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| Quarter | Site | Title | Description | |
|---|---|---|---|---|
| 05000261/FIN-2009002-01 | 2009Q1 | Robinson | Operability of the A emergency diesel generator from February 9 to March 10, 2009 | The inspectors identified an unresolved item associated with two events in which maintenance technicians performed maintenance on the A emergency diesel generator (EDG) without pre-planning and performing the activity in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances. This item is unresolved pending further inspection to determine whether those events rendered the A emergency diesel generator inoperable. On February 9, after the licensee started the A EDG for a routine surveillance test, operators noted a fuel oil leak of approximately 5 drops/minute from the inboard fitting on the fuel oil line from the duplex filter to the fuel injector for the #1 cylinder. After the control-room staff determined that a leak of that magnitude did not render the EDG inoperable, they completed the surveillance test and then initiated work request 369565 to repair the leaking fitting. After the test was complete, as directed by a maintenance supervisor, without any written instructions, and without the knowledge of the control-room staff, a maintenance technician tightened the leaking fitting with a wrench and reported to his supervisor some nut movement. The next day, work order 1496721 was first initiated in response to the subject work request and then cancelled with the following annotation; per the system engineer, there is no concern at this time. On March 9, after the licensee started the A EDG for another surveillance test, operators noted a fuel oil leak of approximately 100 drops/minute from the same fitting that had leaked during the February 9 test. While the March 9 test was underway, after Operations asked Maintenance to investigate the leak, and as directed by a maintenance supervisor, a maintenance technician used a wrench to check the fitting for tightness. Although the technician reported no nut movement, as a result of this activity the leak flow rate increased to what the licensee characterized as a steady stream. Following this increase in the leak rate, the control-room staff shut down the EDG and declared it inoperable. The licensee initiated a work request to repair the leak, converted that work request to a corresponding work order, and then under that work order replaced the leaking fitting on March 10. After removal, visual examination of the subject fitting revealed that it was cracked. After reviewing the related circumstances, the licensee determined that the February 9 maintenance activity had probably cracked the fitting. Because the February 9 maintenance activity and the March 9 maintenance activity both involved maintenance on a safety-related component without any documented instructions, both were performance deficiencies with respect to TS 5.4.1. This TS requires the licensee to establish, implement, and maintain written procedures covering the applicable procedures in Regulatory Guide 1.33. Regulatory Guide 1.33, Appendix A, section 9, Procedures for Performing Maintenance, states that maintenance that can affect the performance of safety-related equipment should be properly pre-planned and performed in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances. These performance deficiencies are in the licensees corrective action program as AR 325384. Planned corrective actions are unknown because the licensee had not completed their investigation of this event before the end of the inspection period. Further inspection is required to determine whether the A EDG was operable during the period from February 9 through March 10. Specifically, further inspection should review any analyses produced by the licensee and/or evidence gathered by the licensee that relates to the operability of the A EDG during the subject period, and subsequently determine whether those results demonstrate that the A EDG was operable during that period. Pending completion of that review and determination, this issue is being treated as an unresolved item, and has been designated URI 05000261/2009002-01, Operability of the A emergency diesel generator from February 9 to March 10, 2009 |
| 05000261/FIN-2009003-01 | 2009Q2 | Robinson | Failure to properly restore service water to the evaporative air coolers resulting in emergency diesel generator inoperability. | A self-revealing finding was identified for the licensees failure to follow procedures while restoring auxiliary building evaporative air coolers to service. Although a violation of regulatory requirements was not identified, this failure was a performance deficiency with respect to licensee procedure PRO-NGGC-0200, Procedure Use and Adherence, Rev. 10, which requires all personnel who use procedures to understand the impact of their actions on personnel or equipment before taking action. As a result, the A emergency diesel generator (EDG) was declared unavailable and inoperable. At the end of this inspection period, the licensee had not yet completed their evaluation of this finding, and had consequently not yet developed corresponding corrective actions. This finding is in the licensees corrective action program as AR 332970. This finding is more-than-minor because it affected the Equipment Performance attribute of the Mitigating Systems cornerstone, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences, in that this finding resulted in unplanned unavailability of an emergency diesel generator. Using Attachment 4 of IMC 0609, the significance of this finding was determined to be of very low safety significance (GREEN), because although the finding could degrade the Emergency AC power function in the Mitigating Systems cornerstone, the finding was not a design or qualification deficiency confirmed not to result in loss of operability or functionality, did not represent a loss of system safety function, did not represent actual loss of safety function of a single train for longer than its TS Allowed Outage Time, did not represent an actual loss of safety function of one or more non-TS Trains of equipment designated as risk-significant, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding has a cross-cutting aspect in the Work Practices component of the Human Performance area because the licensee did not communicate human error prevention techniques such that work activities were performed safely, in that the licensee did not communicate instructions for the sequence of valve operations during the pre-job brief and the licensee proceeded in the face of uncertainty by operating system components when the current system alignment was not verified. (H.4(a)) (Section 1R15.1 |
| 05000261/FIN-2009003-02 | 2009Q2 | Robinson | Failre to follow procedures while performing maintenance on an emergency diesel generator. | The inspectors identified a green non-cited violation of Technical Specification (TS) 5.4.1, Administrative Controls (Procedures) associated with two events in which maintenance technicians performed maintenance on the A emergency diesel generator (EDG) without pre-planning and performing the activity in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances. In both instances maintenance technicians tightened a fuel oil fitting on the A emergency diesel generator which caused increased leakage from that fitting and the unplanned unavailability and inoperability of the diesel generator. In response to this finding, the licensee revised their Maintenance Administration Program to clearly communicate that skill of the craft work on safety related equipment is prohibited without a procedure/work order, and held stand-down meetings to retrain all maintenance and planning personnel on work practices for safety related structures, systems and components. This finding is in the licensees corrective action program as AR 325384. This finding was more-than-minor because it is associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and affected the availability of the emergency diesel generator to respond to a loss of offsite power event. Using Attachment 4 of MC 0609, Significance Determination Process, this finding screened as having very low safety significance (Green) because the finding was not a design or qualification deficiency confirmed not to result in loss of operability or functionality, did not represent a loss of a system safety function, did not represent an actual loss of safety function of a single train, did not represent an actual loss of safety function of one or more non-Tech Spec Trains of equipment designated as risk-significant, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding has a cross-cutting aspect in the Work Practices component of the Human Performance cross cutting area because personnel work practices did not support human performance, in that the licensees work practices did not ensure supervisory and management oversight of work activities such that nuclear safety is supported. (H.4(c)) (Section 1R15.2 |
| 05000261/FIN-2009003-03 | 2009Q2 | Robinson | Failure to address environmental conditions associated with freeze-protection temperature sensors. | A green self-revealing finding was identified for the licensees failure to identify the environmental conditions that temperature sensors in certain freeze-protection circuits could experience after routine installation of cold-weather enclosures during cold-weather operation. Although a violation of regulatory requirements was not identified, this failure was a performance deficiency with respect to the licensees procedure EGR-NGGC-0005 (Engineering Change) which requires, in part, that the licensee identify the functional performance requirements of each structure, system and component being modified in all possible operational configurations. In this circumstance, the licensees modification to the freeze-protection circuits for the steam generator power operated relief valve sensing lines, installed the freeze-protection temperature sensors in a location where a heated enclosure is routinely installed for cold-weather protection. With the heated enclosure surrounding the temperature sensors the freeze protection circuitry failed to energize during freezing conditions and subsequently allowed the sensing line for the B steam generator power operated relief valve to freeze, which in turn caused the B steam generator power operated relief to open at full power operation. This finding is in the licensees corrective action program as AR 339914. At the end of this inspection period, the licensee had not yet completed their evaluation of this finding, and had consequently not yet developed corresponding corrective actions. This finding is more-than-minor because it is associated with the Equipment Performance attribute of the Initiating Events cornerstone and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during power operations, in that this finding created conditions which caused an event that upset plant stability during power operations. Using Appendix A of the Significance Determination Process (SDP) described in MC 0609, this finding did not screen as green because it was a transient initiator contributor and because the finding contributed to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available, in that this finding created conditions that caused a S/G PORV to open during power operation, and rendered inoperable the automatic functions of that PORV. A regional Senior Reactor Analyst performed a Phase 3 evaluation under the Significance Determination Process. The performance deficiency was determined to be of very low safety significance (Green). The evaluation was accomplished using the NRCs Probabilistic Risk Assessment computer model of the plant with basic event MSS-ADV-CC-RV1-2, FAILURE OF SG-B PORV RV1-2, set to always fail. The model was quantified with a one day exposure period. The dominant accident sequences involved Steam Generator Tube Ruptures with complications, partially due to the finding, in depressurizing and cooling down. Consequently, the Residual Heat Removal System was not placed into service resulting in core damage and a Large Early Release. The major assumptions included that recovery of the failed component was possible and common cause inclusion was not appropriate. This finding has a cross-cutting aspect in the Resources component of the Human Performance area because the licensee did not provide and ensure that complete, accurate, up-to-date design documentation were available and adequate to plant personnel, in that the licensee did not ensure that Attachment 7 to EGR-NGGC- 005 was adequate to enable engineers to identify a potential interference between the modification described in EC 70032 and the program described in OP-925 (Cold Weather Preparations). (H.2(c)) (Section 4OA2.2 |
| 05000261/FIN-2016003-01 | 2016Q3 | Robinson | Failure to Scope Tainter Gate Flood Protection Features in Maintenance Rule Resulting in Degraded Performance | A self-revealing Green NCV of 10 CFR 50.65(b)(2)(ii) was identified for the failure to scope the external flood protection function of the Robinson Lake Dam spillway (Tainter) gates in the maintenance rule (MR) monitoring program. The failure to include the Tainter gates in the MR program resulted in ineffective maintenance being performed and subsequent degraded opening capability which challenged the availability of safety-related equipment during design basis rainfall events due to site flooding. The licensee took immediate corrective actions to replace/refurbish the chains to both gates and completed full open testing to restore their functionality. In addition, the licensee has developed and initiated implementation of an action plan to improve and ensure reliability of the gates, and initiated actions to revise the MR scoping program to include the Tainter gates. The issue was entered into the licensees CAP as CR 2035500. The failure to scope the flood protection function of the Lake Robinson Dam Tainter gates in the maintenance rule monitoring program was a PD. The finding is more than minor because it is associated with the protection against external factors (i.e., flood hazard) attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, failure to monitor flood protection features associated with the Tainter gates resulted in degraded gate opening performance that could have resulted in site flooding during design basis rainfall events and adversely impact multiple trains of safety-related equipment due to water intrusion. Using IMC 0609, Appendix A, The SDP for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined the finding involved the degradation of equipment specifically designed to mitigate flooding events. In accordance with Exhibit 4, External Events Screening Questions, the inspectors determined that the finding represented a degradation of two or more trains of a multi-train system or function during an external flooding event, therefore it required a detailed risk evaluation. A regional senior reactor analyst completed a detailed risk evaluation in accordance with NRC IMC 0609 Appendix A, and Appendix M, Significance Determination Process Using Qualitative Criteria, using the latest NRC Robinson Standardized Plant Analysis Risk model. The high uncertainty associated with estimating flood frequencies was the reason for using the NRC IMC Appendix M approach. The major analysis assumptions included a one-year exposure interval, recovery credit for opening the Tainter gates subsequent to binding of the chain, and limited credit for FLEX flooding mitigation strategies. If the rainfall produced a water surface elevation which would overtop the dam, the dam was considered failed and the ultimate heat sink lost. The rainfall frequencies requiring gate operation were estimated using a combination of National Oceanographic and Atmospheric Administration rainfall data and a probabilistic technique to establish precipitation frequency estimates performed by the licensee. The dominant sequence was a flood event inducing a non-recoverable loss of offsite power and loss of the emergency buses with a failure of the operators to manually recover the Tainter gates and failure of the operators to depressurize the steam generators to facilitate FLEX injection leading to a loss of core heat removal and core damage. The risk was mitigated by the low flood frequency, and the likely recovery of the Tainter gates prior to site flooding. There were additional conservatisms which were not applied to the result but would reduce the risk. These included the fact that the plant would be shutdown prior to flooding impacting safety-related equipment, which would reduce decay heat cooling required, and additional FLEX flooding strategies which could provide cooling even if the dam was lost. The risk increase due to the performance deficiency was < 1.0E-6/year, a Green finding of very low safety significance. The licensees analysis and full scope probabilistic risk assessment model produced a similar result. The inspectors determined that since the scoping of plant systems had occurred more than three years in the past, the finding did not represent current plant performance and therefore did not have a cross-cutting aspect associated with it. |
| 05000261/FIN-2016003-02 | 2016Q3 | Robinson | Failure to Assess and Manage Risk for Main Turbine Trip Maintenance Resulting in Turbine/Reactor Trip | A self-revealing Green non-cited violation (NCV) of 10 CFR 50.65(a)(4) was identified for the failure to adequately assess and manage the increase in risk associated with online maintenance activities involving the removal of the cover to the main turbine trip mechanism in order to perform visual inspections. During removal of the cover, the turbine trip mechanism lever was contacted causing an automatic turbine/reactor trip. The licensee took immediate corrective actions to reemphasize the need to enter all applicable types of work activities into the work management process and to conduct formal risk assessments in accordance with the risk management program. The licensee entered this issue into the corrective action program (CAP) as condition report (CR) 2056554. The licensees failure to adequately assess and manage the risk of maintenance associated with visual inspection of the turbine trip mechanism was a performance deficiency (PD). The inspectors evaluated the PD in accordance with IMC 0612, Appendix B, Issue Screening, and determined it to be more than minor because it impacted the human performance attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during power operations. Specifically, the failure to assess and manage the risk associated with removing the turbine trip mechanism cover to conduct visual inspections resulted in a turbine/reactor trip. The inspectors evaluated the finding in accordance with IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process. In accordance with Appendix K, the inspectors requested that a regional Senior Reactor Analyst (SRA) independently evaluate the risk. A Region II SRA performed an analysis of the risk deficit for the unevaluated condition associated with the work activity on the turbine trip mechanism. The latest Robinson Standardized Plant Analysis Risk (SPAR) model was used to calculate an incremental core damage probability deficit (ICDPD). The result was an ICDPD of 3.74E-7 and represented the increase in core damage probability associated with a turbine/reactor trip coincident with the dedicated shutdown diesel generator being out of service at the time of the event. In accordance with IMC 0609, Appendix K, because the calculated ICDPD was not greater than 1E-6, the finding was screened as having very low safety significance (Green). The cause of the PD was directly related to the cross-cutting aspect of work management in the cross-cutting area of human performance because the licensee failed to adequately implement a process of planning, controlling, and executing work activities such that nuclear safety was the overriding priority. Specifically, the licensee failed to adequately assess, manage, and implement risk management actions for activities associated with trip sensitive equipment. |
| 05000261/FIN-2017001-01 | 2017Q1 | Robinson | Failure to Perform General Visual Examinations of Containment Moisture Barriers Associated with Containment Liner Leak -Chase Test Connection | Green . An NRC- identified Green non -cited violation ( NCV ) of 10 CFR Part 50.55a, Codes and Standards, was identified for the failure to perform general visual examinations of moisture barriers in the containment leak -chase channel test connections in accordance with the American Societ y of Mechanical Engineers Boiler and Pressure Vessel Code (ASME BPVC), Section XI, Subsection IWE , Requirements for Class MC and Metallic Liners of Class CC Components of Light -Water Cooled Plants . Following the inspectors identification of this issue, t he licensee initiated actions to conduct the re quired visual examinations during the March 2017 refueling outage and initiated actions to revise the containment inservice inspection (ISI) plan such that the required examinations will be performed in the future . This issue was entered into the licensees corrective action program (CAP) as nuclear condition report (NCR) 02109909. The failure to conduct the required visual examination of moisture barrier material in accordance with the ASME B PVC , Section XI, Subsection IWE , was a performance deficiency (PD) . The finding was of more than minor significance because, if left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, visual examinations of mois ture barriers associated with the containment leak -chase channel test connections provide assurance that the containment metal liner and liner seam welds remain capable of performing its intended safety function. In the absence of such examinations, corro sive conditions at the moisture barrier (concrete -to-tubing interface) could go undetected. As a result, degradation of inaccessible portions of the containment liner could progress to challenge the containment operational capability. Using IMC 0609, A ttachment 4, Initial Characterization of Findings, the finding was determined to affect the Barrier Integrity Cornerstone because it involved ISI program examinations designed to identify degradation of the containment metal liner. The inspectors screen ed the finding using IMC 0609, Appendix A, The Significance Determination Process (SDP) For Findings At -Power, Exhibit 3 Barrier Integrity Screening Questions, and determined that the finding was of very low safety significance (Green) because it did not represent an actual open pathway in the physical integrity of the containment. The inspectors reviewed this performance deficiency for cross -cutting aspects as required by IMC 0310, Components With Cross -Cutting Aspects. The finding was determined to be reflective of present licensee performance because in 2014, the licensee did not take effective corrective actions to implement the ASME BPVC 3 requirements in the Subsection IWE P rogram , when a reasonable opportunity was available through the review of NRC Information Notice (IN) 2014- 07, which highlighted this industry -wide problem. Therefore, the finding was assigned a cross - cutting aspect in the resolution component of the problem identification and resolution cross -cutting area (P.3) |
| 05000261/FIN-2017001-02 | 2017Q1 | Robinson | Failure to Submit Complete and Accurate Information for a Requested License Amendment | Severity Level IV. An NRC -identified severity level IV (SL IV) NCV of 10 CFR 50.9(a), Completeness and Accuracy of Information, was identified for the licensees failure to provide complete and accurate information in a license amendment request (LAR), dated November 19, 2015, requesting extension of the containment leak rate test frequencies required by various containment technical specifications (TS s). In this LAR, the licensee incorrectly stated that they had revised their ASME BPVC, Section XI, Subsection IWE program to include visual examinations of the test connections in the leak -chase channel penetration pressurization system ( PPS) , when in fact, the program had not been revised and the examinations had not been performed . This information was material to the NRC because it was used, in part, as the basis for the approval and issuance of License Amendment 247, dated October 11, 2016, extending the TS containment leak rate test frequencies. The licensees corrective actions included conducting the visual examinations of the test connections in the leak -chase channel PPS during the ongoing refueling outage in March 2017 and initiating actions to add the visual examination requirements to their Subsection IWE program. This issue was entered into the licensees CAP as NCR 02110516. The failure to provide complete and accurate information in accordance with 10 CFR 50.9(a) for the LAR associated with License Amendment 247 is a violation of NRC requirements . This violation was screened against the ROP guidance in IMC 0612, Appendix B, Issue Screening, and no associated ROP finding was identified. The inspectors evaluated this issue using the Traditional Enforcement process because it had the potential to impact the NRCs ability to perform its regulatory function. Specifically, the violation impacted the regulatory process, in that the inaccurate information was material to the NRCs review and acceptance of licensee actions to address the industry -wide operating experience discussed in NRC IN 2014- 07. Based on licensee inaccurate information that they had addressed IN 2014 -07 by revising their containment ISI program to perform visual inspections of accessible tubing in the containment leak -chase channel PPS system, the NRC staff concluded that the licensee was properly implementing the ASME BPVC, Section XI, Subsection IWE program. In accordance with the guidance in Sections 2.2 and 6.9 of the NRC Enforcement Policy, the inspectors determined this is an SL IV violation, because had the information been complete and accurate at the time provided, it likely would have resulted in the need for further clarification of the licensees actions to address NRC IN 2014- 07 , but would not have caused the NRC to change its decision to issue the license amendment or resulted in substantial further inquiry . Also, on March 23, 2017, the licensee completed the visual examinations of the subject tubing in the leak -chase channel system and did not identify any significant degradation. In accordance with IMC 0612, Appendix B, traditional enforcement issues are not assigned a cross -cutting aspect. |
| 05000261/FIN-2017001-03 | 2017Q1 | Robinson | Licensee-Identified Violation | 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, states, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and non- confor mances are promptly identified. Contrary to the above, in March 2014, while performing examinations in steam generator C during forced shutdown RFO229F3, the licensee failed to identify a loose part lodged in contact with tube R37C22. The licensee identified the loose part in March 2017 during refueling outage RO30. The licensee verified that indications of the part were detectable during RFO229F3, retrieved the part, verified that degradation caused by the part met all structural integrity requirement s, plugged the tube, and removed it from service. This issue was identified in the licensees CAP as NCR 0210725. The inspectors evaluated this violation using IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, and 0609 Appendix A, The Significance Determination Process (SDP) for Findings At -Power, and determined that the violation was of very low safety significance (Green) because evaluations demonstrated that the tube could sustain three times the differential pressure across it during normal full power steady state operation and that the steam generator did not violate the accident leakage performance criterion |
| 05000324/FIN-2013004-01 | 2013Q3 | Brunswick | Failure to Identify and Correct Nuclear Service Water Pump Shaft Degradation | An NRC identified Green non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, was identified for the failure of the licensee to identify and correct a condition adverse to quality (CAQ) on the 1B nuclear service water pump (NSWP). Specifically, between June 26, 2012, and January 12, 2013, the licensee failed to identify or correct the pump shaft degradation on the 1B Nuclear Service Water Pump (NSWP) pump. This resulted in the shaft bearing delaminating and bearing material becoming dislodged and trapped in the pump strainer which caused the 1B NSWP to become inoperable. The licensee replaced the pump shaft and returned the pump to operable. The licensee entered this issue into the corrective action program (CAP) as nuclear condition report (NCR) 582584. The inspectors determined that the failure of the licensee to identify and correct the 1B NSWP shaft degradation before the pump failed was a performance deficiency. The finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the shaft degradation resulted in the 1B NSWP being inoperable. Using IMC 0609, Appendix A, issued June 19, 2012, the SDP for Findings At-Power, the inspectors determined the finding was of very low safety significance (Green) because the finding did not affect the design or qualification of a mitigating structure, system and component (SSC), the finding did not represent a loss of system and/or function, the finding did not represent an actual loss of a function of a single train for greater than the technical specifications (TS) allowed outage time, the finding did not represent an actual loss of a function of one or more non-TS trains of equipment, and did not screen as potentially risk-significant due to a seismic, flooding, or severe weather initiating event. The finding has a cross-cutting aspect in the area of problem identification and resolution associated with the CAP attribute because the licensee failed to implement a CAP with a low threshold for identifying issues, specifically the licensee did not enter this issue into the CAP in June 2012. |
| 05000324/FIN-2013004-02 | 2013Q3 | Brunswick | Inadequate Preventative Maintenance Procedure for the Service Water Pump Breakers | A self-revealing Green NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified for the failure of the licensee to have an adequate preventative maintenance procedure for the service water pump breakers. Specifically, from December 1, 2004, through the end of this inspection period (September 30, 2013), the licensee failed to have an adequate preventative maintenance procedure to ensure the 52S mechanism was securely bolted to the breaker for the 2C conventional service water pump (CSWP). This resulted in both discharge valves failing to open when the 2C CSWP was started, and the inoperability of the 2C CSWP. The licensee securely bolted and tightened the 52S mechanism to the breaker. The licensee entered this issue into the CAP as NCR 604452. The inspectors determined the failure to have an adequate preventative maintenance procedure for the service water pump breakers was a performance deficiency. The finding was more than minor because it was associated with the configuration control attribute of the Mitigating Systems Cornerstone and affects the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to ensure the 52S mechanism was securely bolted to the 2C CSWP breaker resulted in the failure of both 2C CSWP discharge valves to open. Using IMC 0609, Appendix A, issued June 19, 2012, the SDP for Findings At-Power, the inspectors determined the finding was of very low safety significance (Green) because the finding did not affect the design or qualification of a mitigating SSC, the finding did not represent a loss of system and/or function, the finding did not represent an actual loss of a function of a single train for greater than the TS allowed outage time, the finding did not represent an actual loss of a function of one or more non-TS trains of equipment, and did not screen as potentially risk-significant due to a seismic, flooding, or severe weather initiating event. The finding does not have a cross-cutting aspect since the performance deficiency is not indicative of current plant performance. The 2C CSWP breaker was refurbished in December 2004 and installed in the plant in January 2005. |
| 05000338/FIN-2018003-01 | 2018Q3 | North Anna | Licensee-Identified Violation | This violation of very low safety significance 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 Section 2.3.2.a of the Enforcement Policy. Violation: TS 5.4.1.a, requires in part, that written procedures shall be established per Revision 2 of Regulatory Guide 1.33, Appendix A, of which part 9.a requires written procedures and documented instructions appropriate to the circumstances for performing maintenance that can affect the performance of safety related equipment. Contrary to the above, on June 12, 2018, the licensee failed to adequately establish a procedure appropriate to the circumstances during maintenance on the safety-related main control chillers. Specifically, licensee mechanical preventative maintenance procedure, 0-MPM-0806-02, Inspection of Control Room Chillers, Revision 0, did not provide a proper method to adequately monitor the Freon level in main control room chillers. Consequently, the licensee discovered a low Freon level condition on main control room chiller 1-HV-3-4B, which rendered the chiller inoperable. Significance: The inspectors reviewed Exhibit 2 Mitigating Systems Screening Questions of IMC 0609 Appendix A, The Significance Determination Process (SDP) for findings at Power and determined this finding was of very low safety significance, Green, because there was no design deficiency, it did not represent a loss of system or function, and did not represent an actual loss of function for greater than its TS allowed outage time. Corrective Action Reference: CR109958 |
| 05000369/FIN-2000009-03 | 2000Q4 | Mcguire McGuire | Availability of the Charging Pumps for Fire Damage to the Volume Control Tank Outlet Valve | Fire Areas 2 and 14 contained redundant trains of systems, equipment, and cables necessary to accomplish post-fire SSD conditions. In the event of an unmitigated fire in either of these areas the licensees SSA credited the use of an alternative shutdown capability designated as the SSS. The SSS was comprised of existing plant safety related systems as well as certain dedicated equipment that would be used in the event of a fire which required shutdown of the unit from the SSF. Section III.L.3 of Appendix R to 10 CFR 50 requires, in part, that the alternative shutdown capability be physically and electrically independent of the fire area under evaluation. The team assessed the adequacy of electrical independence provided for the alternative shutdown capability (i.e., SSS) in the event of a fire in Fire Area 2 or Fire Area 14. The SSS relied on the use of the turbine driven CA pump to accomplish the decay heat removal shutdown function. Hence, the routing of power and control cables associated with a sample of components required to assure the operability of the turbine driven CA pump was reviewed. This review determined that certain cables associated with the turbine driven CA pump suction valve (Valve 1CA7AC) could be subject to damage as a result of a fire in either of these areas. Specific cables included: 1*CA516, 1*CA517, 1*CA519, 1*CA761, and 1*CA763. Fire damage to these cables had the potential to cause valve 1CA7AC to fail closed. Should this occur, the turbine driven CA pump would be damaged in a short period of time due to a loss of pump suction. Although this scenario had been identified by the licensee, operator actions credited in the SSA to mitigate this event were not appropriate. Specifically, the team determined that the licensees credited recovery actions in the SSA could not be completed in a sufficiently timely manner necessary to prevent pump damage. The operator actions in the SSA had not been translated to appropriate operations procedures (e.g., AP/1/A/5500/24). Additionally, local manual actions to reopen valve 1CA7AC would require an operator to traverse Fire Area 2. This issue will be tracked as Unresolved Item (URI) 50-369, 370/00-09-01: Potential for Loss of Auxiliary Feedwater Flow for an Appendix R Fire in Fire Areas 2 or 14. The team noted that certain byproduct associated circuits issues (e.g., fire-induced spurious operations or mal-operations) are the subject of an ongoing, voluntary industry initiative. This URI is considered an example of a \"byproduct\" associated circuits issue and will be tracked as a URI pending generic resolution of the related issue. This issue was entered in the licensees corrective action program as Problem Investigation Process (PIP) No. M-00-04480. |
| 05000369/FIN-2011004-01 | 2011Q3 | Mcguire McGuire | Failure to Establish Adequate ND Venting Procedures | The inspectors identified a NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for the failure to establish acceptance criteria to determine operability in surveillance procedures used to vent the decay heat removal system in Modes 5, 6, and No-Mode in preparation for Mode 6. The issue was entered into the licensees corrective action program as PIP M-11-04745 the licensees failure to establish adequate acceptance criteria for ND venting surveillance procedures PT/1/A/4200/036 and PT/2/A/4200/036 was a performance deficiency. The PD was determined to be more than minor because if left uncorrected, the failure to establish acceptance criteria for surveillance tests which establish the basis for the ND system operability in modes 5 and 6 would have the potential to lead to a more significant safety concern in that conditions which could impact system operability could remain undetected. In addition, the finding adversely affected the equipment performance attribute of the mitigating systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using IMC 0609, Appendix G, Shutdown Operations Significance Determination Process, Attachment 1, the finding was determined to be of very low safety significance (Green) because a quantitative assessment was not required based on the criteria in Attachment 1. The finding had a cross-cutting aspect of implementation of operating experience in the Operating Experience component in the area of Problem Identification and Resolution because the licensee failed to implement operating experience from Generic Letter (GL) 2008-01 into station procedures |
| 05000369/FIN-2011004-02 | 2011Q3 | Mcguire McGuire | Licensee-Identified Violation | 10 CFR 50, Appendix B, Criterion V, required, in part, that procedures for performing maintenance that can affect the performance of safety-related equipment should be properly preplanned and performed in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances. This requirement included written procedures for post-maintenance testing of the RN system. Contrary to the above, written procedures for post-maintenance testing of the RN system were not appropriate to the circumstances. On July 24, 2010, following maintenance on the 2A RN strainer dP instrument loop, the licensee failed to include adequate post-maintenance instructions which would have included testing the 2A RN strainer dP instrument loop during high flow conditions experienced on the RN SNSWP supply and return header. This violation was determined not to be greater than very low safety significance (Green) because it did not represent a loss of safety function of the 2A RN train. This condition was placed in the licensees corrective action program as PIP M-10-05982 |
| 05000369/FIN-2011004-03 | 2011Q3 | Mcguire McGuire | Availability of the Charging Pumps for Fire Damage to the Volume Control Tank Outlet Valves | McGuire Unit 1 Operating License Condition 2.C.(4) and Unit 2 Operating License Condition 2.C.(4) required the licensee to implement and maintain in effect all provisions of the approved Fire Protection Program as described in the Final Safety Analysis Report, as updated, for the facility and as approved in the NRC Safety Evaluation Report dated March 1978, and Supplements 2, 5, and 6 dated March 1979, April 1981, and February 1983, respectively, and the NRC Safety Evaluation Report dated May 15, 1989. The licensees approved fire protection program committed to 10 CFR 50, Appendix R, Section III.G. Section III.G specified that fire protection features shall be capable of limiting fire damage so that systems necessary to achieve and maintain cold shutdown from either the control room or emergency shutdown stations can be repaired within 72 hours. Contrary to the above, since the Safety Evaluation Report issued in 1978, the licensee failed to have fire protection features capable of limiting fire damage so that systems necessary to achieve and maintain cold shutdown conditions could be repaired within 72 hours. These conditions have existed since original plant construction and were applicable to both Unit 1 and Unit 2. The NRC is exercising enforcement discretion for this nonconformance in accordance with the NRC Enforcement Policy, Interim Enforcement Policy Regarding Enforcement Discretion for Certain Fire Protection Issues (10 CFR 50.48), because the licensee documented their commitment to adopt NFPA 805 and change their fire protection licensing basis to comply with 10 CFR 50.48(c) prior to December 31, 2005, and it was likely this issue would have been identified and addressed during the licensees transition to NFPA 805, it was entered into the licensees corrective action program, immediate corrective action and compensatory measures were taken, it was not likely to have been previously identified by routine licensee efforts, it was not willful, and it was not associated with a finding of high safety significance. |
| 05000369/FIN-2012002-01 | 2012Q1 | Mcguire McGuire | Faulure to Maintain Operable Fire Assembly in Unit 2 Auxiliary Feedwater Pump Room | An NRC-identified non-cited violation (NCV) of Technical Specification (TS) 5.4.1.d was identified for failure to maintain an operable fire assembly resulting in an unsealed pipe penetration through a 3-hour rated fire barrier wall separating the Unit 2 Train A/B motor driven auxiliary feedwater pump room from the Unit 2 mechanical penetration equipment room. The licensee reinstalled pipe caps on each end of the unsealed pipe. The performance deficiency (PD) was more than minor because it was associated with the protection against external events attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective in that the unsealed opening adversely impacted the ability of the fire barrier to perform its intended safety function. The finding was of very low safety significance because the fire barrier deficiency represented a low fire degradation rating. The finding was directly related to the cross-cutting area of Human Performance under the Procedural Compliance aspect of the Work Practices component because station personnel failed to follow fire protection impairment procedures for breaching a fire assembly. |
| 05000369/FIN-2012002-02 | 2012Q1 | Mcguire McGuire | Failure to Enter Condition Adverse to Quality Into the CAP | A NRC-identified finding was identified for the failure to follow the sites corrective action program (CAP) procedure which required the initiation of a PIP for a degraded 2B emergency diesel generator (EDG) Bellofram seal. The degraded Bellofram seal contributed to the improper setup of the 2B EDG governor actuator which resulted in the 2B EDG not achieving the required 105 percent full power output. The performance deficiency was more than minor because it was associated with the human performance attribute of the Mitigating Systems cornerstone and adversely impacted the cornerstone objective in that the capability of the EDG to provide continuous and adequate load margin was affected. The finding was of very low safety significance because it did not represent an actual loss of safety function of the system or train. The finding was directly related to the cross-cutting aspect of implements the CAP with a low threshold in the Corrective Action Program component in the area of the Problem Identification and Resolution because the licensee did not enter the condition into the CAP. |
| 05000369/FIN-2012003-01 | 2012Q2 | Mcguire McGuire | Failure to Implement Planned Compensatory Measures for Impaired Auxiliary Building Fire Hose Stations | An NRC-identified non-cited violation (NCV) of Technical Specification (TS) 5.4.1.d was identified for failure to implement adequate compensatory measures for multiple impaired manual fire hose stations (FHSs) in accordance with the approved fire protection program. Gated wye valves were not installed as required during a periodic flush of multiple auxiliary building (AB) FHSs rendering them inoperable. The licensee took actions to install the gated wye valves in the affected FHSs to restore them to operable. This violation was entered into the licensees corrective action program (CAP) as Problem Investigation Program (PIP) M-12-2816. The performance deficiency (PD) was more than minor because it was associated with the protection against external events attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective in that manual fire suppression capability was impaired. The finding was determined to be of very low safety significance because it represented a low degradation of the manual fire suppression function. The cause of this finding was directly related to the cross-cutting aspect of planning and coordination of work activities in the Work Control component of the Human Performance area, in that the licensee did not plan and coordinate work activities to ensure that adequate compensatory measures were established for impaired fire hose stations. |
| 05000369/FIN-2012003-02 | 2012Q2 | Mcguire McGuire | Licensee-Identified Violation | TS 3.6.10 required two AVS trains to be operable in Modes 1, 2, 3 and 4. TS 3.6.10, Condition B, specified if one or more AVS train pre-heaters was inoperable, the pre-heater must be restored to operable status within 7 days or a report must be initiated within 7 days, detailing the reason for the pre-heater inoperability and corrective actions to restore the pre-heater, and submitted to the NRC within 30 days. TS 3.6.10, Condition C, specified if the required actions and completion times of Condition B were not met, the plant shall be shut down to Mode 3 within 6 hours and Mode 5 within 36 hours. Contrary to the above, from October 29, 2011, to December 10, 2011, the 2B AVS pre-heater was not operable and the licensee failed to meet either Condition B or Condition C. This violation is not greater than Green because the degraded pre-heater did not result in a loss of safety function of the 2B AVS. This violation was documented in the licensees corrective action program as PIP M-11-9216. |
| 05000369/FIN-2012004-01 | 2012Q3 | Mcguire McGuire | Failure to Correctly Implement Technical Specifications Adversely Affects Requalification Operating Test Quality | An NRC-identified finding was identified associated with the quality of the simulator scenarios developed by the licensee for the licensed operator requalification annual operating test. The licensee failed to follow the Technical Specification (TS) rules of usage for concurrent inoperability as shown in TS Example 1.3-3. The licensee entered this issue into their corrective action program (CAP) as PIP M-12-4157. The performance deficiency (PD) was determined to be more than minor because it was associated with the Human Performance attribute of the Mitigating Systems Cornerstone, and adversely affected the cornerstone objective in that it impacted the licensees ability to evaluate and ensure operator performance. The significance determination was performed in accordance with Manual Chapter 0609, Appendix I, and determined to be of very low safety significance (Green). The cause of the finding was directly related to the crosscutting aspect of personnel training and qualifications in the Resources component of the cross-cutting area of Human Performance, in that the licensee failed to ensure the quality of the operating tests used to evaluate the knowledge, skills, abilities, and training provided to operators to assure nuclear safety. |
| 05000369/FIN-2012005-01 | 2012Q4 | Mcguire McGuire | Failure to Maintain Complete and Accurate Pre-Fire Plans | An NRC-identified Green non-cited violation (NCV) of the Unit 2 Facility Operating License, Condition 2.C.4, Fire Protection Program, was identified for failure to maintain prefire plans in areas that contain safety-related equipment. The inspectors identified that all copies of fire strategy plan view for the Unit 2 lower annulus and containment were missing from their pre-fire plans and unavailable to the Fire Brigade Leader and Operations personnel in the event of a fire in the Unit 2 reactor building. Corrective actions included replacement of the missing fire strategy plan views and additional review of the fire strategy books located in the Fire Brigade Leaders Kit, Control Room, and Emergency Preparedness office. This violation was entered into the licensees corrective action program (CAP) as Problem Investigation Program (PIP) M-12-08270. The performance deficiency (PD) was more than minor because it was associated with the Mitigating Systems cornerstone attribute of Protection Against External Events (Fire) and adversely affected the cornerstone objective, in that, it degraded the manual fire suppression capability. The finding was determined to be of very low safety significance (Green) because the fire brigade consisted of plant personnel familiar with the plant layout and associated fire hazards and appropriate fire-fighting equipment was available. The cause of the PD was directly related to the aspect of complete, accurate, and up-to-date procedures of the Resources Component in the cross-cutting area of Human Performance because the Fire Brigade Program Administrator failed to include all approved plan view updates into the fire brigade response strategies. |
| 05000369/FIN-2012005-02 | 2012Q4 | Mcguire McGuire | Evaluation of the Occupational Radiation Dose Assigned to a Worker from a Piece of Contaminated Wire | While working in the reactor building an individual received a puncture wound in their hand from a piece of contaminated wire. Licensee attempts to decontaminate the wound were unsuccessful and the radioactive material from the contaminated wire remained inside the individuals hand. The licensee was reviewing that data and determining what dose to assign to the individual. The NRC will review the methodologies used once the licensee has completed its assessment to determine if a violation of regulatory requirements existed. This issue is identified as URI 05000369,370/2012005-02, Evaluation of the Occupational Radiation Dose Assigned to a Worker from a Piece of Contaminated Wire. |
| 05000369/FIN-2012005-03 | 2012Q4 | Mcguire McGuire | Licensee-Identified Violation | Technical Specification 5.7, High Radiation Area, required areas with radiation levels greater than 1,000 millirem (mrem) per hour at 30 centimeters (cm) from the radiation source or from any surface which the radiation penetrates to be provided with locked or continuously guarded doors to prevent unauthorized entry. Contrary to the above, on September 23, 2011, an area with radiation levels greater than 1,000 mrem per hour at 30 cm from the radiation source or from any surface which the radiation penetrates was not locked or continuously guarded to prevent unauthorized entry. The locking method for a LHRA door leading to the reactor head stand did not prevent unauthorized entry. The padlock used to secure retaining bolts on the doors was supposed to be installed through openings in the bolts preventing them from being removed. Instead, the padlock was installed around the bolts allowing them to be removed. Corrective actions included identifying other HRA, LHRA, and VHRA barriers with the unique locking mechanism, photographing the proper locking method, providing proper instructions to individuals during key issuance, and clarifying procedural guidance on the proper use of the locking mechanism. The corrective actions were documented under PIP M-11-07009. The violation was evaluated using the Occupational Radiation Safety Significance Determination Process and was determined to be not more than very low safety significance (Green) because this finding did not have a substantial potential for over-exposure because of additional controls and warnings present such as personal ED alarming devices and LHRA posting. |
| 05000369/FIN-2013002-01 | 2013Q1 | Mcguire McGuire | Failure to Revise Turbine Inlet Pressure Calibration Procedures During Implementation of High Pressure Turbine Replacement Design Modific | A self-revealing finding was identified for the licensees failure to follow the requirements of the station modification program manual EDM 601 during implementation of a high pressure turbine replacement modification revision. This resulted in Anticipated Transient Without Scram Mitigation System Actuation Circuitry (AMSAC) calibration procedures not being revised with the proper setpoints. The performance deficiency (PD) was more than minor because it affected the Design Control attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective in that AMSAC actuated causing a turbine trip. The finding was determined to have very low safety significance because it did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available. The cause of this finding was related to the cross-cutting aspect of the need for work groups to maintain appropriate interfaces and communicate, coordinate with each other during important work activities as described in the Work Control component of the Human Performance cross-cutting area because necessary revisions to the AMSAC input device calibration procedures were not adequately communicated. |
| 05000369/FIN-2013002-02 | 2013Q1 | Mcguire McGuire | Licensee-Identified Violation | TS 3.6.3 required that each containment isolation valve be operable in Modes 1, 2, 3, and 4. TS 3.6.3, Condition A, specified if one containment isolation valve is inoperable, the flow path must be isolated within 4 hours and verified isolated once per 31 days. Contrary to the above, from November 2, 2012, to November 4, 2012, with Unit 2 in Mode 4, manual containment isolation valve 2NV-1053 was inoperable and the licensee failed to isolate the flow path within 4 hours. This violation was determined to be of very low safety significance (Green) due to the small size of the piping and that a control room air-operated valve (i.e., 2NV-840) located downstream of 2NV-1053 could have been used to isolate the penetration. This violation was documented in the licensees CAP as PIP M-12-09347. |
| 05000369/FIN-2013003-01 | 2013Q2 | Mcguire McGuire | Failure to Implement Adequate Venting Instructions for Condensate Booster Pump Trip Instrumentation Resulting in Reactor Trip | A self-revealing finding was identified for the licensees failure to implement adequate instructions for venting condensate booster pump (CBP) emergency low suction pressure trip instrumentation which resulted in air entrainment causing a non-conservative shift in the trip setpoint. During a subsequent secondary side transient involving a heater drain tank pump trip, the non-conservative trip setpoint resulted in a premature trip of all three CBPs ultimately causing a reactor trip. The performance deficiency was more than minor because it affected the Procedure Quality attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective, in that, the inadequate venting allowed air entrainment in the instrumentation lines resulting in a reactor trip. This finding was determined to have very low safety significance (Green) because it did not contribute to the likelihood of both a reactor trip and that mitigation equipment or functions would not be available. No cross cutting aspect was identified. |
| 05000369/FIN-2013502-01 | 2013Q4 | McGuire | Licensee-Identified Violation | The following violation of very low safety significance (Green) was identified by the licensee and is a violation of NRC requirements which met the criteria of the NRC Enforcement Policy for being disposition as a NCV.TS 3.3.2, Function 6.f, requires that all four instrumentation channels of the TDCA pump suction transfer function be operable in Modes 1, 2, and 3. TS 3.3.2, Condition N, specifies if one or more of the pressure switch instrumentation channels are inoperable, the channel must be restored to operability within 48 hours or the associated TDCA pump must be declared inoperable. Contrary to this requirement, from September 1993 to May 30, 2013, the channel associated with pressure switch 1CAPS5390 was inoperable and the licensee failed to declare the Unit 1 TDCA inoperable within the required TS completion time. This violation was determined to be of very low safety significance (Green) because the channel would still have been capable of actuating and aligning the TDCA to its assured water source within the timeframe necessary for the pump to perform its intended safety function. This violation was documented in the licensees CAP as PIP M-13-05935. |
| 05000369/FIN-2014002-02 | 2014Q1 | Mcguire McGuire | Failure to Adequately Control the Use of Self- Extinguishing Fire Lids | An NRC-identified NCV of the McGuire Unit 1 and Unit 2 Renewed Facility Operating License Condition 2.C.4, FPP, was identified for the licensees failure to adequately control the storage of transient combustibles in waste receptacles equipped with self-extinguishing fire lids in accordance with the FPP requirements. The licensee took actions to correct all waste receptacles in the plant that were filled beyond the manufacturers specification or had loosely fitted lids. This condition was placed in the licensees corrective action program. The licensees failure to control the storage of transient combustibles in accordance with the requirements of NSD-313 was more than minor because it was associated with the Mitigating Systems cornerstone attribute of Protection Against External Factors (Fire) and adversely affected the cornerstone objective in that the self-extinguishing function was not retained which could allow the spread of the fire and adversely affect mitigating system equipment in the area. The finding was determined to be of very low safety significance (Green) because it did not affect the ability of the reactor to reach and maintain cold shutdown conditions. A cross-cutting aspect was not assigned because the performance deficiency does not reflect current licensee performance. |
| 05000369/FIN-2014002-03 | 2014Q1 | McGuire | Failure to Implement Adequate Design Control Measures for Rod Control Power Supply Replacement Resulting in Reactor Trip | A self-revealing finding (FIN) was identified for the licensees failure to implement adequate design control measures for the rod control power supply modification which resulted in the loss of 24VDC power in the 1AC rod control power cabinet. The inspectors determined that the licensees failure to implement adequate design control measures was more than minor because it affected the Design Control attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective, in that, the insufficient margin in the rod control power supply OVP function caused a multiple drop rod event which resulted in a reactor trip. This finding was determined to have very low safety significance (Green) because it did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available. A cross-cutting aspect was not assigned because the performance deficiency does not reflect current licensee performance. |
| 05000369/FIN-2014003-01 | 2014Q2 | Mcguire McGuire | Licensee-Identified Violation | 10 CFR Part 50, Appendix B, Criterion XVI, requires, in part, the licensee to establish measures 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 the above, in March 2008 and September 2012, the licensee failed to promptly identify a condition adverse to quality. A flaw in the Cold Leg 2D Nozzle 4-1 weld was missed during UT examinations of this component. During the most recent outage, the licensee reexamined this weld and identified an 85 percent through-wall flaw originating from the inner diameter in the weld. Destructive testing and analysis established that this flaw most likely existed since 2005. This violation was determined not to be greater than very low safety significance (Green) because it could not result in exceeding the RCS leak rate for a small loss of coolant accident (LOCA) and could not have likely affected other systems used to mitigate a LOCA. This violation was documented in PIP M-14-03544. |
| 05000369/FIN-2014003-02 | 2014Q2 | Mcguire McGuire | Licensee-Identified Violation | TS 5.4.1.a requires that written procedures shall be established, implemented, and maintained as recommended in Regulatory Guide (RG) 1.33, Rev. 2, Appendix A, February 1978. Section 3.d of RG 1.33 recommends that appropriate procedures be prepared for energizing, filling, venting, draining, startup, shutdown, and changing modes of operation for the ECCS. Contrary to the above, the licensee failed to develop an adequate procedure to fill and vent the Unit 2 NV system following system draining and argon gas introduction associated with NV 2A mixed bed demineralizer valve modification work conducted December 1-12, 2013. On December 14, 2013, following this modification work, a large gas void was identified in the ECCS piping at high point vent valve 2NV-1056, located in the suction of the ECCS pumps during design basis accident conditions involving cold-leg recirculation. The licensee determined that the use of an inadequate fill and vent procedure during the system restoration from the modification work resulted in the accumulation of a significant amount of the gas at this location. This violation was determined to be of very low safety significance (Green) because the licensee provided reasonable evidence that the ECCS pumps would have been capable of performing their intended safety function had the gas void been ingested into the suction of the pumps. This violation was documented in the licensees CAP as PIP M-13-11181. |
| 05000369/FIN-2014004-01 | 2014Q3 | McGuire | 1B/1C Reactor Coolant System Loop Safety Injection Piping Flaws | The licensee identified flaws with ultrasonic testing in the 1B and 1C cold leg safety injection pipe welds as part of their extent of condition from Unit 2 for MRP- 146, Thermal Fatigue. Further evaluation determined these flaws were a circumferential flaw with an axial component on the nozzle side for 1B and an axial flaw from the centerline of the weld into the base metal for 1C. The licensee completed examinations on all welds included in the MRP-146 program and found them to be within the acceptance criteria. The licensee also removed and repaired the 1B and 1C nozzles. Welding of the new components have been examined and have passed all quality assurance examinations. The licensee determined that the flaws were a result of thermal fatigue. The licensee has performed all required examinations and repairs and is completing a metallurgical analysis of the flaws. This is an unresolved item pending review of the licensees metallurgical analysis of the flaws to determine if there is a performance deficiency. This issue will be tracked as URI 05000369/2014004-01, 1B/1C Reactor Coolant System Loop Safety Injection Piping Flaws. |
| 05000369/FIN-2014004-02 | 2014Q3 | McGuire | Review NOED 14-2-002 Granting Exercise of Enforcement Discretion to Complete 1B EDG Repairs | The inspectors reviewed NOED 14-2-002 and related documents to determine the accuracy and consistency with the licensees assertions and implementation of the licensees compensatory measures and commitments. The inspectors independently verified the proper implementation of these compensatory measures which included deferring non-essential surveillances and other maintenance activities on the 1A EDG, TDCA pump, SSF, switchyard, and posting dedicated fire watches in selected risk significant areas. Additional inspection of this issue will be conducted as part of the NRCs review of the subsequent Licensee Event Report (LER) to be submitted by the licensee within 90 days. This LER will describe the circumstances of the 1B EDG failure, the root cause, and planned licensee corrective actions. This URI is identified as URI 05000369/2014004-02, Review NOED 14-2-02 Granting Exercise of Enforcement Discretion to Complete 1B EDG Repairs. |
| 05000369/FIN-2014005-01 | 2014Q4 | Mcguire McGuire | Failure to Adequately Control Transient Combustible Materials and Ignition Sources in Accordance with the Fire Protection Program | An NRC-identified Green NCV of the McGuire Unit 1 and Unit 2 Renewed Facility Operating License Condition 2.C.4, Fire Protection Program (FPP), was identified for the licensees failure to adequately control fire ignition sources in the Unit 1 and Unit 2 exterior doghouses in accordance with the FPP requirements of Nuclear System Directive (NSD)- 313, Control of Transient Fire Loads. Specifically, temporary electric portable heaters were energized for several days without implementing required hourly fire watches, locating the energized heaters greater than prescribed separation distances from safety-related equipment, and preventing other transient combustible materials from being located near the heaters. The licensee placed this issue into their corrective action program (CAP) and took corrective actions to de-energize the heaters, distance the heaters away from safety related feedwater isolation valve electrical cables, and remove unnecessary transient combustibles from the area. The failure to control fire ignition sources in accordance with NSD-313 was a performance deficiency (PD) . The PD was more than minor because it was associated with the mitigating systems cornerstone attribute of protection against external factors (fire) and adversely affected the cornerstone objective in that, a fire could have affected nearby safety-related feedwater isolation valve electrical cables which provide a shutdown mitigation function. The finding was determined to be of very low safety significance (Green) because it did not affect the ability of the reactor to reach and maintain cold shutdown condition. This finding had a cross cutting aspect of teamwork in the human performance area because individuals failed to effectively communicate and coordinate their activities to ensure that the temporary heaters were energized following prescribed fire protection control measures and written instructions. |
| 05000369/FIN-2014005-02 | 2014Q4 | McGuire | Failure to Adequately Implement Containment Closeout Resulting in Loose Debris and Unanalyzed Materials Left in Containment | An NRC-identified Green NCV of Technical Specification 5.4.1.a, Procedures, was identified for the failure to properly implement containment cleanliness and material control closeout procedures in accordance with procedure PT/1A/4600/003F, Containment Cleanliness and ECCS Operability Inspection, prior to entering Mode 4, following the Unit 1 refueling outage. Specifically, a large amount of unanalyzed general loose debris, as well as scaffolding with aluminum walkboards and fibrous lead blankets, were left in containment that could either contribute to emergency core cooling system (ECCS) recirculation sump screen blockage or containment hydrogen generation during design basis accidents. The licensee placed this issue into their CAP and took corrective actions to remove the loose debris and unanalyzed materials and performed re-inspections of containment to identify any additional loose debris or unanalyzed materials left in containment. The failure to perform an adequate containment cleanliness and material control closeout following the Unit 1 refueling outage in accordance with procedure PT/1/A/4600/003F was a PD. The PD was more than minor because it was associated with the equipment performance attribute of the mitigating systems cornerstone, and adversely affected the cornerstone objective in that, loose debris in containment could result in the debris being transported to the ECCS recirculation sump screens in the event of design basis accident and adversely affect the sump performance. In addition, the PD was associated with the configuration control attribute of the barrier integrity cornerstone and adversely affected the cornerstone objective in that, the failure to control scaffolding that contained unanalyzed amounts of aluminum in containment challenged the existing analysis for containment aluminum inventory limitations. The finding was determined to be of very low safety significance (Green) because it did not result in an actual loss of safety function of the ECCS sumps, was not safety significant due to external events, and no actual open pathway in the physical integrity of containment occurred. The finding had a cross-cutting aspect of field presence in the human performance area because the licensee failed to ensure that adequate supervisory and management oversight of the containment closeout process was conducted to ensure proper performance of procedure PT/1/A/4600/003F prior to entering Mode 4. |
| 05000369/FIN-2015002-01 | 2015Q2 | Mcguire McGuire | Failure to Establish Compensatory Actions for Obstructed Fire Sprinkler Spray Nozzle | An NRC-identified Green NCV of Technical Specification (TS) 5.4.1.d, Procedures, was identified for failure to evaluate and establish adequate compensatory measures for an impaired fire protection automatic water sprinkler system. Specifically, a solid deck scaffold platform was erected below a sprinkler system spray nozzle that would have obstructed the nozzle spray pattern protecting safe shutdown equipment involving the 2B2 component cooling water pump/motor. The licensee entered the issue into the corrective action program (CAP) as nuclear condition report (NCR) 01931412 and implemented immediate corrective actions to remove the scaffolding obstructing the sprinkler nozzle. The failure to evaluate scaffolding obstruction of a sprinkler system spray nozzle and implement required fire protection compensatory actions was a performance deficiency (PD). The PD was more than minor because it was associated with the mitigating systems cornerstone attribute of protection against external factors (fire) and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to provide adequate compensatory actions for an obstructed sprinkler nozzle would have reduced the licensees ability to quickly extinguish fires in the area. The finding was screened in accordance with NRC IMC 0609, Significance Determination Process, Attachment 4, Initial Characterization of Findings. Using the guidance in IMC 0609, Appendix F, Attachment 1, Fire Protection SDP Phase 1 Worksheet, the finding was assigned a category of fixed fire protection systems. The inspectors determined the finding to be of very low safety significance (Green), because it was assigned a low degradation rating that was based upon meeting the criteria described in IMC 0609, Appendix F, Attachment 2, Degradation Rating Guidance Specific to Various Fire Protection Program Elements. Specifically, less than ten percent of the sprinkler nozzles were nonfunctional, there were functional nozzles within five feet of the combustibles of concern, and the system was nominally code compliant. The finding had a cross-cutting aspect of procedure adherence in the human performance area, because the licensee failed to follow scaffolding erection procedures which explicitly required not erecting scaffolding that could obstruct sprinkler nozzles unless approved by a fire protection engineer and necessary compensatory actions were implemented. |
| 05000369/FIN-2015003-01 | 2015Q3 | Mcguire McGuire | Failure to Adequately Implement a Temporary Modification for a Leak Enclosure | A self-revealing Green finding (FIN) was identified for failure to adequately implement the modification procedural requirements of engineering directives manual (EDM)-601, Engineering Change Manual, for a temporary modification that installed a valve leak seal enclosure on main steam drain valve 2SM-27. Specifically, EDM-601 required the weight and vibration response of the enclosure to be evaluated as part of the installation. The failure to consider this resulted in vibration induced piping failure upstream of the valve and an unexpected rapid plant down power. The failure to adequately implement a temporary modification in accordance with EDM- 601 was a performance deficiency (PD). The PD was more than minor because it was associated with the design control attribute of the initiating events cornerstone and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability during power operations. Specifically, the performance deficiency resulted in a rapid down power to approximately 20 percent and subsequent actions to take the Unit 2 turbine generator offline to repair the leak. Using NRC IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, the finding was determined to be of very low safety significance because the it did not contribute to both the cause of a reactor trip and affect mitigation equipment. The finding had a cross cutting aspect of consistent process, as described in the human performance crosscutting area because the licensee failed to use a consistent, systematic approach to make de.cisions during implementation of a temporary modification. |
| 05000369/FIN-2016001-01 | 2016Q1 | Mcguire McGuire | Failure to Maintain Fire Extinguishers in Contaminated Radiation Control Zones in Accordance with the Fire Protection Program | An NRC-identified Green non-cited violation (NCV) of the McGuire Nuclear Station Unit 1 and Unit 2 Renewed Facility Operating License Condition 2.C.4, Fire Protection Program (FPP), was identified for failure to perform annual maintenance on fire extinguishers located in contaminated radiation control zones (RCZs). The licensee took immediate corrective action to replace the past due fire extinguishers and entered the issue into their corrective action program as action request (AR) 02009794. The performance deficiency (PD) was more than minor because if left uncorrected the PD could have the potential to lead to a more significant safety concern, in that, fire extinguishers located in any contaminated RCZs may not be functional for firefighting purposes due to lack of maintenance. Every fire extinguisher, five total, located in a contaminated RCZ, did not have its annual maintenance up-to-date. The longest duration without annual maintenance was six years for two of the five extinguishers. The finding was determined to be of very low safety significance (Green) within the mitigating system cornerstone because it would not affect the ability to reach and maintain a safe shutdown condition, in that, for each of the fire areas where the out-of-date extinguishers were present, there were also properly maintained fire extinguishers and hose stations outside of the RCZ. The out-of-date extinguishers were weighed and it was determined that they would have performed their function, if needed. The cause of the PD was directly related to the cross-cutting aspect of field presence in the cross-cutting area of human performance because the licensee failed to correct deviations from the FPP and ensure proper oversight of the vendor contracted to perform fire extinguisher maintenance. |
| 05000369/FIN-2016002-01 | 2016Q2 | Mcguire McGuire | Failure to Perform General Visual Examinations of Containment Moisture Barriers Associated with Containment Liner Leak Chase Test Connections | An NRC-identified Green non-cited violation (NCV) of 10 CFR Part 50.55a, Codes and Standards, was identified for the licensees failure to perform general visual examinations of moisture barrier material in the reactor containment leak-chase channel test connections in accordance with the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME BPV Code), Section XI, Subsection IWE. The licensee performed the required examinations in Unit 1 during the March 2016 refueling outage and initiated corrective actions to revise the Containment Inservice Inspection (ISI) Plan. The licensee also planned to perform similar examinations in Unit 2 prior to the end of the first containment ISI period. Additionally, the licensee performed a containment operability determination to justify continuous operation of the Unit 1 and Unit 2 containment based on the results of all visual examinations, extent of condition activities, and the results of containment integrated leak rate tests. The licensee entered this issue into their corrective action program as action request (AR) 02038505. The failure to conduct the required visual examination of moisture barrier material in accordance with the ASME BPV Code was a performance deficiency (PD). The PD was of more than minor significance per IMC-0612, Appendix B, Issue Screening, because the current Containment ISI Plan did not adequately implement the ASME BPV Code requirements for the examination of moisture barriers, and if left uncorrected, it had the potential to lead to a more significant concern. The finding was of very low safety significance (Green) per IMC-0609 because it did not represent an actual open pathway in the physical integrity of the reactor containment and did not involve an actual reduction in function of hydrogen igniters in the reactor containment. The finding had a cross-cutting aspect of resolution in the problem identification and resolution cross-cutting area because the licensee did not take effective corrective actions to implement the ASME BPV code requirements in the Containment ISI Plan when a reasonable opportunity was available through the review of NRC Information Notice (IN) 2014-07. |
| 05000369/FIN-2016002-02 | 2016Q2 | McGuire | Failure to Ensure Containment Equipment Hatch Was Properly Closed During Fuel Movements | An NRC-identified Green NCV of Technical Specification (TS) 5.4.1.d, Procedures, was identified for the licensees failure to adequately implement the commitments in Selective Licensee Commitment (SLC) 16.9.25, Refueling Operations Containment Equipment Hatch, which required the containment equipment hatch to be closed during the movement of non-recently irradiated fuel inside containment. Specifically, during reactor vessel fuel reload activities, the inspectors identified that the equipment hatch was left partially open due to the failure to properly tighten the bolts evenly around the hatch resulting in direct communication of the containment atmosphere with the environment. The licensee took immediate corrective action to suspend fuel movements and properly tighten the equipment hatch bolts prior to resuming fuel movements and entered the issue into their corrective action program as ARs 02018605 and 02018701. The PD was more than minor because it impacted the configuration control attribute of the barrier integrity cornerstone and affected the cornerstone objective to provide reasonable assurance that containment protects the public from radionuclide releases caused by accidents or events. Additionally, if left uncorrected, the PD would have the potential to lead to a more significant safety concern. Specifically, the radiological barrier functionality of the containment equipment hatch was degraded due to the gap opening which could have allowed direct access of radiological releases from the containment atmosphere to the outside environment during a potential fuel handling accident inside containment. The inspectors screened the finding in accordance with IMC 0609, Appendix G, Shutdown Operations Significance Determination Process, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings. Because the finding degraded the ability to close or isolate the containment, it required review using IMC 0609, Appendix H, Containment Integrity Significance Determination Process. While the containment boundary function was considered degraded, the incident occurred eight days after the beginning of the refueling outage when short lived volatile radioisotopes had decayed sufficiently such that the potential radiological releases to the public would not likely contribute to the large early release frequency (LERF). Based on this, the finding was screened as having very low safety significance (Green). The cause of the PD was directly related to the cross-cutting aspect of procedure adherence in the cross-cutting area of human performance because the licensee failed to follow containment equipment hatch closing procedures which explicitly required performing a visual inspection that the containment equipment hatch was sealed and secured with metal-to-metal contact with the containment hatch flange and had no visual gaps. |
| 05000369/FIN-2016003-01 | 2016Q3 | McGuire | Licensee-Identified Violation | Technical Specifications 5.4.1.a, Procedures, requires, in part, that procedures for certain activities recommended in Regulatory Guide 1.33, Rev. 2, Appendix A, be established, implemented, and maintained. Administrative procedures for shift and relief turnover is one of the identified activities. Administrative procedure AD-OP-ALL-1000, Conduct of Operations, Rev. 4, implements the licensees shift and relief turnover standards. This procedure requires shift turnovers to contain detailed information on equipment and system status, alignments, and activities, to ensure watchstanders have a complete understanding of plant status. Contrary to the above, from August 10 to August 13, 2015, operators were not aware of the required nuclear service water system alignment which required a continuous vent (passing water flow) to be maintained in the condenser cooling water (RC) suction supply to the Unit 1 turbine driven auxiliary feedwater pump. The continuous vent mitigates the potential for air entrainment in the RC piping high point and is needed in order for the standby shutdown system to be functional during an Appendix R fire event when the suction of the turbine driven auxiliary feedwater pump is transferred from the auxiliary feedwater storage tank to the long term water supply provided by the RC system. This lack of operator awareness stemmed from a misunderstanding in the operator turnovers that the nuclear service water system was in a standby nuclear service water pond cooling alignment, which does not require the continuous vent to be maintained. The discrepancy was subsequently identified by oncoming shift operations personnel and the continuous vent was re-established on August 16, 2015, after removing material that obstructed the continuous vent line. As a result of not maintaining the continuous vent at the suction of the turbine driven auxiliary feedwater pump, the standby shutdown system was rendered non-functional for a period of eleven days, which was in excess of the 7-day limit allowed by Selected Licensee Commitments 16.9.7. This violation was determined to be of very low safety significance (Green) because it only affected the non-safety related Appendix R water supply to the turbine driven auxiliary feedwater pump. This violation was entered into the licensees corrective action program as NCR 01943414. |
| 05000370/FIN-2013005-01 | 2013Q4 | Mcguire | Evaluation of Gas Void Identified in Unit 2 ECCS Piping | During the performance of Unit 2 ECCS pipe gas void inspections using ultrasonic test (UT) equipment, a large gas void was found in a 5 foot section of 8 inch diameter piping at high point vent valve 2NV-1056. 2NV-1056 was located on the suction side of both trains of the NI and NV pumps downstream of valve 2ND-58A, which is opened during design basis accident conditions involving cold-leg recirculation to provide the piggyback alignment from the residual heat removal (ND) system. Excessive gas accumulation at 2NV-1056 could result in gas being drawn into the NI/NV pumps causing pump degradation or failure. The licensee vented the piping by opening 2NV-1056, which returned the ECCS piping to water solid conditions. Additional ECCS piping locations were checked for possible gas accumulation and none were identified. The licensee implemented increased frequency UT monitoring for gas accumulation at 2NV-1056 (every 6 hours and subsequently every 12 hours) to ensure timely detection of abnormal gas accumulation until the source was determined. Based on the UT measurements, the licensee determined the size of the gas void to be approximately 2 ft3, which exceeded the existing 0.35 ft3 maximum allowable void volume for this location. The licensee initiated a past operability evaluation to determine if the NI/NV pumps would have been capable of performing their safety function during design basis accident conditions with the void in the piping. In addition, on December 18, the inspectors observed how licensee personnel were conducting the increased frequency UT measurements at location 2NV-1056 using Enclosure 13.7, Supplemental Venting, of procedure PT/2/A/4200/019, ECCS Pumps and Piping Vent. The inspectors noted that personnel were conducting the UT measurement on the 1.5 inch diameter vent piping associated with 2NV-1056 versus the 8 inch ECCS header piping that the vent valve is connected to. The procedure contained a note stating that UT measurement is performed at piping adjacent to valve due to flow being limited by 1/8 inch diameter hole in piping header. The 2NV-1056 vent piping was previously added via a modification to enhance the licensees ECCS piping gas management program. It was installed using a wet tap with a 1/8 inch drilled hole into the top of the header piping with a coupling welded over the hole to connect the vent piping. Due to the small 1/8 inch opening, water tension and/or small trash/debris can inhibit the proper communication of water between the ECCS header pipe and the vent piping. It appeared to the inspectors that the note was directing that the UT measurement needed to be conducted on the ECCS header piping and not the vent piping due to concerns that the vent piping might remain water solid while the ECCS header piping could be voiding. Following discussions with the licensee regarding this note, personnel were directed to conduct the UT measurement in the ECCS header piping. The licensee initiated PIP M-13-11297 to address this issue and to investigate how prevalent past UT measurements were conducted in the vent piping versus the header piping. This issue remains unresolved pending completion of the licensees evaluation of the impact that the gas void would have on the operation of the NI/NV pumps during design basis accident conditions and investigation into the mechanism that resulted in the excessive gas voiding not being identified during routine surveillances designed to identify such conditions. This issue is identified as URI 05000370/2013005-01, Evaluation of Gas Void Identified in Unit 2 ECCS Piping. |
| 05000370/FIN-2014002-01 | 2014Q1 | Mcguire | Failure to Adequately Control Transient Combustible Materials in Accordance with the Fire Protection Program | An NRC-identified NCV of the McGuire Unit 1 and Unit 2 Renewed Facility Operating License Condition 2.C.4, Fire Protection Program (FPP), was identified for the licensees failure to adequately control the storage of transient combustibles in the 2A residual heat removal (ND)/containment spray (NS) heat exchanger room near safe shutdown equipment in accordance with the FPP requirements. The licensee initiated immediate corrective actions to evaluate the transient combustible fire loading and remove all the unapproved transient combustibles from the area. This condition was placed in the licensees corrective action program (CAP). The licensees failure to control the storage of transient combustibles in accordance with procedure NSD 313 was more than minor because it was associated with the Mitigating Systems cornerstone attribute of Protection Against External Factors (Fire) and adversely affected the cornerstone objective in that a fire involving transient combustibles could have affected nearby power cables and motor operator for valve 2ND-58A which provides a safe shutdown mitigation function. The finding was determined to have very low safety significance (Green) because it did not affect the ability of the reactor to reach and maintain cold shutdown condition. This finding had a cross cutting aspect of Teamwork in the Human Performance area because multiple groups were responsible for bringing the transient combustibles into the area and the individuals failed to effectively communicate and coordinate their activities to ensure that transient combustible control processes were appropriately implemented. |
| 05000370/FIN-2015004-01 | 2015Q4 | Mcguire | Failure to Report Unit 2 Unplanned Valid Auxiliary Feedwater Actuation in Mode 4 | An NRC identified Severity Level (SL) IV non-cited violation (NCV) of 10 CFR 50.72(b)(3)(iv)(A) was identified for the licensees failure to make a required NRC event notification within eight hours for an unplanned valid actuation of the auxiliary feedwater (CA) system. The unplanned valid actuation occurred during main turbine and main feedwater pump safety injection (SI) train trip function testing with Unit 2 in Mode 4 on October 7, 2015. The licensee entered this issue into their corrective action program and subsequently reported this CA actuation to the NRC on October 15, 2015. The failure to submit an event notification to the NRC within eight hours of occurrence of an unplanned valid CA system actuation in accordance with 10 CFR 50.72(b)(3)(iv)(A) was a performance deficiency (PD). Since the failure to submit an event report within the time requirements may impact the ability of the NRC to perform its regulatory oversight function, this PD was dispositioned under the traditional enforcement process and was determined to be a SL IV violation. Because this SL IV violation was not repetitive or willful, and did not have an underlying technical violation that would be considered more-than-minor, a cross-cutting aspect was not assigned to this violation. |
| 05000395/FIN-2007002-01 | 2007Q1 | Summer | nadequate Corrective Actions Results in Repetitive Spurious Tripping of EDG Room Ventilation Fan Molded Case Circuit Breakers | A Green self-revealing non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, was identified for inadequate corrective actions which resulted in repetition of a significant condition adverse to quality involving the spurious tripping of safety-related molded case circuit breaker associated with the A emergency diesel generator (EDG) room ventilation cooling fan A due to asymmetrical in-rush starting current. The licensee documented this failure in their corrective action program and implemented breaker trip setpoint changes to preclude spurious tripping from this phenomenon. This finding is more than minor because it affected the equipment performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The finding is of very low safety significance because it did not result in a loss of safety function of one or more trains of the EDGs and was not potentially risksignificant due to possible external events. The cause of this finding involved the thorough evaluation of identified problems aspect of the Problem Identification and Resolution cross-cutting area, in that, the extent of condition evaluation for the previous spurious trip of the A EDG room cooling fan B failed to consider the need to readjust the trip setpoint of A EDG room ventilation cooling fan A in order to mitigate the possibility of spurious tripping. (Section 1R15). |
| 05000395/FIN-2007004-01 | 2007Q3 | Summer | Review Risk Assessment Credit for Dedicated Manual Operator Actions During EDG Surveillance Testing | No findings of significance were identified. However, associated with the plant risk assessment review for testing the A EDG (Work Week 2007-31) in accordance with surveillance test procedure (STP)-125.002A, Diesel Generator Operability Test, the inspectors noted that dedicated manual operator risk management compensatory actions were being relied upon to allow the A EDG to remain available. The dedicated operators were required to be stationed in the EDG room area in order to reposition critical components manipulated during testing in the event of an EDG emergency start demand. Establishing prescribed compensatory measures to restore the EDG to service, if needed, allows the plant maintenance risk assessment to remain at Normal Risk versus Elevated Risk for a non-functional EDG. Further details related to the conduct of this surveillance test are discussed in Section 1R22 of this report. At the time of the inspection, the licensee could not provide complete information to support that these actions could be taken within the time necessary for the EDG to satisfy its safety function as assumed in the Probabilistic Risk Assessment (PRA) success criteria. Pending further NRC review of this issue and additional information from the licensee, this issue is identified as unresolved item (URI) 05000395/2007004-01, Review Risk Assessment Credit for Dedicated Manual Operator Actions During EDG Surveillance Testing. |
| 05000395/FIN-2008003-01 | 2008Q2 | Summer | Untimely Corrective Actions to Resolve Feedwater Regulating valve Malfunction Resulted in Reactor Trip | A Green self-revealing finding was identified for the failure to implement effective and timely corrective actions to prevent failure of a main feedwater regulating valve IFV000498 that resulted in a reactor trip. This valve failed due to previously identified pneumatic positioner pilot valve malfunction caused by either pilot valve stem fretting and/or foreign material intrusion from various internal air supply sources. All three loop feedwater regulating valve positioners and air supply components subject to potential sources of contamination were replaced prior to startup from the reactor trip. During Refueling Outage 17, modifications were completed to reduce vibration induced wear of control air system components and improve air quality to the positioners until the current positioner models can be replaced with a new design. This finding was entered into the licensee=s corrective action program as Condition Report 08-00292. This finding is greater than minor because it is associated with the Initiating Event Cornerstone attribute of equipment performance, and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during at-power operations. The finding was evaluated using Phase 1 of the At-Power SDP, and was determined to be of very low safety significance (Green) because it did not contribute to both the likelihood of a reactor trip and the likelihood that mitigating equipment or functions were not available. The cause of this finding was directly related to the aspect of appropriate and timely corrective action in the cross-cutting area of Problem Identification and Resolution (Corrective Action component) because actions to address previously identified feedwater regulating valve positioner pilot valve fretting and foreign material intrusion were not implemented in a timely manner (P.1.d). (Section 4OA2.3 |
| 05000395/FIN-2008004-01 | 2008Q3 | Summer | Failure to Perform EKG Tests During the Biennial medical Exam for Licenses Operators | The inspectors identified a Green NCV of 10 CFR 55.21 for failure to perform electro-cardiogram (EKG) tests during the biennial medical exam for licensed operators. Specifically, the inspectors identified three licensed operators who had not received EKGs as part of their biennial medical exams. The licensee conducted an extent of condition review and identified ten licensed operators who had not received EKGs during their biennial medical exams. The licensee scheduled those operators for EKGs. This issue is documented in the licensees corrective action program as Condition Report (CR) 08-03456. This finding is more than minor because if left uncorrected, it could become a significant safety concern if an undetected cardiovascular condition impacted an operators ability to direct or perform licensed activities. The finding affects the human performance attribute of the Mitigating Systems cornerstone because licensed operator response to initiating events mitigates undesirable consequences. Using the Significance Determination Process, this finding was determined to be of very low safety significance (Green) because the performance deficiency did not result in an actual operator performance error or plant event. The finding directly involved the cross-cutting area of Human Performance, component of Work Practices, and the aspect of supervisory and management oversight of work activities, including contractors, such that nuclear safety is supported, in that, the cause of the finding was the licensees lack of oversight of the off-site contract physicians clinic (H.4.c). (Section 1R11.2 |
| 05000395/FIN-2008004-02 | 2008Q3 | Summer | Failure to Maintain the Cotnrol Room Pressure Boundary Operable and Complete the Required TS Actions | The inspectors identified a Green NCV for failure to comply with Technical Specifications (TS) Limited Conditions for Operation (LCO) 3.7.6, Control Room Normal and Emergency Air Handling System, for the failure to maintain the control room boundary intact and operable, and complete the required TS actions. Specifically, the control room pressure boundary (CRPB) was breached and inoperable, and the Control Room Normal and Emergency Air Handling System was not capable of performing its TS function for a period of 17 days. The licensee completed repairs to the ductwork, restored compliance with the TS, and documented this issue in their corrective action program as CR-08- 00944 and CR-08-00972. This finding was more than minor because it affected the barrier performance attribute of the Barrier Integrity cornerstone and affected the cornerstone objective of providing reasonable assurance that the control room maintains radiological barrier functionality and protects the plant operators from radionuclide releases caused by accidents or events. The finding was evaluated using Inspection Manual Chapter 0609, Significance Determination Process, Phase I Worksheet for barrier integrity. The finding was determined to be of very low safety significance (Green) because it represented a degradation of the radiological barrier function provided for the control room. The finding directly involved the cross-cutting area of Human Performance, component of Resources, and aspect of Complete, Accurate and up-to-date Design Documentation and Procedures, in that, the post maintenance test for XAH0048 failed to include the verification of CRPB restoration through complete testing of the control room envelope (H.2.c). (Section 1R15.b) |