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| Quarter | Site | Title | Description | |
|---|---|---|---|---|
| 05000334/FIN-2018011-01 | 2018Q3 | Beaver Valley | Duties of the Shift Technical Advisor for Control Room Evacuation during a Fire Event. | The inspectors identified a Green non-cited violation (NCV) of Technical Specification (TS) 5.4.1(a), Procedures, related to the duties of the Shift Technical Advisor (STA) in response to a serious fire requiring control room evacuation. Specifically, procedure 1OM-56C.4.E, Shift Technical Advisors Procedure, Revision 23, directs the STA to perform substantial plant equipment operations outside of the control room (i.e., opening breakers, operating valves, electrical switching, etc.). These duties preclude the STA from maintaining sufficient independence to provide advisory technical support to the Unit 1 and 2 Operating Shift Crews as required by NOP-OP-1002 Conduct of Operations, Revision 12, and Unit 1 TS 5.2.2.f. |
| 05000293/FIN-2018002-06 | 2018Q2 | Pilgrim | Minor Violation | This violation of minor significance was identified by the licensee and has been entered into the licensees corrective action program and is being treated as a minor violation, consistent with the NRC Enforcement Policy. On June 22, 2015, Entergy submitted a licensee event report in accordance with 10 CFR 50.73 that contained information that was not complete or accurate in all material respects, contrary to the requirements in 10 CFR 50.9. Specifically, the licensee submitted Licensee Event Report 2015-004-00 to communicate the failure during testing of time delay Agastat relay 27A-B1X/TDDO intended to provide under-voltage protection for 480V emergency bus B6 by transferring power from bus B1 to bus B2. In the licensee event report, Entergy incorrectly documented that due to the failure, bus B6 would have continued to receive power from bus B1 with degraded voltage. Upon identifying the issue, on March 8, 2016, Entergy submitted a revised licensee event report with the correct information. Enforcement: 10 CFR 50.9 requires that information provided to the Commission by a licensee shall be complete and accurate in all material respects. Contrary to the above, on June 22, 2015, Entergy provided information to the Commission that was not complete and accurate in all material respects. In the licensee event report, the licensee documented that due to the failure, bus B6 would have continued to receive power from bus B1 with degraded voltage. However, bus B6 would actually have tripped from bus B1 and lost power completely. This information was material to the NRC because the NRC requires timely and accurate reporting of information related to events in order to evaluate the potential safety significance and required NRC response. Entergy identified the inaccuracy and entered the issue into its corrective action program (CR-PNP-2015-9762). On March 8, 2016, Entergy submitted a revision to the licensee event report (2015-004-01) that corrected the report. This failure to comply with 10 CFR 50.9 constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy. The disposition of this violation closes Licensee Event Report 05000293/2015-004-01. |
| 05000293/FIN-2018002-05 | 2018Q2 | Pilgrim | Licensee-Identified Violation | This violation of very low safety significance was identified by the licensee and has been entered into the licensees corrective action program and is being treated as a NCV, consistent with Section 2.3.2 of the Enforcement Policy. Violation: 10 CFR 50.72(b)(3)(v)(C) requires licensees to a notify the NRC within 8 hours any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material. Contrary to the above, Entergy did not make a required notification pursuant to 10 CFR 50.72(b)(3)(v)(C). Specifically, on June 20, 2017, secondary containment was declared inoperable due to simultaneous opening of both airlock doors, and Entergy did not make the required notification until June 22, 2017. Significance/Severity: This violation is being treated under the NRCs traditional enforcement process, for impeding the regulatory process, specifically Entergy did not make a required notification, as outlined in Inspection Manual Chapter 0612, Appendix B. The Reactor Oversight Processs significance determination process does not specifically consider the regulatory process impact in its assessment of licensee performance. Therefore, it is necessary to address this violation which impedes the NRCs ability to regulate using traditional enforcement to adequately deter non-compliance. The severity of this violation was determined to be Severity Level IV, as outlined in Example 9 from Section 6.9.d. of the NRC Enforcement Policy. Corrective Action References: CR-PNP-2017-06380 and CR-PNP-2017-07015 The disposition of this finding closes Licensee Event Report 2017-011-00. |
| 05000293/FIN-2018002-04 | 2018Q2 | Pilgrim | Licensee-Identified Violation | This violation of very low safety significance was identified by the licensee and has been entered into the licensees corrective action program and is being treated as a NCV, consistent with Section 2.3.2 of the Enforcement Policy. Violation: 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality shall be prescribed by documented instructions appropriate to the circumstances and shall be accomplished in accordance with the instructions. Contrary to the above, from January 1994 to June 2017, Entergy modified site surveillance procedure 8.M.3-18, Standby Gas Treatment System Exhaust Fan Logic Test and Instrument Calibration, without prescribing adequate documented instructions for the condition caused by the testing. Specifically, Entergy failed to identify that the procedurally prescribed lineup of the standby gas treatment system resulted in secondary containment being inoperable due to the large opening introduced into the system. Significance/Severity: The inspectors evaluated this finding using Inspection Manual Chapter 0609.04, Initial Characterization of Findings, and Inspection Manual Chapter 0609, Appendix A, Exhibit 3, Barrier Integrity Screening Questions. The inspectors determined that the finding was of very low safety significance. Corrective Action Reference: CR-PNP-2017-11714 The disposition of this violation closes Licensee Event Reports 05000293/2017-013-00 and 05000293/2017-013-01. |
| 05000286/FIN-2018002-01 | 2018Q2 | Indian Point | Reactor Pressure Boundary Leakage Due to Weld Failure in Reactor Vessel Head Penetration #3 | A self-revealing Severity Level IV NCV of Technical Specification (TS) 3.4.13.a, Reactor Coolant System Operational Leakage, was identified when Entergy operated the reactor in Mode 1 with pressure boundary leakage for a period of time longer than the allowable limiting condition of operation. Specifically, a leak in the J-weld around reactor pressure vessel (RPV) head penetration #3 occurred during the last operating cycle and was not identified until after the reactor was shutdown for a refueling outage. |
| 05000293/FIN-2018002-03 | 2018Q2 | Pilgrim | 480V Bus B6 Auto Transfer Function Degraded Due to Time Delay Relay Failure | The inspectors identified a Severity Level IV NCV of TS 3.5.A.2 because a component of the low pressure coolant injection system was inoperable between May 12, 2015, and May 3, 2017, during which time, on occasions, core spray systems were also not operable. Specifically, a relay, used to transfer the power feed for the low pressure coolant injection valves to the backup source in the event of a degraded voltage condition, failed during testing. As a result, under certain conditions, the transfer would not have automatically occurred. This condition existed through the operating cycle, during which time the core spray pumps were also inoperable when removed from service for scheduled maintenance. |
| 05000293/FIN-2018002-02 | 2018Q2 | Pilgrim | Loss of Secondary Containment Integrity due to Simultaneously Opened Airlock Doors | A self-revealed Green finding was identified when personnel did not implement a procedure requiring the closure and verification of doors credited with specific design functions. Procedure 1.3.135, Control of Doors, requires station personnel to ensure closing and latching of doors. Failure to meet this requirement caused the loss of secondary containment integrity and unplanned entry into Technical Specification (TS) condition 3.7.C.1. |
| 05000293/FIN-2018002-01 | 2018Q2 | Pilgrim | Failure to Properly Implement the Fatigue Management Program Work Hour Controls for Covered Workers | The inspectors identified a Green NCV of Title 10 of the Code of Federal Regulations (10 CFR) 26.205(d). During the period December 2017 to April 2018, Entergy did not properly control the work hours of several workers who performed work covered under 10 CFR 26.4(a). Specifically, on eleven occasions, workers exceeded one of the following work hour limits: (1) 16 work hours in any 24-hour period; (2) 72 hours in any 7-day period; or (3) 54 hours per week average over a 6-week rolling time period. |
| 05000387/FIN-2018010-01 | 2018Q1 | Susquehanna | Licensee-Identified Violation | This violation of very low safety significant was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a non-cited violation, consistent with Section2.3.2 of the Enforcement Policy.Violation: 10 CFR 50.49(e)(5) requires, in part, that the electrical equipment qualification program must replace or refurbish the equipment at the end of its designated life.Contrary to the above, on November 16, 2017, the licensee identified that thirteen Unit 1, NAMCO limit switches in environmentally qualified (EQ) applications inside primary containment were not installed in their fully qualified configuration. Specifically, contrary to vendor instructions and EQAR-004 requirements, the limit switches for several containment isolation valves (CIV) have had their covers removed and reinstalled without replacing the gasket and cover screw O-rings. For this application, opening and/or removing the limit switch gasket /and cover screws O-ring constituted the end of the gasket/O-ring designated life. Significance/Severity Level: The inspectors evaluated this finding using IMC 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix A, Exhibit 3, Barrier Integrity Screening Questions. The inspectors determined that the finding was of very low safety significance (Green), because the limit switches provide only an open or closed signal indication in the main control room, so that operators are aware of the valve position, and can make appropriate assessment of plant conditions. The safety function of the containment isolation valves was not affected. |
| 05000247/FIN-2017007-02 | 2017Q4 | Indian Point | Failure to Maintain B.5.b Mitigating Strategies | An NRC-identified finding of very low safety significance (Green) and NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50.54(hh)(2), Conditions of Licenses, the Unit 2 FOL Condition 2.N, and the Unit 3 FOL Condition 2.AC was identified for failure to maintain strategies for addressing large fires and explosions. Specifically, Entergy failed to maintain the B.5.b strategies when the sites diesel contingency pump (B.5.b pump), B5B-101-PMP, was declared non-functional and unavailable on March 20, 2017, due to a deficiency associated with the pumps engine and failed to promptly restore the pump to a functional status or establish any 3 compensatory measures. Entergy entered this issue into its CAP and promptly completed repairs to the B.5.b pump. This finding was more than minor because it is associated with the Protection Against External Factors (e.g., fire) attribute of t he Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent core damage. The team evaluated the significance of the finding in accordance with IMC 0609, Appendix L, B.5.b Significance Determination Process. The finding was determined to be of very low safety significance (Green) because although the B.5.b pump was considered unavailable, the team concluded that the pump was recoverable. This finding had a cross-cutting aspect of Resources, in the area of Human Performance, because leaders did not ensure that personnel, equipment, procedures, and other resources were available and adequate to support nuclear safety. Specifically, procedural guidance and equipment were not available to operators to implement adequate compensatory measures when the B.5.b pump became non-functional and unavailable. (H.1) |
| 05000247/FIN-2017007-03 | 2017Q4 | Indian Point | Licensee-Identified Violation | The following violation of very low safety significance (Green) was identified by Entergy and is a violation of NRC requirements which meets the criteria of the NRC Enforcement Policy for being dispositioned as an NCV. Indian Point Unit 2 Technical Specification 5.4.1.k and Indian Point Unit 3 Technical Specification 5.4.1.d require written procedures shall be established, implemented, and maintained covering FPP implementation. Procedure 0-PT-M004, Fire Extinguisher Inspection, Revision 9 implements monthly inspections of portable fire extinguishers to verify hydrostatic testing and periodic maintenance was performed within the periodicity specified by NFPA 10-1990, Standard for Portable Fire 17 Extinguishers. Procedure 0-PT-M004 requires portable fire extinguishers to be removed from service and replaced if their periodic maintenance or hydrostatic testing are not current within the specified periodicity. Contrary to the above, from July 1, 2015, to August 25, 2017, approximately 200 portable fire extinguishers were not removed from service and replaced when they exceeded their specified periodicity for maintenance and/or hydrostatic testing. Specifically, plant staff did not properly implement procedure 0-PT-M004 to verify portable fire extinguishers were periodically maintained and hydrostatically tested at intervals specified by NFPA 10-1990. The application software used in conjunction with 0-PT-M004 to perform monthly fire extinguisher inspections was revised in June 2015 and plant staff did not set up the software to require verification that fire extinguisher maintenance and hydro tests were current. Fire protection engineers identified the issue in August 2017, evaluated the subject fire extinguishers, and determined they were acceptable for continued use until December 31, 2017, based on the relatively short period of untimely maintenance/testing, satisfactory monthly verifications of physical condition, and the availability of additional portable fire extinguishers in the affected fire areas. Additional corrective actions included initiation of an accelerated maintenance and hydro test program to ensure all portable fire extinguishers met NFPA-10 maintenance and test requirements by December 21, 2017, and revision of 0-PT-M004 and the associated application software. Entergy entered the issue into the CAP (CRs IP3-2017-04084 and IP3- 2017-02945). The inspectors evaluated this finding using IMC 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions. The inspectors determined that the finding was of very low safety significance (Green) because the finding did not impact the frequency of a fire and did not involve a loss or degradation of equipment or function specifically designed to mitigate an external event. |
| 05000286/FIN-2017007-01 | 2017Q4 | Indian Point | Inadequate Alternative Post-Fire Safe Shutdown Procedure | An NRC-identified finding of very low safety significance (Green), involving an NCV of Indian Point Unit 3 Facility Operating License (FOL) Condition 2.H, was identified because Entergy did not implement and maintain in effect all provisions of the Fire Protection Program (FPP), as approved by the NRC. Specifically, Entergy did not have an adequate post-fire operating procedure for its alternative shutdown capability to ensure that safe shutdown (SSD) equipment analyzed to be available during the postulated fire in fire area ETN-4(2), Upper Electrical Tunnel, were credited in the procedure. Entergy entered this issue into its corrective action program (CAP) and promptly implemented compensatory measures by establishing a fire watch. This finding was more than minor because it was associated with the Protection Against External Factors (e.g., fire) attribute of t he Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the reliability and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). The team performed a Phase 2 Significance Determination Process screening for this issue, in accordance with NRC IMC 0609, Appendix F, Fire Protection Significance Determination Process. This finding affected the post-fire SSD category because the implementing procedures were adversely affected. The team determined that this finding screened to very low safety significance (Green) based upon task 2.3.5, because no credible fire ignition source scenarios were identified in fire area ETN-4(2) that could affect both electrical channels I and II cables. This finding did not have a cross-cutting aspects because it was a legacy issue and was considered to not be indicative of current licensee performance. |
| 05000278/FIN-2017008-02 | 2017Q1 | Peach Bottom | Untimely Corrective Actions to Address Elevated Primary Containment Isolation Valve Leakage | Green. The inspectors identified a self-revealing non-cited violation of 10 CFR 50 Appendix B, Criterion XVI, Corrective Action, because Exelon did not promptly implement corrective actions to address a condition adverse to quality on two containment isolation valves. Specifically, drywell air sampling valves SV-3-7D-3671A and SV-3-7D-3671D failed to perform their primary containment isolation function on March 15 and September 26, 2016, respectively, as a result of untimely corrective actions to address elevated leakage. The valve internals were repaired, declared operable, and the issue was entered into the corrective action program (IR 3990490). The finding was more than minor, because it was associated with the barrier performance attribute of the Barrier Integrity cornerstone and adversely affected the cornerstones objective to provide reasonable assurance that the containment design barrier protect the public from radionuclide releases caused by accidents or events. In accordance with IMC 0609.04, Initial Characterization of Findings, dated October 7, 2016, and Exhibit 1 of IMC 0609, Appendix A, The SDP for Findings At-Power, dated June 19, 2012, the inspectors determined this finding was of very low safety significance, because the finding did not result in an actual open pathway in the physical integrity of the reactor containment or involve an actual reduction in the function of hydrogen igniters in the reactor containment. The inspectors determined this finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Resolution, because Exelon did not perform effective corrective actions in a timely manner commensurate with the safety significance of the issue. Specifically, corrective actions to address a CAQ on SV-3-7D-3671A and SV-3-7D-3671D were delayed which resulted in the valves failing their LLRT and being declared inoperable. (P.3) |
| 05000277/FIN-2017008-01 | 2017Q1 | Peach Bottom | Untimely Corrective Actions to Address 2C Core Spray Motor Elevated Vibrations | Green. The inspectors identified a non-cited violation of 10 CFR 50 Appendix B, Criterion XVI, Corrective Action, because Exelon did not implement corrective actions in a timely manner to correct a condition adverse to quality on the 2C core spray motor. Specifically, Exelon did not perform appropriate corrective actions to evaluate and address an increasing motor bearing vibration trend that had existed for over ten years. Consequently, motor vibration reached the fault level established in Exelons vibration analysis procedure. The finding was more than minor, because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone and adversely impacted the cornerstone objective to ensure the reliability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). In accordance with IMC 0609.04, Initial Characterization of Findings, dated October 7, 2016, and Exhibit 1 of IMC 0609, Appendix A, The SDP for Findings At-Power, dated June 19, 2012, the inspectors determined this finding was of very low safety significance because the performance deficiency did not impact the design or qualification of the component, did not result in a loss of system function, did not result in the loss of function of a train greater than its Tech Spec allowed outage time, and did not represent an actual loss of function for a high safety significant component in accordance with Exelons maintenance rule program. The inspectors determined the finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Resolution, because Exelon did not take effective corrective actions in a timely manner commensurate with the safety significance of the issue. Specifically, corrective actions to address the elevated vibrations on the 2C core spray motor were not implemented before motor vibration reached the fault level and adversely impacted the long-term reliability of the motor. (P.3) |
| 05000387/FIN-2016007-01 | 2016Q4 | Susquehanna | Failure to Specify and Maintain Safety-Related Quality Standards and Materials Essential for Reactor Core Isolation Cooling | The team identified a Green non-cited violation of Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion III, Design Control, for the failure to classify and maintain reactor core isolation cooling (RCIC) system components as safety-related as specified by Updated Final Safety Analysis Report Table 3.2-1 and Section 7.1.1. Specifically, although Talen, the operator of Susquehanna Steam Electric Station, classified the RCIC system as safety-related, this classification did not extend to the Unit 1 and Unit 2 RCIC barometric condenser relief valves. The team determined failure of the non-safety related barometric condenser relief valves could result in a loss of RCIC lube oil cooling and failure of RCIC to perform its design basis safety function. Talen entered the issue into the corrective action program as condition report 2016-23615 and performed an immediate operability determination, which concluded RCIC remained operable. The finding was more than minor because it was associated with the Design Control attribute of the Mitigating Systems cornerstone, and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team evaluated this finding using IMC 0609, Attachment 4, Initial Characterization of Findings, and IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 2 - Mitigating System Screening Questions. The team determined the finding screened as very low safety significance (Green), because the finding was a design deficiency which did not result in an actual loss of functionality of the RCIC system. This finding was not assigned a cross-cutting aspect because the performance deficiency occurred during original plant design and did not reflect current licensee performance. |
| 05000387/FIN-2016007-02 | 2016Q4 | Susquehanna | Licensee-Identified Violation | Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Appendix B, Criterion XVI, Corrective Actions, requires measures to be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. Contrary to 10 CFR 50, Appendix B, Criterion XVI, station personnel did not promptly correct a condition adverse to quality. Specifically, from August 1, 2013 to February 26, 2016, Talen did not implement corrective actions to establish a PM program for molded case circuit breakers (MCCBs) found in distribution control panels that protect containment penetration conductors. The need for a MCCB PM program was originally identified during the 2013 NRC CDBI and documented as NCV 05000387 (05000388)/2013010-01, Failure to Verify Operation of Safety-Related 125Vdc Molded Case Circuit Breakers (CR 1732454). Talen identified this untimely implementation of corrective action during a self-assessment in preparation for the 2016 NRC CDBI. Plant staff entered the issue into the corrective action program (CRs 2016-04833; 23373; 23971 and 24015) and established a MCCB PM program. The team evaluated this finding using IMC 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix A, Exhibit 3, Barrier Integrity Screening Questions. The team determined that the finding was of very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containment (valves, airlocks, etc.), containment isolation system (logic and instrumentation), and heat removal components, and did not involve an actual reduction in function of hydrogen igniters in the reactor containment. |
| 05000289/FIN-2016004-01 | 2016Q4 | Three Mile Island | Licensee-Identified Violation | Technical specification 3.2.12.1, "LTOP Protection", requires when the reactor vessel head is installed and indicated reactor coolant system temperature is 313F, high pressure injection pump breakers shall not be racked in unless injection valves (MU-V16A/B/C/D and MU-V217) are closed with their associated breakers open and that pressurizer level is maintained 100 inches, or restore pressurizer level to 100 inches within 1 hour. Contrary to technical specification 3.2.12.1, during reactor coolant system filling with the vessel head installed and temperature < 313F, high pressure injection pump breakers were racked in while pressurizer level was >100 inches for greater than 1 hour. The condition existed for 2 hours and 49 minutes until recognized by the operating crew when questioned by a senior reactor operator trainee, at which time the crew took immediate actions to reduce pressurizer level <100 inches within 1 hour. Additional corrective actions included crew remediation, additional main control room supervisory oversight, and procedure changes. Exelon entered this issue into the corrective action program as issue report 3949713. The inspectors determined that the finding was of very low safety significance (Green) in accordance with NRC IMC 0609, Appendix G, Shutdown Operations, Attachment 1, Exhibit 4, since the finding did not represent an inadvertent safety injection and did not render the power-operated relief valve (LTOP Protection) unavailable or degraded. |
| 05000333/FIN-2016004-01 | 2016Q4 | FitzPatrick | Failure to Ensure Proper Configuration Control of a PCIV During Planned Maintenance | Green. The inspectors identified a Green NCV of Technical Specification (TS) 5.4, Procedures, because Entergy staff did not implement procedure AP-12.06, Equipment Status Control, as required. Specifically, Entergy personnel did not recognize the impact of a change associated with the tagout of a C residual heat removal (RHR) system primary containment isolation valve (PCIV). This resulted in motor operated valve 10MOV-13C being electrically isolated in the open position without being recognized as a PCIV and without proper entry into TS 3.6.1.3. Entergy restored the valve to operable status, entered this issue into their corrective action program (CAP) as condition report (CR)-JAF-2016-4419, and conducted meetings with each operating crew to discuss the event and reinforce standards for equipment status control and maintaining a questioning attitude. Training was also provided to operators to review the scenario and discuss requirements associated with PCIVs. This finding is more than minor because it was associated with the configuration control attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers (containment) protect the public from radionuclide releases caused by accidents or events. Specifically, Entergy staff did not recognize the impact of a change associated with the tagout of a containment isolation valve. The change in the tagout resulted in a failure to isolate the containment isolation valve and enter TS 3.6.3.1 prior to maintenance. The finding was similar to Example 3.j in Appendix E of IMC 0612, Examples of Minor Issues, issued August 11, 2009. Since the PCIV was in an open position with power removed, a reasonable doubt of operability existed due to the valves inability to close to perform its containment isolation function. The inspectors evaluated this finding using IMC 0609.04, Initial Characterization of Findings, issued October 7, 2016; Exhibit 3 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2012; and Appendix H of IMC 0609, Containment Integrity Significance Determination Process, issued May 6, 2004. Using Exhibit 3 of IMC 0609, Appendix A, Section B, Reactor Containment, the finding directed the use of IMC 0609, Appendix H because it represented an actual open pathway in the physical integrity of reactor containment (i.e. valve). Using IMC 0609, Appendix H, the finding was classified as a Type B finding because it was related to a degraded condition that had potentially important implications for the integrity of containment, without affecting the likelihood of core damage (i.e. containment isolation was precluded by the isolation valve being failed in the open position, however the low pressure coolant injection function remained 4 available). Using Table 6.1, Phase 1 Screening-Type B Findings at Full Power, for a boiling water reactor, Mark 1 Containment, the inspectors were directed to perform a Phase 2 Assessment because the structure, system, and component (SSC) affected by the finding was a containment isolation valve. Using Table 6.2, Phase 2 Risk Significance-Type B Findings at Full Power, the inspectors determined that the failure of the containment isolation valve critical to suppression pool integrity/scrubbing was less than 3 days, and therefore was of very low safety significance (Green). This finding has a cross-cutting aspect in the area of Human Performance, Challenge the Unknown, because Entergy failed to maintain a questioning attitude to identify an improper configuration associated with a PCIV tagout during maintenance planning and execution. Specifically, a tagout writer modified the configuration for a containment isolation valve, which was not challenged or questioned during subsequent reviews. This resulted in the PCIV being tagged out in the open position, a condition that rendered the valve inoperable. (H.11) |
| 05000333/FIN-2016007-02 | 2016Q2 | FitzPatrick | Failure to adequately evaluate a procedure change impacting a PRA-credited time critical operator action | The team identified a Green finding involving Entergys inability to complete a time critical operator action within the assumed probabilistic risk assessment (PRA) credited accident mitigation time limit to prevent undesirable consequences (i.e., core damage) under a postulated scenario (i.e., using the residual heat removal service water (RHRSW) system as an alternate injection source into the reactor pressure vessel (RPV) via the residual heat removal (RHR) system during a loss of coolant accident (LOCA)). Specifically, in response to a known degraded condition impacting an RHRSW valve, Entergy did not adequately evaluate an associated temporary procedure change to EP-8, Alternate Injection Systems, to ensure operator actions could be accomplished to initiate RHRSW injection to the RPV within the PRA-credited time. Entergy entered the issue into their CAP as CR 2016-1396 and CR 2016-1429 and completed corrective actions to pre-stage a ladder for operator use and provide additional guidance to plant operators. The finding was more than minor because it was associated with the design control (plant modifications) attribute of the Mitigating Systems cornerstone and adversely affected the cornerstones objective of ensuring reliability, availability, and capability of systems and operators that respond to initiating events to prevent undesirable consequences (i.e., core damage). The team evaluated the finding in accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings at Power, Exhibit 2 Mitigating Systems Screening Questions, and concluded it required a detailed risk evaluation (DRE). A Region I Senior Reactor Analyst performed the DRE and concluded that the failure of an operator action to align RHRSW for RPV alternate injection within the assumed PRA accident mitigation time limit results in an estimated increase in core damage frequency in the mid E-8/year range, or very low safety significance (Green). The finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, because Entergy did not thoroughly evaluate issues to ensure that resolutions address causes and extent-of-conditions commensurate with their safety significance. Specifically, Entergy did not thoroughly evaluate the effect of an alternate injection procedure change on PRA-credited time critical operator actions. (PI.2) |
| 05000333/FIN-2016007-01 | 2016Q2 | FitzPatrick | Failure to ensure design basis of EDG LO storage facility | The team identified a Green NCV of Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Appendix B, Criterion III, Design Control, because Entergy did not ensure that FitzPatricks emergency diesel generator (EDG) lubrication oil (LO) supply storage facility was designed to withstand the effects of natural phenomena. Specifically, additional LO, maintained and inventoried monthly to ensure an adequate LO supply to meet the EDGs seven day mission time, was stored in a non-Class I structure that was not designed to withstand the effects of natural phenomena induced plant events that the EDGs were designed to mitigate. Entergy entered the issue into the corrective action program (CAP) as condition report (CR) 2016-1471 and promptly relocated the LO reserve inventory from warehouse No. 2 to the EDG building, which is constructed to Class I seismic and tornado protection design criteria. The finding was more than minor because it was associated with the protection against external factors attribute of the Mitigating Systems cornerstone and adversely affected the cornerstones objective of ensuring reliability and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). The team evaluated the significance of this finding using IMC 0609, Appendix A, The Significance Determination Process for Findings at Power, Exhibit 2 Mitigating Systems Screening Questions. The team determined the finding screened as very low safety significance (Green) because the finding was a design deficiency which did not result in an actual loss of functionality of the EDGs. This finding did not have a cross-cutting aspect because the underlying cause occurred in 1988 during warehouse No. 2 construction and was not indicative of current Entergy performance. |
| 05000289/FIN-2016007-01 | 2016Q1 | Three Mile Island | Licensee-Identified Violation | Title 10 CFR 50.55a (g)(4), In-service Inspection Requirements, requires in part, that throughout the service life of a boiling or pressurized water-cooled nuclear power facility, components (including supports) that are classified as ASME Code Class 1, must meet the requirements, except design and access provisions, and preservice examination requirements set forth in Section XI of editions and addenda of the ASME Boiler Pressure and Vessel Code (BPVC) that become effective subsequent to editions specified in paragraphs (g)(2) and (g)(3) of this Section, and that are incorporated by reference in paragraph (b) of this Section, to the extent practical within the limitations of design, geometry, and materials of construction of the components. Section XI of the ASME BPVC, 2001 Edition with 2003 Addenda, Table IWF-2500-1, Examination Category F-A Supports, requires VT-3 examination of 100 percent of the ASME Class 1 supports, other than piping supports, every ISI Interval (examination item F1.40), as modified by Notes 1, 2, 3 and 5 of Table IWF-2500-1. Contrary to this requirement, from initial plant operation until November 14, 2015, (when Exelon staff completed the initial required VT-3 examination), Exelon failed to perform the required VT-3 examination of ASME Class 1 supports, other than piping supports, (i.e. seismic support plates and associated load path components) on the TMI control rod drive mechanism assemblies. Exelon staff entered the issue into their corrective action program as IR 01678190. The inspectors evaluated this finding using IMC 0609.04, Initial Characterization of Findings, IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, and IMC 0609, Appendix A, Exhibit 3, Barrier Integrity Screening Questions. The finding is more than minor because it is associated with the protection against external factors attribute of the mitigating systems cornerstone and adversely affects the objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined that the finding was of very low safety significance (Green) because the finding did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic initiating event and was not associated with pressurized thermal shock of the reactor coolant system boundary. |
| 05000390/FIN-2016001-01 | 2016Q1 | Watts Bar | Failure to Use a Procedure Appropriate to the Circumstances for the Auxiliary Control Air System Train A | A self-revealing non-cited violation (NCV) of 10 Code of Federal Regulations (CFR) 50, Appendix B, Criterion V, Procedures was identified for the licensees failure to use a procedure appropriate to the circumstances for work associated with the A-A auxiliary control air system (ACAS) compressor. Specifically, the licensee used a section of procedure 0-SOI-32.02, Auxiliary Air System, Revision 2, that placed the air compressor in OFF when it was intended to place it in A-Auto. The licensee restored the compressor to A-Auto and entered this issue into their corrective action program as condition report (CR) 1131261. The performance deficiency was more than minor because it affected the equipment performance attribute of the mitigating system cornerstone to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the ACAS train A was nonfunctional for approximately 19.5 hours on January 29, 2016 and as a supported system, the auxiliary feedwater system was inoperable during this time. The inspectors determined that this finding was of very low safety significance (Green) because the finding did not represent an actual loss of function of a single train for greater than its TS allowed outage time. The finding has a cross cutting aspect in the Work Management component of the Human Performance area because the licensee failed to implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. Specifically, the planning and execution of work on the A-A ACAS compressor on January 29, 2016 lacked sufficient rigor to ensure the activity was performed as intended. |
| 05000390/FIN-2016001-02 | 2016Q1 | Watts Bar | Inadequate Immediate Determination of Operability for the Auxiliary Control Air System Train A | The NRC identified an NCV of 10 CFR 50, Appendix B, Criterion V, Procedures, for the licensees failure to follow TVA procedure OPDP-8, Operability Determination Process and Limiting Conditions for Operation Tracking, Revision 21. Specifically, the licensee failed to base an immediate determination of operability (IDO) for the auxiliary control air system on information sufficient to conclude that a reasonable expectation of operability/functionality existed. The licensee subsequently implemented compensatory measures and entered this issue into their corrective action program as CR 1129322. The performance deficiency was more than minor because it affected the equipment performance attribute of the mitigating system cornerstone to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, reasonable assurance of operability/functionality did not exist for the A train of auxiliary control air from January 13, 2016, until January 14, 2016, and it therefore should have been declared inoperable/nonfunctional. The inspectors determined that this finding was of very low safety significance (Green) because the finding did not represent an actual loss of function of a single train for greater than its TS allowed outage time. This finding had a cross-cutting aspect in the area of Human Performance, conservative bias, because the licensee failed to make the conservative decisions. Specifically, the licensee reinstalled a degraded valve in the auxiliary control air system without fully understanding the failure mechanism or its impact on system operability/functionality. |
| 05000390/FIN-2016001-03 | 2016Q1 | Watts Bar | Failure to Adequately Implement the Administration of Site Technical Procedures for TDAFW Pump Governor Calibration | The NRC identified an NCV of TS 5.7.1.1.a, Procedures, for the licensees inadequate implementation of procedure NPG-SPP-01.2, Administration of Site Technical Procedures, Revision 8. Specifically, the licensee determined applicable acceptance criteria steps in technical procedures were not applicable (N/A) in lieu of performing a procedure change. This resulted in challenging the operability of safety-related plant equipment. The licensee entered this issue into their corrective action program as CR 1125256. The performance deficiency was more than minor because, if left uncorrected, it could lead to a more significant safety concern with the use of N/A and implementation of site technical procedures. Specifically, if further adjustments outside of the acceptance criteria or additional acceptance criteria were not met, it could have resulted in the turbine-driven auxiliary feedwater pump becoming inoperable. The inspectors determined this finding to be of very low safety significance (Green) because it was a deficiency affecting the design or qualification of equipment and operability was maintained. The finding had a cross-cutting aspect of Procedure Adherence, as described in the Human Performance cross-cutting area because the licensee failed to comply with NPG-SPP-01.2. |
| 05000390/FIN-2016001-04 | 2016Q1 | Watts Bar | Failure to Place the RHR System in ECCSStandby Mode Prior to Exceeding an RCS Temperature of 212 oF | The NRC identified an NCV of TS 5.7.1.1.a, Procedures, for the licensees failure to place the residual heat removal (RHR) system into ECCS-Standby Mode prior to the reactor coolant system (RCS) temperature exceeding 212 oF as required by procedure 1-GO-1, Unit Startup from Cold Shutdown to Hot Standby, Revision 4. The licensee entered this issue into their corrective action program as CR 1127691. The performance deficiency was determined to be more than minor because, if left uncorrected, a failure to align a safety system under the proper plant conditions could lead to that system being inoperable or degraded. The inspectors determined that this finding was of very low safety significance (Green) because the system temperatures never rose high enough to allow the RHR pump suction header to form steam voids. The performance deficiency had a cross-cutting aspect of Avoid Complacency in the area of Human Performance because licensee personnel were complacent and failed to question the long held idea that the particular step just needed to be started prior to exceeding an RCS temperature of 212 oF. |
| 05000390/FIN-2016001-05 | 2016Q1 | Watts Bar | Failure to Use Approved Procedures to Place RHR Letdown In Service | The NRC identified an NCV of TS 5.7.1.1.a, Procedures, for the licensees failure to use any approved procedures to place RHR Letdown in service. The licensee entered this issue into their corrective action program as CR 1127691. The performance deficiency was determined to be more than minor because if left uncorrected a failure to use procedures to place systems or portions of systems in service could result in equipment being operated incorrectly and that system could then become inoperable or degraded. The inspectors determined that this finding was of very low safety significance (Green) because the way that the system was placed in service did not cause any safety-related components to become inoperable nor did it represent an actual loss of function of one or more non-TS trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. The performance deficiency had a cross-cutting aspect of safety conscious work environment (SCWE) policy in the area of Safety Conscious Work Environment because the licensee organization failed to effectively implement a policy that supports individuals rights and responsibilities to raise safety concerns, and does not tolerate harassment, intimidation, retaliation, or discrimination for doing so. |
| 05000390/FIN-2016001-06 | 2016Q1 | Watts Bar | Failure to Track Applicable Technical Specification Action Statement for Charging Pump Inoperability | The NRC identified an NCV of TS 5.7.1.1.a, Procedures, for the licensees failure to implement OPDP-8, Operability Determinations and LCO tracking. Specifically, the licensee failed to track the applicability of action statement B of TS LCO 3.5.3, ECCS- Shutdown, during planned testing. The licensee entered this issue into their corrective action program as CR 1134949. The licensees failure to track applicable TS LCOs, as required by Section 3.5.1 of OPDP-8 was a performance deficiency. The performance deficiency was more than minor because, if left uncorrected, it would have had the potential to lead to a more significant safety concern in that, the failure to track an applicable TS action statement could lead to plant operations outside of TS analyzed conditions. The inspectors determined that this finding was of very low safety significance (Green) because the finding did not represent an actual loss of function of a single train for greater than its TS allowed outage time nor did it represent an actual loss of function of one or more non-TS equipment for greater than 24 hours. The performance deficiency had a cross-cutting aspect of Challenge the Unknown in the area of Human Performance because licensee personnel did not appropriately stop, question, and evaluate the risks before proceeding when the 1A-A CCP oil cooler low flow alarm came in during flow testing. |
| 05000390/FIN-2016001-10 | 2016Q1 | Watts Bar | Failure to Maintain an Adequate Surveillance Procedure for Emergency Core Cooling System Venting | The inspectors identified an apparent violation of TS 5.7.1.1.a, Procedures, for the licensees failure to maintain procedure 1-SI-63-10.1-A, ECCS Discharge Pipes Venting Train A Inside Containment, Revisions 11-16, in accordance with the requirements of Regulatory Guide 1.33. Specifically, the procedure did not have provisions for quantifying accumulated gases during venting which allowed emergency core cooling system (ECCS) piping to be vented without being evaluated for potential adverse impacts on system operability. The licensee implemented manual ultrasonic testing (UT) of gas accumulation and entered this issue into their corrective action program as CR 1136359. The performance deficiency was more than minor because, if left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, if left uncorrected, the potential existed for an unacceptable void affecting ECCS operability to develop prior to the next scheduled surveillance. The inspectors determined the finding could not be screened to GREEN and may require a detailed risk evaluation following a determination of whether the finding represents a loss of system and/or function. Because the safety characterization of this finding is not yet finalized, it is being documented with a significance of To Be Determined (TBD). The inspectors determined that the finding had a cross-cutting aspect of Change Management in the area of Human Performance because the licensee failed to use a systematic process to implement changes to the ECCS venting procedure to ensure that Generic Letter 2008-01 commitments would continue to be met. |
| 05000390/FIN-2016001-11 | 2016Q1 | Watts Bar | Licensee-Identified Violation | Watts Bar Operating License Condition 2.F requires that the licensee shall implement and maintain in effect all provisions of the approved fire protection program, as described in the Fire Protection Report for Watts Bar Unit 1, as approved in Supplements 18 and 19 of the SER (NUREG-0847). Fire Protection Report, Part V, Section 2.1, Safe Shutdown Procedures states, in part, the fire safe shutdown procedures contained in AOI-30.2 were developed based on calculations WBN-OSG4-031, Equipment Required for Safe Shutdown per 10 CFR 50 Appendix R, and WBN-OSG4-165, Manual Actions Required for Safe Shutdown Following a Fire. Calculation WBN-OSG4-165 is contained within drawing 1-45A897-1, Manual Actions Required for Safe Shutdown Following a Fire to 10 CFR 50 Appendix R. Contrary to the above, since initial plant licensing, the licensee failed to perform an adequate calculation to support fire safe shutdown procedure AOI-30.2. Specifically, for certain fire scenarios, the licensee failed to identify all equipment required to ensure availability of the TDAFW pump; and, for certain fire scenarios, the licensee established a non-conservative time requirement to mitigate spurious opening of a pressurizer PORV to prevent an undesired safety injection. This violation is of very low safety significance (Green). This issue was determined to be of very low safety significance based on the results of the IMC 0609, Appendix F, Fire Protection Significance Determination Process, Phase II Quantitative Screening Approach. A bounding risk assessment performed by a regional SRA reviewed the licensee and inspector risk evaluations and confirmed the CDF risk increase due to this condition was less than 1E-6, and therefore Green. This violation was documented in the licensees corrective action program as CRs 946764 and 999926. |
| 05000390/FIN-2016001-08 | 2016Q1 | Watts Bar | Charging Pump 1B-B Room Cooler Fan Bearing Failure | Inspectors identified an unresolved item (URI) associated with the failure of the 1B-B charging pump room cooler. This item is unresolved pending review of an equipment apparent cause evaluation that was performed after deficiencies were identified by inspectors in the past operability evaluation. On September 27, 2015, the licensee installed a new bearings on the 1B-B CCP room cooler fan shaft as part of planned maintenance (PM) under WO 115790759. The WO noted the room cooler had a broken lubrication line close to the point where it is attached to the outboard fan shaft bearing, but the new bearing on the fan shaft, including the outboard shaft bearing, were installed without an immediate repair of the lubrication line. The bearing replacements for WO 115790759 were accomplished in accordance with maintenance procedure 0-MI-0.16, Maintenance Guidelines for Belt Driven Equipment, Rev. 7. Appendix D, Bearing Installation, Step 14 requires, All remote lubrication lines, remote vibration attachments, etc. shall be verified as attached prior to return to service. The work order noted at this step that the lubrication line to the outboard fan shaft bearing was broken in half and will need to be replaced prior to return to service and the step was left blank. The licensee did not initiate a CR for this degraded condition. Due to the broken lubrication line, the outboard fan shaft bearing was the only fan shaft bearing that was not greased during installation. October 15, 2015, the licensee completed the PMT for the room cooler and noted it to be satisfactory. The broken lubrication line documented in the PM WO was identified and CR 1093983 was initiated to document the condition. This CR stated that the broken lubrication line did not affect the functionality of the fan and could be repaired at the next scheduled PM. This assessment was not questioned during the review of the CR for operability. The fan was returned to service and declared operable. On December 4, 2015, the room cooler failed in service. The licensee declared the 1BB charging pump inoperable and entered the applicable TS LCO. Investigation revealed that the outboard fan shaft bearing had failed. At this point, the inappropriate treatment of the degraded lubrication line under 0-MI-0.16 and the associated PMT was identified. This issue was documented in the licensees CAP in CR 1111791. The licensee performed a past operability evaluation (POE) for CR 1111791 which concluded the fan was operable until several hours before the time of the failure. The POE was based largely on statements from the bearing vendor indicating that the new bearing was pre-lubricated at the factory and should have performed under load for a long period of time without needing to be pre-greased at installation. The POE was hampered by the fact that the licensee did not retain the damaged bearing for failure analysis. The inspectors reviewed the POE and determined that it failed to adequately document sufficient information to either discount the broken lubrication line as a cause of the bearing failure or to identify another cause. In response, the licensee opened an investigation of the cause of the bearing failure under an equipment apparent cause evaluation. Because more information is necessary to evaluate the cause of the 1B-B CCP room cooler fan shaft bearing failure, future inspection is required to determine if a more than minor performance deficiency or violation exists associated with this issue. Specifically, the inspectors need to review the equipment apparent cause evaluation, which was not completed by the end of the inspection period. This is identified as URI 05000390/2016001-08, Charging Pump 1B-B Room Cooler Fan Bearing Failure. |
| 05000390/FIN-2016001-09 | 2016Q1 | Watts Bar | Appropriateness of Corrective Actions Associated with Safety Related Pump Mechanical Seal Issues and the Effect on Plant Response | The inspectors identified an URI associated with the timely and effective corrective action associated with an adverse trend in safety related pump performance, including mechanical seal degradation and failure. This item is unresolved pending review and evaluation of the licensees response to the CRs generated to determine if a performance deficiency exists. During Unit 1, 2015 fall outage, the 1A Safety Injection (SI) pump mechanical seal was replaced. The mechanical seal had degraded to a point at which the leakage was greater than the Technical Specification limit for ECCS leakage outside of containment. The inspectors identified several issues during a review of the Prompt Determination of Operability for CR 1125623 and WO 116050574 to replace the seal. Specifically, inspectors found that non-QA1 parts were being used for seal replacement, the seal was the original equipment manufacturer part from startup, the failure mechanism was not clearly understood, and an extent of condition review was not performed. The inspectors reviewed other safety related pump mechanical seal performance and corrective action program entries. The inspectors are awaiting the completion of the licensees evaluation to determine the licensees compliance with applicable procedures and TS relative to pump operability and ECCS leakage limits outside containment. Additional inspection activities are needed to determine the extent of condition and compliance with the procedures and TS. Pending the results of this additional inspection, an URI will be opened and designated as URI 05000390/2016001-09, Appropriateness of Corrective Actions Associated with Safety Related Pump Mechanical Seal Issues and the Effect on Plant Response. |
| 05000390/FIN-2016001-12 | 2016Q1 | Watts Bar | Licensee-Identified Violation | Technical Specification 5.7.1, Procedures, requires, in part, that written procedures shall be established, implemented, and maintained covering activities described in Regulatory Guide (RG) 1.33, Revision 2, Appendix A, February 1978; Appendix A, Section 6.v, requires procedures for Combating Emergencies and other Significant Events such as Plant Fires. Contrary to the above, the licensee provided operators inadequate procedural instructions to support fire safe shutdown. Specifically, since 2012, for certain fire scenarios, fire SSD procedures did not contain necessary steps to secure all reactor coolant pumps to prevent inadvertent RCS depressurization due to spurious opening of a pressurizer spray valve. Additionally, since initial plant licensing, for certain fire scenarios, fire SSD procedures did not contain necessary steps to isolate the normal charging line to prevent inadvertent RCS depressurization due to spurious opening of an auxiliary pressurizer spray valve. This violation is of very low safety significance (Green). This issue was determined to be of very low safety significance based on the results of the IMC 0609, Appendix F, Fire Protection Significance Determination Process, Phase II Quantitative Screening Approach. A bounding risk assessment performed by a regional SRA reviewed the licensee and inspector risk evaluations and confirmed the CDF risk increase due to this condition was less than 1E-6, and therefore Green. This violation was documented in the licensees corrective action program as CRs 954895 and 954957. |
| 05000390/FIN-2016001-07 | 2016Q1 | Watts Bar | Failure to Maintain Operating Logs | The NRC identified a NCV of 10 CFR 50, Appendix B, Criterion XVII, Quality Assurance Records, for the licensees failure to maintain sufficient records to furnish evidence of activities affecting quality. The licensee entered this issue into their corrective action program as CR 1127691. The inspectors determined that the licensees failure to document plant operations in the operating logs in accordance with OPDP-1 was a violation of 10 CFR 50, Appendix B, Criterion XVII, Quality Assurance Records. This violation constitutes a traditional enforcement violation because it impacts the NRC's ability to carry out its regulatory function. The failure to maintain accurate logs was more than minor because it would have likely caused the NRC to undertake further inquiry and was consistent with Enforcement Policy section 6.9.d.1 for a SL-IV violation. Crosscutting aspects are not assigned to traditional enforcement violations. |
| 05000277/FIN-2015004-01 | 2015Q4 | Peach Bottom | Failure to Ensure Design Basis of Emergency Diesel Generator Lubrication System | The inspectors identified a non-cited violation (NCV) of very low safety significance of 10 Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion III, Design Control, for not ensuring that the adequacy of PBAPS emergency diesel generator (EDG) lubrication oil (LO) supply was designed to withstand the effects of natural phenomena. Specifically, additional LO, evaluated by PBAPS to meet their EDG technical specification (TS) mission time of seven days of continuous operation, was housed in a non-class I structure that would be unable to withstand the effects of natural phenomena. PBAPS entered the issue into the correction action program (CAP) as issue report (IR) 02603369 and took immediate corrective actions to relocate the LO reserve inventory from their warehouse to the 135 elevation of the PBAPS radwaste building, which is a seismic class I structure The finding is considered more than minor because it is associated with the Protection Against External Factors attribute of the Reactor Safety Mitigating Systems cornerstone and adversely affected the cornerstones objective of ensuring reliability and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). The inspectors evaluated the significance of this finding using IMC 0609 Appendix A, The SDP for Findings at Power, Exhibit 2, Mitigating Systems Screening Questions. The inspectors determined that this finding was of very low safety significance (Green) because the finding is a design deficiency which did not result in an actual loss of functionality of the EDGs. This finding did not have a cross-cutting aspect because the most significant contributor of the performance deficiency (PD) occurred during the 1994 conversion to improved technical specifications (ITS) and, thus, was not reflective of current plant performance. Specifically, PBAPS current engineering change request (ECR) process would evaluate for natural phenomena considerations such as seismic, tornado, flood, etc. |
| 05000390/FIN-2015004-04 | 2015Q4 | Watts Bar | Shield Building Operability Requirements | The inspectors identified an unresolved item (URI) associated with the requirements of Watts Bar Unit 1 technical specification (TS) 3.6.15, Shield Building. Additional inspection is required to determine if the requirements of 3.6.15.B applied during a specific testing alignment. On September 10, 2015, the licensee conducted 0-SI-65-6-A, Emergency Gas Treatment System (EGTS) Train A 10-Hour Operation. During the 10-hour time period of the test when the EGTS was in service, the auxiliary gas building treatment system was also in service for a Unit 2 construction test. This unique ventilation combination is not normally experienced during the 0-SI-65-6-A surveillance. As a result, shield building annulus differential pressure fell below the limit established by TS surveillance requirement (TSSR) 3.6.15.1 limits for the entire duration of the 10-hr EGTS surveillance. TS limiting condition for operation (LCO) 3.6.15.B requires annulus pressure be restored when it is outside of limits with a required completion time of 8-hrs. The licensee considered the note associated with TS LCO 3.6.15.B, which states that the annulus pressure requirement is not applicable during ventilating operations, required annulus entries, or auxiliary building isolations not exceeding one hour in duration. The licensee considered the alignment they were in at the time to be ventilating operations and thus the requirements of TS LCO 3.6.15.B did not apply. The licensee further considered that the note, as written, allowed grace from the annulus pressure requirement for ventilating operations for an unlimited amount of time. The inspectors were concerned about a possible allowance in the TS to have grace from annulus pressure requirements for longer than the allowed LCO required action completion time. Furthermore, a basis for the note and what can be considered ventilating operations was not immediately apparent. Because more information is necessary to evaluate the proper applicability of TS LCO 3.6.15.B and the associated note, future inspection is required to determine if a more than minor performance deficiency or violation exists associated with this issue. Specifically, the inspectors need to determine if a TS compliance issue exists. This is identified as URI 0500390/2015004-04, Shield Building Operability Requirements. |
| 05000390/FIN-2015004-02 | 2015Q4 | Watts Bar | AFWST Permanent Plant Modification | The inspectors identified an unresolved item (URI) associated with the 50.59 screening performed for the installation of the auxiliary feedwater storage tank (AFWST). Additional inspection is required to determine if the plant modification which installed the tank would have required NRC permission in the form of a license amendment prior to the change. The AFWST is a 500,000 gallon source of clean water for the auxiliary feedwater (AFW) pumps. It was installed as part of the licensees post-Fukushima (FLEX) modifications to meet the mitigating strategies order (EA-12-049). The new tank was needed because the licensee determined they could not credit their existing condensate storage tanks (CSTs) for FLEX strategies due to seismic requirements necessary to survive the extended loss of AC power (ELAP) event. The AFWST was connected to the existing condensate system in the AFW supply piping upstream from the AFW pumps and downstream from the CSTs. The modification was evaluated in two separate DCNs, each with its own 50.59 applicability screening. DCN 60060 evaluated the installation of the tank and DCN 61422 evaluated the piping connections to the condensate system. The piping connections included new check valves in the CST piping to prevent AFWST inventory loss in the event the CSTs are damaged in the ELAP event. There were also two air-operated supply valves on AFWST outlet piping which automatically open on low pressure in the downstream condensate piping and also fail open on a loss of power or air. Inspectors noted a number of deficiencies in the 50.59 screening for DCN 61422. Inspectors determined that several potentially adverse impacts were introduced by the modification and were not adequately considered in the 50.59 screening. The licensee re-performed the screening and concluded that the modification would require a 50.59 evaluation due to adverse impacts brought up by the inspectors. Because more information is necessary to properly evaluate the 50.59 evaluation that was completed late in the quarter, future inspection is required to determine if a more than minor performance deficiency or violation exists associated with this issue. Specifically, the inspectors need to determine if prior NRC approval was required for the installation of the AFWST. This is identified as URI 05000390/2015004-02, AFWST Permanent Plant Modification. |
| 05000219/FIN-2015008-02 | 2015Q2 | Oyster Creek | Untimely Corrective Actions to Restore Design Conformance of Two SDV Vent & Drain Valves Pressure Regulator Valves | The NRC identified an NCV of Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion XVI, Corrective Action, for failure to promptly correct a condition adverse to quality. Specifically, corrective actions to restore design conformance of scram discharge volume (SDV) vent and drain valve pressure regulator valves V-6-961 and V-6-962 were not taken at the first opportunity of sufficient duration which was refueling outage 25 (1R25). Additionally, justification of the basis for deferral of corrective actions beyond the restart from 1R25 on October 2014, was not documented, reviewed, or approved by site management and/or oversight organizations as required by station procedure OP-AA-108-115, Section 4.5.5. Consequently, two non-conforming pressure regulator valves which perform a safety-related function remained installed following plant startup from 1R25, without appropriate evaluation and approval. Immediate corrective action included licensee determination that V-6-961 and 962 and the associated SDV vent and drain valves (V-15-119 and 121) remained operable, but non-conforming. Exelon entered the issue into their corrective action program as IR 2482851. The finding was more than minor because it was associated with the design control and barrier performance attributes of the Barrier Integrity cornerstone and adversely affected the cornerstone objective of ensuring the operational capability of the containment barrier to protect the public from radionuclide releases caused by accidents or events. Additionally, the finding was similar to example 5.c in Appendix E of Inspection Manual Chapter (IMC) 0612, because the control rod drive system was returned to service following 1R25 with two non-conforming (non-safety-related) pressure regulator valves installed in a safety-related application. The team determined the finding was of very low safety significance because it did not affect the reactor coolant system (RCS) boundary; did not affect the radiological barrier function of the control room, auxiliary building, or spent fuel pool systems or boundaries; and did not represent an actual open pathway in containment or involve a reduction in the function of hydrogen igniters. The team assigned a cross-cutting aspect in the area of Human Performance, Consistent Process (aspect H.13) because the organization did not use a consistent systematic approach to evaluate component operability after Exelon upgraded the classification of three pressure regulator valves from a non-safety to a safety-related status. |
| 05000219/FIN-2015008-01 | 2015Q2 | Oyster Creek | Use of an Analytical Method to Determine the Core Operating Limits Without Prior NRC Approval | The NRC identified a Severity Level lV non-cited violation (NCV) of Technical Specification (TS) 6.9.1.f.2 in that Exelon did not obtain NRC approval prior to using a specific analytical method to determine the core operating limits. Specifically, Exelon used an analytical method (TRACG04P) to determine the core operating limits (the average power range monitor protection settings which were identified in the Core Operating Limits Report (COLR)); however, that particular analytical method was not previously reviewed and approved by the NRC prior to Exelons use. Exelon submitted a corrective action issue report (IR) to evaluate the condition (IR2482042). The team determined that Exelon did not comply with TS 6.9.1.f.2 requirements in that Exelon used an analytical method to determine the core operating limits without prior NRC approval. The team determined that this was a performance deficiency that was within Exelons ability to foresee and correct. Because the issue had the potential to affect the NRCs ability to perform its regulatory function, the team evaluated this performance deficiency in accordance with the traditional enforcement process. Using the Enforcement Manual, the team characterized the violation as Severity Level IV because the underlying analytical method required NRC approval prior to use. Because this violation involves the traditional enforcement process and does not have an underlying technical violation that would be considered more-than-minor within the Reactor Oversight Process (ROP), the team did not assign a cross-cutting aspect to this violation in accordance with IMC 0612, Power Reactor Inspection Reports, Section 07.03.c. |
| 05000289/FIN-2015007-01 | 2015Q1 | Three Mile Island | Deficient Design Control for Verifying Reactor Building Fan Assembly Capability to Perform Design Basis Function | The NRC identified an NCV of Title 10 of the CFR, Part 50, Appendix B, Criterion III, Design Control, for failure to establish and implement adequate design control measures to assure that the reactor building (RB) fan assemblies were capable of performing their design function to mitigate a design basis loss of coolant accident (LOCA) event. Specifically, testing and design calculations used a non-conservative RB ventilation system alignment to determine the brake horsepower of the RB fan motors during a LOCA. As a result, engineers had not evaluated the capability of the RB fan motors to operate above their nameplate full load rating to perform their intended safety function. Additionally, RB fan motor electrical overload protection analyses were incorrect. Immediate corrective actions included interim calculations which demonstrated that the RB fan assemblies would remain capable of performing their safety functions and that the emergency diesel generators were capable of supplying the additional electrical load requirements. Exelon entered the issues into their corrective action program as IRs 2458932, 2458929, and 2451855. This finding was more than minor because it was associated with the design control attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective of ensuring the operational capability of the containment barrier to protect the public from radionuclide releases caused by accidents or events. Additionally, the finding was similar to example 3.j in Appendix E of IMC 0612, in that the engineering calculation error resulted in a condition where there was reasonable doubt of the operability of the RB fan assemblies to perform their safety function during a design basis LOCA. The team determined the finding was of very low safety significance because it: did not affect the reactor coolant system (RCS) boundary; did not affect the radiological barrier function of the control room, auxiliary building, or spent fuel pool systems or boundaries; and did not represent an actual open pathway in containment or involve a reduction in the function of hydrogen igniters. This finding was not assigned a cross-cutting aspect because the underlying cause was not indicative of current performance in that the non-conservative calculation error occurred in 1993. |
| 05000289/FIN-2015007-02 | 2015Q1 | Three Mile Island | Untimely Identification and Correction of Degraded BWST Level Transmitter Cold Weather Protection Equipment | The NRC identified an NCV of Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Appendix B, Criterion XVI, Corrective Action, for failure to promptly identify and correct degraded borated water storage tank (BWST) level transmitter instrument line cold weather protection equipment. Specifically, station personnel performed periodic maintenance and testing activities to verify the adequacy of cold weather protection for the BWST level transmitters prior to the onset of cold weather, but did not identify existing uninsulated sections of the instrument lines or degraded heat trace circuit continuity. Consequently, on February 15, 2015, the sensing line for BWST level transmitter DH-LT-808 froze which challenged the operators capability to successfully perform a critical design basis manual action. Namely, swapover from the injection to recirculation phase of ECCS operation following a LOCA. Immediate actions included entering the applicable technical specification (TS) limiting condition of operation (LCO), thawing the frozen instrument line, restoring DH-LT-808 to service, and exiting the TS LCO. Exelon entered the cold weather protection issue into their corrective action program as issue reports (IR) 2445164, 2451342, 02452858, and 02454925. This finding was more than minor because it was associated with the equipment and human performance attributes of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team determined that the finding was of very low safety significance because it did not affect design or qualification, did not represent a loss of system, did not cause at least one train of BWST level instrumentation to be inoperable for greater than its TS LCO allowed outage time, and did not involve external event mitigation systems. The team assigned a cross-cutting aspect in the area of Human Performance, Procedure Adherence (aspect H.8), because station personnel did not follow processes, procedures, and work instructions when performing maintenance and operational activities that should have identified degraded BWST level instrument cold weather protection equipment associated with missing insulation and loss of heat trace circuit continuity. |
| 05000220/FIN-2014007-02 | 2014Q4 | Nine Mile Point | Deficient Design Control of NMP Unit 1 Electrical Protection Design to Ensure Survivability of Safety-Related Loads | The team identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for failure to verify the adequacy of the Nine Mile Point Unit 1 electrical design during a design basis loss of coolant accident (LOCA) event with sustained degraded grid voltage (DGV). Specifically, Exelon did not verify Class 1E loads would not be damaged or become unavailable for a design basis LOCA with a degraded voltage condition between the degraded voltage setpoint and the loss of voltage setting for the degraded voltage time delay of 21 +/- 3 seconds and subsequent reconnection to the emergency diesel generator. Immediate corrective actions included preliminary evaluation of the safetyrelated MOV that operate during the first 21 seconds of the accident, which determined there was reasonable assurance the MOV protective devices would not actuate during sustained DGV concurrent with a design basis LOCA. Exelon entered this issue into their corrective action program as issue reports 2387818 and 2392780. The finding was more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team determined the finding was of very low safety significance (Green) because it was a design deficiency confirmed not to result in loss of operability or functionality. The team determined this issue had a cross-cutting aspect in the area of Problem Identification and Resolution, Operating Experience (Aspect P.5), because the organization did not effectively collect, evaluate, and implement relevant internal and external operating experience in a timely manner. Despite NRC Regulatory Issue Summary 2012-11, Adequacy of Station Electric Distribution System Voltages, and NRC Component Design Bases Inspections identifying similar performance deficiencies at other Exelon facilities during the last 3 years, the Nine Mile Point staff did not effectively evaluate and resolve this operating experience. |
| 05000410/FIN-2014007-01 | 2014Q4 | Nine Mile Point | Deficient Design Control of NMP Unit 1 Electrical Calculations to Evaluate Minimum Voltages to Class 1E Accident Initiated Motors and MOVs during a Design Basis Event | The team identified a Green NCV of Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Appendix B, Criterion III, Design Control, for failure to verify and assure, in Nine Mile Point Unit 1 design basis calculations, that adequate voltages would be available to Class 1E accident initiated motors, motor-operated valves (MOV), and control circuits powered from the 4160 V, 600 V, and 120 V distribution systems during a design basis loss-of-coolant accident (LOCA) with offsite power available. Specifically, Exelon did not identify and evaluate the minimum transient voltage for the design basis LOCA event regarding accident initiated motors, MOVs, and control circuits, and did not evaluate the capability of the safety-related main steam isolation valve motor brakes. Immediate corrective action included preliminary calculations using the design grid voltage sag, which determined the Reserve Service Station Transformer load tap changers, motor control center (MCC) control circuits, MOVs, and the main steam isolation valve motor brakes would have adequate voltage to remain capable of performing their safety functions. Exelon entered the issues into their corrective action program as issue reports 2386719, 2386824, 2387652, 2387888, 2392928, and 2393299. This finding was more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team determined the finding was of very low safety significance because it was a design deficiency confirmed not to result in a loss of safety-related MCC MOV operability or functionality. This team assigned a cross-cutting aspect associated with this finding because the long-standing performance deficiency continued during and after Exelons review of related internal and external operating experience from 2012 to 2014. The team determined this finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Operating Experience (Aspect P.5), because Nine Mile Point Unit 1 staff did not effectively collect, evaluate, and implement relevant internal and external operating experience in a timely manner. |
| 05000220/FIN-2014007-03 | 2014Q4 | Nine Mile Point | Deficient Design Control of Protective Device Sizing for Unit 1 Core Spray Injection Motor-Operated Valves | The team identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, because Exelon did not verify the design adequacy of Nine Mile Point Unit 1 electrical power to safety-related MOVs to support their design function during design basis events. Specifically, Exelon did not verify that the thermal/magnetic breaker (TMB) protection on core spray (CS) loop injection MOV circuits 1V-40-01, 1V-40-09, 1V-40-10, and 1V-40-11 were properly sized to support the design function of repetitive MOV operation (throttling) in response to a design basis loss-of-coolant accident (LOCA). Routine throttling operation of the CS injection valves could potentially cause a TMB trip and loss of power to the MOV leading to the valve failing in an indeterminate position and not being capable of performing its design function to control reactor pressure vessel (RPV) level. Immediate corrective action included guidance to control room operators to close three of the MOVs when required to maintain RPV level and only use MOV 1V-40-09 which had a TMB tripping design of 17 seconds Exelon entered this issue into its corrective action program as issue report 2393386. The finding was more than minor because it is associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team determined that the finding was of very low safety significance (Green) because it was a design deficiency confirmed not to result in loss of operability or functionality. The team determined that the central cause of this finding was not reflective of current performance or current plant modification processes. Therefore no cross-cutting aspect was assigned. |
| 05000220/FIN-2014007-04 | 2014Q4 | Nine Mile Point | Licensee-Identified Violation | 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. In addition, NRC letter dated June 2, 1977, required all licensees to verify existing plant design or propose modifications to ensure onsite emergency power systems met certain criteria including Staff Position 1. Staff Position 1(c)(2) stated the DGV protection time delay duration, shall minimize the effect of short duration disturbances from reducing the availability of the offsite power source(s). Staff Position 1(c)(3) stated the DGV protection time delay shall be established such that the allowable time duration of a DGV condition at all distribution system levels shall not result in failure of safety systems or components. Contrary to the above, prior to October 9, 2014, Exelon did not adequately evaluate the sequencing of Unit 2 safety-related loads and associated transient voltages to the Unit 2 Class 1E accident initiated motors and MOVs on the-safety related buses and MCCs during the initiation of a design basis loss of coolant event, subsequent unit trip, and resulting sag of the 115kV grid. Specifically, Exelon did not ensure the chosen DGV protection time delay duration: (1) minimized the effect of short duration disturbances from reducing the availability of offsite power sources; and, (2) maintained voltage requirements for safety-related loads to ensure that failures of safety-related systems would not occur. Exelon did not identify the resulting voltage transients, minimum 4kV motor and 600V MOV starting voltages, associated motor actuator output torque, and control circuit voltages to safety-related MOV motors. Exelon identified these deficiencies as a result of their review of NRC RIS 2011-12R1 and contracted in September 2014 to have the electrical calculations revised. Initial results, using the grid operator specified grid sag of 3.5 percent following a unit trip, indicated that the transient 4KV safety bus voltage would be too low to reset the DGV relays. This would result in the unintended disconnection of offsite power and transfer to the EDG. Immediate actions included reevaluation of the postulated grid sag and transient 4 kV bus voltages. Exelons subsequent assessment concluded that grid stability had improved and offsite power remained operable. The team reviewed Exelons assessment and immediate actions and found them to be reasonable. Exelon entered this issue into the corrective action program as IRs 2393336 and 2392930. The team determined that the finding was of very low safety significance (Green) in accordance with IMC 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, Mitigating Systems, because it was a design deficiency confirmed not to result in loss of operability or functionality. |
| 05000286/FIN-2014007-01 | 2014Q2 | Indian Point | Deficient Design Control Results in Non-Qualified Component Installed in Harsh Environment for Unit 3, BFD-FCV-406B Actuator | The team identified a Green non-cited violation of Title 10 Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion III, Design Control, because Entergy did not ensure the control air pressure regulator (IA-PCV-1548) for Unit 3 auxiliary boiler feedwater (ABFW) flow control valve BFD-FCV-406B was suited and designed to perform its safety-related function. Specifically, IA-PCV-1548 was not designed or qualified for use in the harsh environment area where it was located. Immediate corrective actions included evaluation of IA-PCV-1582 and BFD-FCV-406B to verify component operability. The issue was entered into the corrective action program as condition report IP3-2014-1364, to further evaluate both the extent-ofcondition and the stations processes for maintaining configuration control over mechanical components installed in harsh environment areas. The finding was more than minor because the finding was associated with the Design Control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of assuring the reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Additionally, the issue was similar to example 3.j in Appendix E of Inspection Manual Chapter 0612, in that the design control issue resulted in a reasonable doubt of operability. The team determined the finding was of very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in a loss of operability. The finding had a cross-cutting aspect in the area of Human Performance, Design Margin (H.6), because Entergy did not maintain the operational temperature design margin for the control air pressure regulator to the ABFW flow control valve. The margin between the ABFW pump room peak environmental temperature and the design/qualified temperature of IA-PCV-1582 was not carefully guarded and changed only through a systematic and rigorous process. |
| 05000336/FIN-2014003-03 | 2014Q2 | Millstone | Failure to Adequately Maintain EALs | The inspectors identified a Green NCV associated with emergency preparedness planning standard Title 10 of the Code of Federal Regulations (10 CFR) 50.47(b)(4) and the requirements of Sections IV.B and IV.C of Appendix E to 10 CFR 50. Specifically, Dominion did not maintain the Millstone Units 2 and 3 emergency action level (EAL) schemes for assessing a loss of forced flow cooling during refueling operations. Dominion entered this issue into the CAP and implemented temporary corrective actions which included procedure changes to direct operators to the shutdown safety assessment checklists to determine representative reactor coolant system (RCS) temperature increases in order to assess the initiating conditions (ICs) for this situation. The inspectors determined that the failure by Dominion to provide site specific criteria for operators to adequately implement the EALs for a loss of forced flow cooling during refueling was a performance deficiency that was reasonably within their ability to foresee and prevent. The finding is more than minor because it is associated with the Procedure Quality attribute of the Emergency Planning Cornerstone and affected the cornerstone objective to ensure that Dominion is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. In accordance with IMC 0609, Appendix B, Emergency Preparedness Significance Determination, the inspectors determined that this finding is of very low safety significance (Green) because the performance deficiency was an issue where two EAL ICs had been rendered ineffective such that an Unusual Event and an Alert would not be declared, or declared in a degraded manner for a loss of forced flow cooling during refueling. The finding has a cross-cutting aspect in the area of Problem Identification and Resolution, in that Dominion did not implement a CAP with a low threshold for identifying issues. Dominions self-assessment for two previous NCVs regarding EAL deficiencies failed to identify the lack of specific criteria to assess the ICs for EALs EU1.2 and EA2.1 for a loss of forced cooling flow during refueling. |
| 05000336/FIN-2014003-02 | 2014Q2 | Millstone | Failure to Utilize Respiratory Protection as Specified in Work Control Documents | A self-revealing Green NCV of Technical Specification (TS) 6.8.1; Regulatory Guide 1.33, Appendix A; Radiation Work Permits (RWP); and as low as reasonably achievable (ALARA) procedures was identified for Dominions failure to utilize respiratory protection, as required by the applicable RWP and associated ALARA evaluation for work on replacement of valve 2-SI-227 on April 20, 2014. This failure resulted in an unplanned intake of radioactive material for one worker. Dominion subsequently enforced the respiratory protection requirements to complete the work and entered this issue into their corrective action program (CAP) as condition report (CR) 546439. Failure to use respiratory protection during machining work as required by Dominion procedure was a performance deficiency that was reasonably within Dominions ability to foresee and correct. The inspectors determined that the performance deficiency was more than minor because it affected the Radiation Safety Occupational Radiation Safety Cornerstone attribute of Program and Process associated with exposure/contamination controls, because it resulted in the unintended internal exposure of a worker. A crosscutting aspect of Human Performance, Conservative Bias, was associated with the finding. Specifically, radiation protection staff did not adhere to the RWP requirements. |
| 05000336/FIN-2014003-01 | 2014Q2 | Millstone | Failure to Maintain Adequate Procedure For RCS Drain/Fill | The inspectors identified a Green NCV of TS 6.8.1, Procedures, for Dominions failure to maintain an adequate procedure for reactor filling and draining that incorporates guidance contained in NRC Generic Letter 88-17. Specifically, OP2301E, Draining the RCS, permitted operation in a reduced RCS inventory condition without ensuring redundant means of level indication contrary to the inventory control requirements of OU-M2-201, Shutdown Safety Assessment Checklist. The failure to maintain an adequate procedure for operating in reduced inventory conditions is a performance deficiency. The inspectors determined this performance deficiency is more than minor because it is associated with the Initiating Events cornerstone attribute of equipment performance and affects the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown operations. Specifically, inadequate procedural guidance increased the likelihood that operators could experience a loss of level indication during the reduced inventory condition. The inspectors evaluated the significance of the finding using IMC 0609 Appendix G, Shutdown Operations Significance Determination Process, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings, and the issue screened to a Phase 2 analysis. Using the guidance contained in IMC 0609, Appendix G, Attachment 2, Phase 2 Significance Determination Process Template for PWR During Shutdown, the inspectors worked with regional and headquarters senior reactor analysts to determine the issue screened to Green. The inspectors determined this issue had a cross-cutting aspect in the area of Human Performance, Avoid Complacency, where individuals recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Specifically, the latent error of considering L-112 and LI-112 as independent level instruments even though a single failure impacted both instruments contributed to the issue. |
| 05000410/FIN-2014002-02 | 2014Q1 | Nine Mile Point | Invalid Low Reactor Water Level Results in Unit 2 Automatic Reactor Scram | Inspectors documented a self-revealing Green NCV of Technical Specification (TS) 5.4, Procedures, for CENGs failure to ensure proper communication of a change in work scope prior to implementation. Specifically, on March 10, 2014, valve label replacements at Unit 2 commenced in a trip sensitive area while the plant was on-line, although the work was previously scheduled to be conducted when the reactor was shut down. This change in work scope was not properly reviewed and communicated to the supporting work group prior to implementation. As a result, a reactor scram occurred when an instrumentation and control (I&C) technician inadvertently contacted an instrument rack located in a trip sensitive area while performing a valve label replacement. CENG generated condition report (CR)-2014- 001963 to document the Unit 2 reactor scram due to the technician contacting the instrument line while cutting the valve label. Immediate corrective actions included developing site communications to enhance awareness of trip sensitive equipment and to provide additional flagging barriers to ensure trip sensitive components are not inadvertently contacted. This finding is more than minor because it is associated with the human performance attribute of the Initiating Events Cornerstone and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. Specifically, CENG staff did not properly ensure that the scope change was properly reviewed and communicated to the supporting work group prior to implementation. This resulted in work being performed while Unit 2 was online and a subsequent automatic reactor scram when an instrument rack was inadvertently contacted. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 1 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2012, the inspectors determined that this finding is of very low safety significance (Green) because while the performance deficiency caused a reactor scram, it did not result in the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The finding has a cross-cutting aspect in the area of Human Performance, Conservative Bias, because CENG failed to use proper decision makingpractices that emphasize prudent choices over those that are simply all. |
| 05000220/FIN-2014002-01 | 2014Q1 | Nine Mile Point | Inadequate Design Control Measures Employed During Control Room HVAC Modification | The inspectors identified a Green NCV of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion III, Design Control, because CENG did not implement adequate design controls to ensure piping in the Reactor Building Closed Loop Cooling (RBCLC) system remained operable while implementing a modification to the Unit 1 control room heating and ventilation system. Specifically, while implementing the modification, CENG personnel removed permanent plant supports and piping for the safetyrelated RBCLC system and did not fully assess how this change could impact the operability of the system with respect to a hydraulic shock or seismic acceleration event. In response to this observation, CENG initiated CR-2014-001676 and evaluated the condition for operability. Existing temporary supports were enhanced to provide additional margin by bracing the structure for horizontal loads. An extent of condition walkdown was performed and no additional issues of concern were identified. Subsequently, CENGs operability review determined the RBCLC system had remained operable. This finding was more than minor because it was associated with the design control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, while implementing the modification, CENG removed permanent plant supports and piping for the safety-related RBCLC system and did not fully assess how this change could impact the operability of the system if a hydraulic shock or seismic acceleration occurred. This finding is also similar to examples 3.j and 4.k in IMC 0612, Appendix E, Examples of Minor Issues, where a temporary modification was installed without adequate design information and adequate design controls were not implemented leading to a reasonable doubt of operability of plant components. 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 this finding is of very low safety significance (Green) because the performance deficiency was a design or qualification deficiency that did not result in the inoperability of the RBCLC system. The finding has a cross-cutting aspect in the area of Human Performance, Work Management, because CENG failed to implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. Specifically, CENG failed to ensure that the installed temporary supports were adequate to ensure the RBCLC piping would not be stressed above code allowable values in the event of a seismic acceleration or hydraulic shock event prior to removing the permanently installed seismic supports. |