protection against external events attribute of the Mitigating Systems cornerstone and affected the objective to ensure the availability and reliability of systems that respond to external events to prevent undesirable consequences. Specifically, the failed flood barriers provided an external flood water pathway that could potentially impact the reliability and availability of both electrical switchgear rooms during a design basis flood event. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 4 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2012, the inspectors determined that this finding was of very low safety significance (Green) because, in spite of the failed flood barriers, sufficient water removal capability was available to ensure there was no loss of electrical switchgear safety function. The switchgear would still have been able to perform its function because the water level would have been maintained below floor level using the additional sump pump capacity available on site. The inspectors determined that the finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, because Entergy did not identify these issues in a timely manner commensurate with their safety significance. Specifically, Entergy had opportunities as part of the extent of condition from a 2012 flood seal failure and as part of the Fukushima flooding walkdowns to identify the additional possible flood paths and did not. |
| 05000271/FIN-2013003-02 | 2013Q2 | Green | H.4 H.5 | NRC identified | Failure to Maintain Compensatory Measures Required for a Barrier Breach Permit for the A Emergency Diesel Generator Room | The inspectors identified a NCV of Technical Specification (TS) 6.4, Procedures, because Entergy did not implement a barrier breach permit required by procedure. Specifically, Entergy personnel created three open penetrations to the A emergency diesel generator (EDG) room when the barrier breach permit for the planned modification allowed only one hole at a time. Entergys corrective action to restore compliance consisted of placing the issue into their corrective action program and installing the tubing and grouting the three holes. The inspectors determined that Entergys failure to properly implement a barrier breach permit by opening three holes in the A EDG room west wall instead of only one was a performance deficiency that was reasonably within Entergys ability to foresee and correct and should have been prevented. This finding is more than minor because it is associated with the design control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the capability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Specifically, the capability of the A EDG room west wall to limit the intrusion of a high energy line break into the A EDG room was reduced more than allowed and accepted by the barrier breach permit, and the equipment within the A EDG room was only qualified for a mild environment (i.e. not a steam or high temperature environment). Additionally, the finding is similar to IMC 0612, Appendix E, Examples of Minor Issues, examples 3.i. and 3.j., more than minor descriptions, because the accident analysis calculation had to be re-performed to assure the accident analysis requirements were met and there was reasonable doubt on the operability of the equipment without the re-analysis. In accordance with IMC 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012, the inspectors determined that this finding is of very low safety significance (Green) because the finding is a deficiency affecting the qualification of a mitigating structure but the structure maintained its functionality. Specifically, the A EDG room west wall would have sufficiently limited the intrusion of steam from a design basis high energy line break (HELB) to maintain the operability of equipment within the A EDG room. The inspectors determined that the finding has a cross-cutting aspect in the Human Performance area, Work Control component, because Entergy personnel 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 barrier breach permit (i.e., drilling only one hole at a time), but the necessary actions were not sufficiently communicated to maintenance or operations personnel (H.3(b)) |
| 05000271/FIN-2013003-03 | 2013Q2 | Green | H.7 | Self-revealing | Inadequate Procedure for Configuration Control Results in a Dislodged Secondary Containment Blowout Panel Due to Reactor Building Ventilation System Pressurization | A self-revealing NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified because Entergy did not establish a procedure controlling plant equipment appropriate to the circumstances for activities affecting quality. Specifically, the procedure for the control of plant equipment did not require identifying and tagging the switches of deenergized equipment that continued to have control power. As a result, the reactor building ventilation was operated in a manner that dislodged an engineered blowout panel rendering secondary containment inoperable. Entergys corrective action to restore compliance consisted of entering the issue into their corrective action program and implementing a night order to place a caution tag on the control switches of components that are deenergized and continue to have control power available. The inspectors determined that Entergys inadequate procedure for the control of plant equipment such that deenergized loads that continued to have control power were not identified or tagged was a performance deficiency that was reasonably within Entergys ability to foresee and correct and should have been prevented. This finding is 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 (containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the lack of identification and/or tagging of the switches for A reactor building ventilation resulted in dislodging a secondary containment blowout panel that rendered secondary containment inoperable. In accordance with IMC 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix G, Shutdown Operations Significance Determination Process, issued February 28, 2005, the inspectors determined that this finding is of very low safety significance (Green) because it did not increase the likelihood of a loss of reactor coolant system inventory, did not degrade the ability to terminate a leak path or add reactor coolant system inventory, and did not degrade the ability to recover decay heat removal if it was lost. Specifically, at the time secondary containment was rendered inoperable, the reactor coolant system was fully flooded, the event did not increase the likelihood of any initiating event, and secondary containment was not required to be operable at the time given no fuel movement, core alterations, or operations with a potential for draining the reactor vessel in progress. The inspectors determined that the finding had a cross-cutting aspect in the Human Performance area, Resources component, because Entergys procedures did not ensure that components in the field were labeled sufficiently and completely, during the bus deenergization, to assure nuclear safety (H.2(c)). |
| 05000271/FIN-2013008-01 | 2013Q2 | Green | NRC identified | Improper Maintenance Rule Scoping of the Reactor Building HVAC System | The inspectors identified a NCV of Title 10 Code of Federal Regulations (10 CFR) 50.65(b)(2) because Entergy did not properly scope the reactor building heating, ventilation and air conditioning (HVAC) system within the stations maintenance rule program. Specifically, the inspectors determined Entergy did not properly scope the reactor building HVAC system, specific to the systems function to run and assist in area temperature control, into the maintenance rule program as required. The system is directly used in the emergency operating procedure (EOP)-4, Secondary Containment Control, to assist in mitigating a high temperature condition. The inspectors determined that this finding was more than minor because it is associated with the equipment performance 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. Specifically, reliably starting reactor building HVAC system could mitigate or lessen the severity of a high temperature condition in the reactor building during an event or system which requires EOP-4 entry. The performance deficiency was also determined to be similar to more than minor example 7.d per IMC 0612, Appendix E, Examples of Minor Issues. The inspectors completed a Phase 1 screening of the finding per IMC 0609, Attachment 4, Phase 1 Initial Screening and Characterization of Findings, and determined the finding to be 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 allowed outage time, and did not screen as potentially risk-significant due to a seismic, flooding, or severe weather initiating event. The inspectors did not identify a cross-cutting aspect associated with the finding because the underlying performance aspects occurred in the late 1990s and no recent operating experience was identified that would reasonably have prompted Entergy to review their scoping adequacy. | |
| 05000271/FIN-2013202-01 | 2013Q2 | Green | H.13 | NRC identified | Security | |
| 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. |
| 05000271/FIN-2013002-04 | 2013Q1 | Green | H.14 | Self-revealing | Failure of the B Emergency Diesel Generator from Jacket Water Leakage Due to Inadequate Corrective Action | A self-revealing 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 EDG. 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 a condition report. 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 EDG 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 inspectors and a Region I Senior Reactor Analyst (SRA) completed the Detailed Risk Evaluation (DRE) in accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings At- Power, and determined this finding to be of very low safety significance (Green). The DRE estimated the increase in core damage frequency (ACDF) for internal initiating events in the range of 1 core damage accident in 2,000,000 years of reactor operation, in the mid-E-7 range per year. In addition, external initiating events such as fire, seismic and flooding would not have increased the total ACDF above 1 E-6 per year, and the increase in the frequency of a large early release of radioactive material (ALERF) associated with the internal event ACDF core damage sequences would not be above 1E-7 per year. 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-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-2013002-03 | 2013Q1 | Green | P.5 | NRC identified | Inadequate Corrective Action for Maintaining Operability of the LOW Pressure Coolant Injection Battery UPS-1A | The inspectors identified an NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, because Entergy did not promptly correct an adverse condition resulting in the failure of the B-UPS-1A low pressure coolant injection (LPCI) uninterruptible power supply (UPS) battery. Specifically, Entergy personnel did not promptly replace a degraded battery cell prior to its exceeding operability limits. Entergys corrective actions included replacing cell 61, replacing all cells with individual cell voltages (ICVs) less than 2.13 V, expediting complete battery bank replacements with a due date of May 30, and initiating a condition report. 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 capability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Specifically, UPS-1A had unplanned inoperability and degraded capacity due to cell 61 being out of service which commenced at some unknown point between December 3 and December 9 and was restored when cell 61 was replaced on December 10. In accordance with IMC 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, the inspectors determined that this finding is of very low safety significance (Green) because it did not represent a loss of system safety function or a loss of safety function for a single train (UPS-1A and A LPCI) for greater than its technical specification allowed outage time (seven days). The inspectors determined that the finding had a crosscutting aspect in the Problem Identification and Resolution area, Operating Experience component, because Entergy personnel did not implement and institutionalize available operating experience guidance contained within IEEE-450, IEEE Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead-Acid Batteries for Stationary Applications, or alternatively, vendor recommendations, to support plant safety. |
| 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-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-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-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-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-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-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-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-2011005-01 | 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-2011404-01 | 2011Q2 | Green | H.12 | Self-revealing | 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-2010403-01 | 2011Q1 | Green | P.2 | NRC identified | Security | |
| 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-2011403-01 | 2011Q1 | Green | H.9 | NRC identified | Security | |
| 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-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-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-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)). |