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	Quarter	Site	Title	Description
05000482/FIN-2011003-08	2011Q2	Wolf Creek	Failure to Maintain Reactor Coolant System Pressure Below Relief Valve Setpoint	The inspectors reviewed a self-revealing noncited violation of Technical Specification 5.4.1.a, Administrative Procedures, for failure to follow procedural requirements to maintain reactor coolant system pressure below 350 psig. Control room operators increased charging flow at too great a rate with the reactor coolant system water-solid which caused the pressurizer power-operated relief valve to cycle three times over several minutes until adjustments to letdown could be made to reduce reactor coolant system pressure. Also, the letdown pressure controller was left in manual when automatic control would have lessened the pressure increase. Wolf Creek wrote Condition Report 35244 to correct the deficiency by changing several procedures for water-solid plant operations. The failure to maintain pressure below the power-operated relief valve setpoint was a performance deficiency. The performance deficiency was more than minor because it impacted the Initiating Events Cornerstone objective of configuration control to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The significance of the finding was determined using Inspection Manual Chapter 0609, Significance Determination Process, Appendix G, Checklist 2, and determined to be of very low safety significance (Green), because it did not cause the loss of mitigating capability of core heat removal, inventory control, power availability, containment control, or reactivity control. Additionally, the finding also did not cause any low temperature overpressure technical specifications to be exceeded. The inspectors found that the cause of the finding had a cross-cutting aspect in the area of human performance. Specifically, operators had to rely on skill of the craft when procedures should have supplied more instruction for manipulating charging and letdown with a water-solid plant.
05000482/FIN-2011003-09	2011Q2	Wolf Creek	Inadequate Fire Watch Defeats Halon Fire Suppression in Vital Switchgear Rooms During Fire	The inspectors reviewed a self-revealing noncited violation of License Condition 2.C.5 for failure to implement adequate fire watches which affected both trains of vital ac and dc switchgear. The inadequate fire watches occurred during an actual fire which negated the Halon system discharge because internal fire doors were not shut, as required, by the fire watch. The inspectors found problems with fire impairments and watches from 2008 that had not been corrected. Subsequent to the fire, Wolf Creek again briefed and trained its personnel on the requirements for fire watches. This issue is captured in the corrective action program as Condition Report 36719. Failure to implement adequate fire impairments such that the fire watches ensured the success of the Halon system was a performance deficiency. The performance deficiency was more than minor because it impacted the Initiating Events Cornerstone and its objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the protection against external factors attribute was impacted by the fire impairment. To determine significance, the inspectors used Inspection Manual Chapter 0609.04 to screen the finding to Inspection Manual Chapter 0609, Appendix F, because the fire protection defense-in-depth strategies involving automatic suppression, fire barriers, and administrative controls were degraded. The senior reactor analyst conducted a Phase 3 review of this finding and concluded that the incremental core damage frequency was 1.6E-8 per year, or very low safety significance (Green). The inspectors found that the cause of the finding had a cross-cutting aspect in the area of problem identification and resolution. Specifically, corrective actions from ineffective fire watches in 2008 did not prevent recurrence of the inadequate fire watch on April 5, 2011.
05000482/FIN-2011003-10	2011Q2	Wolf Creek	Failure to Analyze for Vortexing in Containment Spray Additive Tank	The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to translate the design basis into instructions, procedures, and drawings. The inspectors found that the licensee failed to assess whether vortexing occurred in the containment spray additive tank in the event of a design-basis accident. Wolf Creek entered this issue in the corrective action program as Condition Report 38715. Failure to implement design control measures to analyze whether containment spray piping remained full of water was a performance deficiency. This finding was more than minor because it affected the design control attribute of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of the containment spray system to respond to initiating events and prevent undesirable consequences. Specifically, the inspectors had reasonable doubt on the capability of the containment spray system to properly inject because of vortexing in the containment spray additive tank. The inspectors performed the significance determination using Inspection Manual Chapter 0609.04. The finding was determined to be of very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in loss of operability or functionality. Although the failure to have this calculation had existed since original construction, the inspectors determined this finding reflected current performance since the licensee was required to evaluate likelihood of tanks allowing gas intrusion into the emergency core cooling systems in response to Generic Letter 2008-01, Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems. Consequently, this finding had problem identification and resolution cross-cutting aspects associated with the corrective action program in that the licensee did not thoroughly evaluate the potential for gas intrusion from all possible tanks
05000482/FIN-2011003-11	2011Q2	Wolf Creek	Licensee-Identified Violation	Title 10 CFR 50.54(hh)(2)(ii) states: Each licensee shall develop and implement guidance and strategies intended to maintain or restore core cooling, containment, and spent fuel pool cooling capabilities under the circumstances associated with loss of large areas of the plant due to explosions or fire, to include strategies in the following area of operations to mitigate fuel damage. On April 13, 2011, while performing procedure reviews as part of industry-wide self-assessments in response to the core damage events at Fukushima Daiichi, Wolf Creek engineers identified two instances of mitigating strategy procedures which did not contain sufficient information to accomplish those strategies successfully. The first example was the ability to refill the refueling water storage tank, and the second example involved flashing the diesel generator field using alternate dc sources. These issues were documented in the licensees corrective action program as Condition Report 37374. The inspectors evaluated these findings under Inspection Manual Chapter 0609, Appendix L, and determined these findings to be of very low safety significance because the findings did not involve unrecoverable unavailability of multiple mitigating strategies such that spent fuel pool cooling, injection to the reactor vessel, or injection to steam generators cannot occur, or unrecoverable unavailability of on-site, self-powered, portable pumping capability, or substantial inability to perform command and control enhancements.
05000482/FIN-2011003-12	2011Q2	Wolf Creek	Licensee-Identified Violation	Title 10 CFR Part 50, Appendix B, Criterion XI, Test Control, requires, in part, that a test program be established to assure that all testing required to demonstrate that structures, systems and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in the applicable design documents. On May 13, 2011, Wolf Creek identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XI, test control for stroking residual heat removal containment sump valve 8811B prior to its as-found diagnostic test. Wolf Creek stroked the valve for a clearance order and as such, preconditioned the valve prior to its test. Plant computer data from this stroke, data from the diagnostic stroke, and valve disassembly showed no deficiencies. Using Inspection Manual Chapter 0609.04, the inspectors determined the finding to be of very low safety significance because it was confirmed not to result in the loss of operability or functionality. This issue is captured in Condition Report 37244.
05000482/FIN-2011007-01	2011Q4	Wolf Creek	Failure to Verify Isolation of Associated Circuits on Isolation Switches	The team identified a finding because the licensee was not fully testing the isolation function of local transfer switches located at motor control center breakers for individual components. As a result, the licensee was not performing periodic verifications to confirm that local control circuits would be isolated from the effects of fire damage caused by a control room fire. The licensee documented this deficiency in Condition Report 045434. The failure to maintain adequate written procedures covering fire protection program implementation was a performance deficiency. Specifically, the licensee failed to ensure that component specific transfer switch testing procedures verified proper circuit isolation from the control room in the event of a control room fire. The performance deficiency was more than minor because it was associated with the procedure quality attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team evaluated the finding using Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, because it affected fire protection defense-in-depth strategies involving post-fire safe shutdown. Using Appendix F, Attachment 2, Degradation Rating Guidance Specific to Various Fire Protection Program Elements, the team determined that the finding constituted a low degradation of the safe shutdown area since the control room isolation feature is expected to display nearly the same level of effectiveness and reliability as it would had the degradation not been present. This finding screened as having very low safety significance (Green). Since the failure to test the isolation function had not been verified since initial installation, the team determined that this failure did not reflect current performance.
05000482/FIN-2011007-02	2011Q4	Wolf Creek	Inadequate Alternative Shutdown Procedure	The team identified a non-cited violation of Technical Specification 5.4.1.d for the failure to implement and maintain adequate written procedures covering fire protection program implementation. Specifically, the team identified two examples where the licensee failed to maintain an alternative shutdown procedure that ensured operators would prevent overfilling the pressurizer and steam generators, respectively. The licensee documented this deficiency in Condition Report 045442. The failure to maintain adequate written procedures covering fire protection program implementation was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external events (fire) attribute of the Mitigating Systems Cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team evaluated the significance of this finding using Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, because the performance deficiency affected fire protection defense-in-depth strategies involving post-fire safe shutdown systems. A senior reactor analyst performed a Phase 3 evaluation and determined this finding had very low risk significance based upon a bounding analysis (Green). This finding did not reflect current licensee performance
05000482/FIN-2011007-03	2011Q4	Wolf Creek	Failure to Ensure Post-Fire Safe Shutdown Components Remain Free of Fire Damage	The team identified a non-cited violation of License Condition 2.C(5) because the licensee failed to implement and maintain in effect all provisions of the approved fire protection program. Specifically, the licensee failed to properly analyze for fire damage in the form of shorts-to-ground related to the residual heat removal Train B refueling water storage tank suction valve and the pressurizer power-operated relief valves. Certain postulated shorts-to-ground could spuriously actuate these valves such that safe shutdown would be impacted. The licensee documented these deficiencies in Condition Reports 044912 and 045452, respectively. The failure to protect the residual heat removal Train B suction cables and the pressurizer power operated relief valve cables against all modes of cable failure during post-fire safe shutdown circuit analysis was a performance deficiency. The performance deficiency was more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team used Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, because the performance deficiency affected fire protection defense-in-depth strategies involving post-fire safe shutdown. The team categorized the finding as having a high degradation rating because the post-fire safe shutdown analysis was not complete. Because the Phase 1 screening criteria were not met, the team performed a Phase 2 analysis. The team walked down the affected fire area for each example as part of the Phase 2 quantitative screening. The team identified fire ignition sources and targets, and specific fire growth and damage scenario combinations for each example. The sum of the conditional core damage frequencies for the fire scenarios was 5.15E-7/year, which bounded the total change in core damage frequency associated with this performance deficiency. This performance deficiency had a cross-cutting aspect in the area of human performance associated with decision making because the licensee did not use conservative assumptions during their design review process. Specifically, the licensee did not follow industry guidance related to performing a circuit analysis
05000482/FIN-2011007-04	2011Q4	Wolf Creek	Procedure Inadequacies Related to Cold Shutdown Repairs	The team identified a non-cited violation of License Condition 2.C(5) because the licensee failed to implement and maintain in effect all provisions of the approved fire protection program. Specifically, the licensee failed to provide an adequate procedure for performing cold shutdown repairs required for post-fire safe shutdown. The licensee documented the deficiencies in Condition Reports 045397 and 045417. The failure to ensure that Procedure OFN RP-017A, Hot Standby to Cold Shutdown from Outside the Control Room Due To Fire, Revision 0, could be implemented as written was a performance deficiency. The performance deficiency was more than minor because it was associated with the procedure quality attribute of the Mitigating Systems Cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events (fire) to prevent undesirable consequences. The finding was evaluated for safety significance using NRC Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process. Since the finding was related to the ability to achieve and maintain cold shutdown, the finding screened to Green in Phase 1. This performance deficiency had a cross-cutting aspect in the area of human performance associated with resources because the licensee did not prepare an accurate and up-to-date procedure that assured nuclear safety. Specifically, personnel did not verify that the steps in the revised procedure could be performed as written and that the components had proper labeling
05000482/FIN-2011007-05	2011Q4	Wolf Creek	Licensee-Identified Violation	Licensee Event Report 05000482/2008-009-00 described a failure to establish a 1-hour fire watch compensatory measure in Fire Area A-27 related to a circuit issue. This was a violation of License Condition 2.C(5) and Procedure AP 10-104, Breach Authorization, Revision 25A. The performance deficiency was more than minor because it was associated with the protection against external events (fire) attribute of the Mitigating Systems Cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This finding was associated with fire prevention and administrative controls category. It was determined to be of very low safety significance since it involved a low degradation of the fire protection program. This issue was entered into the corrective action program as Condition Report 2008-05172. This violation is also discussed in Section 4OA3.1.
05000482/FIN-2011007-06	2011Q4	Wolf Creek	Licensee-Identified Violation	Licensee Event Report 05000482/2010-013-00 described three examples where the alternative shutdown capability was not independent of the control room. This was a violation of License Condition 2.C(5). The performance deficiency was more than minor because it was associated with the protection against external events (fire) attribute of the Mitigating Systems Cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. A bounding Phase 3 evaluation concluded that this issue had very low safety significance (Green) since each example was associated with a fire in 1 of the 103 control room cabinets. This issue was entered into the corrective action program as Condition Reports 030350 and 031408. This violation is also discussed in Section 4OA3.4.
05000482/FIN-2014005-02	2014Q4	Wolf Creek	Notice of Enforcement Discretion 14-4-02 for Emergency Diesel Generator B Exciter Cabinet Fire	An unresolved item (URI) is being opened to assess whether the cause for the request for enforcement discretion associated with the fire in the exciter circuit of emergency diesel generator B on October 6, 2014, involved a violation of NRC requirements. On October 6, 2014, at 1:26 p.m., emergency diesel generator B was declared inoperable when it tripped during a 24-hour surveillance test and operators identified a fire in an associated exciter cabinet. An Alert was declared and operators entered Technical Specification 3.8.1, AC Sources Operating, Required Action B.4.1, which required emergency diesel generator B be restored to operable status within 72 hours. Actions in response to the fire were completed, the fire was quickly suppressed, and WCNOC exited the Alert. Following the completion of repairs, WCNOC identified that postmaintenance testing required to demonstrate system operability included completing a 24-hour run. Since the postmaintenance testing and subsequent system restoration was expected to exceed the time remaining in the 72-hour action statement, WCNOC requested that the NRC exercise discretion to not enforce compliance with the actions required in Wolf Creek Generating Station Technical Specification 3.8.1, Required Action B.4.1, and approve an additional 8 hours to restore the system. NOED NO. 14-4-02, documents this request and the NRCs approval. Following postmaintenance testing, emergency diesel generator B was restored to operable status at 5:17 p.m. on October 9, 2014. WCNOC concluded that the most likely cause of the event was the failure of the power current transformers power rectifier bridge. WCNOC postulated that when the bridge failed, power from the power current transformers to the generator field was lost. As a result, the voltage regulator attempted to maintain the field current using only the power potential transformer. Since the power potential transformer is not rated to sustain full field current, the transformer was overloaded, which caused it to overheat and catch fire. Troubleshooting also indicated that the emergency diesel generator B tripped on phase differential current for the same reasons. WCNOC removed the failed rectifier bridge for further analysis in December 2014; at the end of the inspection period, WCNOC personnel were awaiting additional failure analyses of the failed rectified bridge to determine the specific direct causes of the fire and unplanned emergency diesel generator B inoperability. The root cause is being evaluated by Condition Report 88665. When an NOED is issued, Inspection Manual Chapter 0410, Notice of Enforcement Discretion, requires that a URI will be opened to assess whether the cause(s) of the events leading up to the request for the Notice of Enforcement Discretion involved violations of NRC requirements. This issue will be tracked as a URI in order to review and evaluate WCNOCs additional rectifier bridge failure analyses, root cause analysis, and other supporting documentation to determine if a violation exists: URI 05000482/2014005-02, Notice of Enforcement Discretion 14-4-02 for Emergency Diesel Generator B Exciter Cabinet Fire. These activities constitute completion of two event follow-up samples, as defined in Inspection Procedure 71153.
05000482/FIN-2014008-01	2014Q4	Wolf Creek	Inadequate Alternative Shutdown Procedure	The team identified a non-cited violation of Technical Specification 5.4.1.d for the failure to implement and maintain adequate written procedures covering fire protection program implementation. Specifically, the licensee failed to maintain an alternative shutdown procedure that ensured operators could safely shut down the plant under all postulated fire scenarios. A scenario which could impact the operation of the required diesel generator was not adequately addressed. The licensee implemented a fire watch in the control room as a compensatory measure until corrective actions can be taken. The licensee documented the deficiencies with Procedure OFN RP-017, "Control Room Evacuation," Revision 45, in Condition Report 00089788. The failure to maintain adequate written procedures covering fire protection program implementation was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external events (fire) attribute of the Mitigating Systems Cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. A senior reactor analyst performed a Phase 3 evaluation to determine the risk significance of this finding since it involved a postulated control room fire that led to control room evacuation and determined that the finding was of very low safety significance. The finding did not have a cross-cutting aspect since the performance deficiency was more than three years old and not indicative of current performance.
05000482/FIN-2014008-02	2014Q4	Wolf Creek	Procedure Inadequacies Related to Cold Shutdown Repairs	The team identified a non-cited violation of Technical Specification 5.4.1.d for the failure to implement and maintain adequate written procedures covering fire protection program implementation. Specifically, the licensee failed to adequately label equipment and provide an adequate procedure for performing cold shutdown repairs required for post-fire safe shutdown. Since the plant would already be stable in hot shutdown, no immediate compensatory or corrective actions were required to assure safety. The licensee was evaluating corrective actions. The licensee documented the deficiencies in Condition Report 00089130. The failure to ensure that Procedure OFN RP-017A, "Hot Standby to Cold Shutdown from Outside the Control Room Due To Fire," Revision 9, could be implemented as written was a performance deficiency. The performance deficiency was more than minor because it was associated with the procedure quality attribute of the Mitigating Systems Cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events (fire) to prevent undesirable consequences. The finding was evaluated for safety significance using NRC Inspection Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process." Since the finding was related to the ability to achieve safe shutdown and the plant would remain in stable hot shutdown, the finding screened to Green. This performance deficiency had a cross-cutting aspect in the area of human performance associated with documentation because the licensee did not prepare an accurate and up-to-date procedure that assured nuclear safety. Specifically, personnel did not verify that the steps in the revised procedure could be performed as written and that the components had proper labeling.
05000482/FIN-2014008-03	2014Q4	Wolf Creek	Licensee-Identified Violation	On October 9, 2013, Licensee Event Report 05000482/2013-009 described an unanalyzed condition wherein a fire in fire areas containing certain unprotected, shunt, direct current ammeter circuits could result in secondary fires outside the initial fire area. This condition can occur only under specific circuit fault conditions wherein a fire causes a short to ground on cables associated with the DC ammeters, concurrent with a short to ground on a safety-related 125V DC circuit on the negative side of the same battery source. This condition is known as a ground equivalent hot short. The ammeter circuits are not overcurrent protected and, consequentially, could overheat and ignite anywhere along the route of the associated ammeter cables. The licensee determined in a cause analysis and extent of condition review that the only affected circuits were the eight DC ammeters associated with the safety-related 125V DC batteries and battery chargers. The team determined the routing of the affected cables involved five fire areas. The five fire areas were switchboard rooms/battery rooms FA C16, locked cable chase FA C18, locked cable chase FA C24, lower cable spreading room FA C21, and control room FA C27. The team evaluated the various scenarios of where the primary fire could start and where the secondary fire(s) could develop. The team determined that there were no normal ignition sources (electrical cabinets or equipment) in the two locked cable chases or the lower cable spreading room. Transient combustibles in these areas were strictly controlled by Procedure AP 10-102, "Control of Combustible Materials," Revision 18, and the cable chases are protected by a wet pipe automatic sprinkler system and automatic smoke detection alarms in the main control room. A fire in the switchboard rooms/battery room FA C16 is the area where the ammeter shunts are located and a primary fire there would not result in a secondary fire outside this area. The only remaining fire area to evaluate was the control room FA C27, which had a low combustible loading, all cables entering the control room are IEEE 383 rated, and the cables and cable trenches were protected by an automatic halon extinguishing system and automatic smoke detectors. While a primary fire in the control room could cause a secondary fire to develop along the cable route, the team determined that in the event that a secondary fire did occur that the impact would be limited to the same train as the primary fire. Therefore, the redundant post-fire safe shutdown success path would be unaffected by the fire. License Condition 2.C(5)(a) specifies, in part, that the licensee implement and maintain in effect all provisions of the approved fire protection program as described in the Standardized Nuclear Power Plant System Final Safety Analysis Report through Revision 17, the Wolf Creek site addendum through Revision 15, and as approved in the Safety Evaluation Report through Supplement 5, Amendment No. 191, Amendment No. 193, and Amendment No 205. Section 4.4.1.1 of the Fire Hazards Analysis states that "only one fire is postulated to occur at any one time and multiple fires are not postulated." Contrary to the above, since initial construction until November 6, 2014, the licensee failed to implement the fire protection program that ensured only one fire will occur at one time. Specifically, the licensee failed to ensure that direct current ammeter circuits were properly protected to prevent secondary fires from initiating in other areas of the plant. The performance deficiency was more than minor because it was associated with the protection against external factors attribute of the Mitigating Systems Cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events (fire) to prevent undesirable consequences. The finding was evaluated for safety significance using NRC Inspection Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process," dated September 20, 2013. Since the finding was related to the ability to achieve safe shutdown and safe shutdown would be unaffected, the inspectors determined the finding had a very low safety significance (Green). Specifically, no secondary fires resulting from a primary fire could prevent the reactor from achieving safe shutdown. The licensee documented this issue in their corrective action program as Condition Report 00074959. This violation is also discussed in Section 4OA3.
05000482/FIN-2015004-01	2015Q4	Wolf Creek	Inadequate Measures to Assure SGK05A Issues Were Promptly Corrected	The inspectors identified a Green cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees inadequate measures to assure that corrective action was taken to preclude repetition of a significant condition adverse to quality. Specifically, measures to correct train A Class 1E electrical equipment air-conditioning system (SGK05A) issues following two trips of the unit on October 18, 2013, failed to preclude repetition, which resulted in the SGK05A unit tripping twice on May 15, 2015; the train A safety-related batteries, inverters, and alternating and direct current buses being declared inoperable due to the loss of area cooling; two separate Technical Specification 3.0.3 entries; and separate technical specification required reactor power reductions to 93 and 94.7 percent. The licensees immediate corrective actions included troubleshooting to determine the direct cause of the compressor trips, stationing a dedicated operator following the second trip on May 15, 2015, and subsequently implementing Temporary Modification 15-013-GK-00, which restored compliance. Actions to prevent recurrence following the May 15, 2015, SGK05A trips, documented in apparent cause evaluation 96392, included conducting a seminar with station managers to review lessons learned from the event, completing a change package to replace the SGK05A compressor that has been the source of residual contamination that has led to numerous trips of the unit, and tracking of the timely replacement of the SGK05A compressor with a due date of December 15, 2016. Wolf Creek entered this issue into its corrective action program as Condition Reports 96392 and 96397. This finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the train A safety-related batteries, inverters, and alternating and direct current buses became inoperable and their capability to respond to initiating events to prevent undesirable consequences was impacted as a result of the SGK05A unit tripping. In accordance with Inspection Manual Chapter 0609.04, Initial Characterization of Findings, and Exhibit 3 of Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012, and April 29, 2015, respectively, the performance deficiency affects a mitigating structure, system, and component. The performance deficiency does not affect the design or qualification of a mitigating structure, system, and component, and the structure, system, and component did not maintain its functionality. Additionally, the finding does not represent a loss of system and/or function, the finding does not represent an actual loss of function of at least a single train for greater than its technical specification allowed outage time or two separate safety systems out-of-service for greater than their technical specification allowed outage time, and the finding does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. Therefore, the inspectors determined that this finding is of very low safety significance (Green). In accordance with Inspection Manual Chapter 0310, Aspects Within The Cross-Cutting Areas, issued December 4, 2014, the finding has a cross-cutting aspect in the area of human performance, resources, because the licensee did not ensure that personnel, equipment, procedures, and other resources were available and adequate to support nuclear safety. Specifically, senior managers did not ensure successful completion of the replacement of the SGK05A compressor in Refueling Outage 20, which was a missed opportunity that resulted in the SGK05A unit tripping twice on May 15, 2015, as a result of the same direct cause.
05000482/FIN-2015004-02	2015Q4	Wolf Creek	Failure to Ensure Essential Service Water Valves Were Adequately Protected from External Flooding Hazards	The inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to establish measures to assure that applicable regulatory requirements and the design basis, for applicable structures, systems, and components, are correctly translated into specifications, drawings, procedures, and instructions. Specifically, the licensee failed to ensure that safety-related essential service water valves in the control building were adequately protected from external flooding hazards in the event of a design basis local intense precipitation event, which resulted in a reasonable doubt on the operability of safety-related essential service water valves. The stations immediate corrective actions included entering the condition into the corrective action program and performing a prompt operability evaluation that showed the essential service water valves remained operable. Additional corrective actions include accelerating three Fukushima project schedules that include a new sump pump in the turbine building area four cable vault, ground and surface water improvements for non-safety related electrical duct banks, and new sump pumps in electrical manholes near the turbine building. The violation was entered into the licensees corrective action program as Condition Report 102250. This finding is more than minor because it is associated with the design control attribute of the Mitigating Systems Cornerstone and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, during design basis local intense precipitation events, the safety-related essential service water train A and B service water cross-connect motor-operated valves EFHV0023, EFHV0024, EFHV0025, and EFHV0026, and the essential service water train A and B to service water system valves EFHV0039, EFHV0040, EFHV0041, and EFHV0042 were susceptible to external flooding hazards, and there was a reasonable doubt on the operability of these essential service water valves; however, subsequent evaluation determined that the essential service water valves would not have been impacted in the event of a design basis local intense precipitation event, and the valves were determined to be operable. In accordance with Inspection Manual Chapter 0609.04, Initial Characterization of Findings, and Exhibit 2 of Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012, and April 29, 2015, respectively, the performance deficiency affects mitigating structures, systems, and components. The finding is a deficiency affecting the design or qualification of mitigating structures, systems, and components, and the structures, systems, and components maintained their operability and functionality. Therefore, the inspectors determined that this finding is of very low safety significance (Green). In accordance with Inspection Manual Chapter 0310, Aspects Within The Cross-Cutting Areas, issued December 4, 2014, the finding has a cross cutting aspect in the area of human performance, challenge the unknown, because Wolf Creek individuals did not stop when faced with uncertain conditions. Specifically, the licensee did not maintain a questioning attitude during flooding walk-downs performed in accordance with NEI 12-07 or during evaluation of Condition Report 59257 to identify and resolve unexpected conditions like the floor drain pathway from the communication corridor to the control building basement (room 3101), which was an opportunity for the station to identify the open pathway from the exterior of the plant.
05000482/FIN-2015004-03	2015Q4	Wolf Creek	Failure to Perform an Adequate Operability Determination and Consider Design Basis Events	The inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to accomplish activities affecting quality in accordance with Procedure AP 26C-004, Operability Determination and Functionality Assessment, Revision 31. Specifically, the licensee failed to document an operability determination of sufficient scope to address the capability of safety-related essential service water valves in the control building to perform their specified safety functions in the event of a design basis local intense precipitation event. Immediate corrective actions included completing a prompt operability determination and performing analyses that determined the valves remained operable. Additional corrective actions include accelerating three Fukushima project schedules that include a new sump pump in the turbine building area four cable vault, ground and surface water improvements for non-safety related electrical duct banks, and new sump pumps in electrical manholes near the turbine building. The violation was entered into the licensees corrective action program as Condition Report 100299. This finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, during design basis local intense precipitation events, the safety-related essential service water train A and B service water cross-connect motor-operated valves EFHV0023, EFHV0024, EFHV0025, and EFHV0026, and the essential service water train A and B to service water system valves EFHV0039, EFHV0040, EFHV0041, and EFHV0042 were susceptible to external flooding hazards, and there was a reasonable doubt on the operability of these essential service water valves; however, subsequent evaluation determined that the essential service water valves would not have been impacted in the event of a design basis local intense precipitation event, and the valves were determined to be operable. In accordance with Inspection Manual Chapter 0609.04, Initial Characterization of Findings, and Exhibit 2 of Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012, and April 29, 2015, respectively, the performance deficiency affects mitigating structures, systems, and components. The finding is not a deficiency affecting the design or qualification of mitigating structures, systems, and components; the finding does not represent a loss of system and/or function; the finding does not represent an actual loss of function of at least a single train or two separate safety systems out-of-service for greater than their allowed outage times; and the finding does not represent an actual loss of function of one or more non-technical specification trains of equipment. Therefore, the inspectors determined that this finding is of very low safety significance (Green). In accordance with Inspection Manual Chapter 0310, Aspects Within The Cross-Cutting Areas, issued December 4, 2014, the finding has a cross-cutting aspect in the area of human performance, conservative bias, because Wolf Creek did not use decision making-practices that emphasize prudent choices over those that are simply allowable, and proposed action was not determined to be safe in order to proceed, rather than unsafe in order to stop. Specifically, the licensee did not consider long-term consequences or design basis events when determining how to resolve emergent concerns like the unexpected water in room 3101, which resulted in the licensees failure to thoroughly evaluate and assess impacts to the plant when Condition Report 96404 was entered into the corrective action program on May 17, 2015.
05000482/FIN-2016004-01	2016Q4	Wolf Creek	Failure to Adequately Establish and Adjust Preventive Maintenance for Emergency Diesel Generator Excitation System Diodes	Green. The inspectors identified a non-cited violation of Technical Specification 5.4.1.a, for the licensees failure to adequately develop and adjust preventive maintenance activities in accordance with Procedure AP 16B-003, Planning and Scheduling Preventive Maintenance, Revision 5. Specifically, the licensee did not create a preventive maintenance task for emergency diesel generator excitation system diodes, which resulted in degradation of the excitation system diodes in emergency diesel generator B. The licensee restored compliance by establishing preventive maintenance tasks 49286, 49287, 49288, and 49289, which refurbish the power rectifier assemblies and replace the diodes on a 12-year replacement frequency. The licensee entered this issue into the corrective action program as Condition Report 88665. The failure to adequately develop and adjust emergency diesel generator excitation system diode preventive maintenance activities in accordance with Procedure AP 16B-003, Planning and Scheduling Preventive Maintenance, was a performance deficiency. This finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated the finding using IMC 0609, Appendix A, Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, this finding was not a deficiency affecting the design or qualification of a mitigating structure, system, or component that maintained its operability or functionality; the finding did not represent a loss of system and/or function; the finding did not represent an actual loss of function of at least a single train for greater than its Technical Specification allowed outage time; and the finding did not represent an actual loss of function of one or more non-Technical Specification trains of equipment designated as high safety-significant. Therefore, the finding was of very low safety significance (Green). This finding has a cross-cutting aspect in the area of problem identification and resolution, operating experience, because the organization did not systematically and effectively evaluate relevant internal and external operating experience in a timely manner. This issue is indicative of current performance because the station did not take any formal corrective actions to address the stations failure to adequately consider operating experience (P.5)
05000482/FIN-2016008-01	2016Q2	Wolf Creek	Failure to Adequately Establish and Adjust Preventive Maintenance for Emergency Diesel Generator Excitation System Diodes	The inspectors identified a preliminary White finding associated with an apparent violation of Technical Specification 5.4.1.a, for the licensees failure to adequately develop and adjust preventive maintenance activities in accordance with Procedure AP 16B-003, Planning and Scheduling Preventive Maintenance, Revision 5. Specifically, the licensee did not create a preventive maintenance replacement task or schedule for emergency diesel generator excitation system diodes, which resulted in emergency diesel generator B being declared inoperable and unavailable when it tripped during a 24-hour surveillance test. The licensee entered this condition into its corrective action program as Condition Report 88665. The licensee restored compliance by establishing preventive maintenance tasks 49286, 49287, 49288, and 49289, which refurbish the power rectifier assemblies and replace the diodes on a 12-year replacement frequency. This finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, with one emergency diesel generator excitation system diode failed as a result of thermal degradation, emergency diesel generator B was not operable or available to perform its safety function. The inspectors evaluated the finding using Attachment 0609.04, "Initial Characterization of Findings," worksheet to Inspection Manual Chapter (IMC) 0609, Significance Determination Process, issued June 19, 2012. The attachment instructs the inspectors to utilize IMC 0609, Appendix A, Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that the finding required a detailed risk evaluation because it represented an actual loss of function of the emergency diesel generator B for greater than its technical specification allowed outage time. A senior reactor analyst performed a detailed risk evaluation in accordance with Appendix A, Section 6.0, Detailed Risk Evaluation. The calculated change in core damage frequency was dominated by a loss of offsite power initiator leading to station blackout with failures of the turbine-driven and non-safety-related auxiliary feedwater pumps. The analyst did not evaluate the large early release frequency because this performance deficiency would not have challenged the containment. The NRC preliminarily determined that the incremental conditional core damage probability for internal and external initiators was 1.54E-06, in the low to moderate risk significance range (White). This finding has a cross-cutting aspect in the area of problem identification and resolution, operating experience, because the organization did not systematically and effectively evaluate relevant internal and external operating experience in a timely manner. Specifically, Condition Report 55103 documented industry operating experience regarding emergency diesel generator excitation system diodes failing at an increased rate, and the operating experience was not effectively implemented and institutionalized through changes to station processes, procedures, equipment, and training programs, and at least one emergency diesel generator excitation system diode failed due to aging (P.5).
05000482/FIN-2016008-02	2016Q2	Wolf Creek	Failure to Promptly Identify and Correct a Condition Adverse to Quality Associated with the Emergency Diesel Generator B Excitation System Diodes	The inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to assure that conditions adverse to quality, such as failures, malfunctions, and deficiencies are promptly identified and corrected. Specifically, the licensee failed to promptly identify and correct a failed rectifier bridge diode after smoke was observed coming from the three power potential transformers in the emergency diesel generator exciter cabinet NE106 on June 11, 2014, which contributed to the emergency diesel generator B being declared inoperable and unavailable when it tripped during a 24-hour surveillance test on October 6, 2014. To address the failure to take adequate corrective actions Wolf Creek entered this issue into its corrective action program as Condition Report 105480 and plans to implement a modification to install overcurrent detection for each emergency diesel generators power potential transformer. This finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to identify and correct the failed emergency diesel generator excitation system diode contributed to the emergency diesel generator B failure on October 6, 2014. The inspectors evaluated the finding using Attachment 0609.04, "Initial Characterization of Findings," worksheet to Inspection Manual Chapter (IMC) 0609, Significance Determination Process, issued June 19, 2012. The attachment instructs the inspectors to utilize IMC 0609, Appendix A, Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012. The inspectors determined this finding is not a deficiency affecting the design or qualification of a mitigating structure, system, or component that maintained its operability or functionality, the finding does not represent a loss of system and/or function, the finding does not represent an actual loss of function of at least a single train for greater than its Technical Specification allowed outage time, and the finding does not represent an actual loss of function of one or more non-Technical Specification trains of equipment designated as high safety-significant. Therefore, the inspectors determined the finding was of very low safety significance (Green). The inspectors determined that in accordance with Inspection Manual Chapter 0310, Aspects Within The Cross-Cutting Areas, issued December 4, 2014, the finding has a cross-cutting aspect in the area of human performance, conservative bias, because when smoke was identified coming from the power potential transformers on multiple occasions, licensee personnel did not use decision making-practices that emphasize prudent choices over those that are simply allowable, and a proposed action is determined to be safe in order to proceed, rather than unsafe in order to stop. As a result, the licensee missed an opportunity to identify and correct the condition of the failed diode in the static exciter (H.14).
05000482/FIN-2017008-01	2017Q4	Wolf Creek	InadequateEvaluation of Spurious Valve Operation	The team identified a non-cited violation of License Condition 2.C.(5) for failure to implement and maintain in effect all provisions of the approved fire protection program. Specifically, the licensee failed to adequately evaluate the potential impacts on post-fire safe shutdown of two motor operated valves spuriously closing due to fire damage.The failure to adequately evaluate the impact of pressure operated relief valve block valves spuriously closing on post-fire safe shutdown was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external events (fire) attribute of the Mitigating Systems cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Because the finding affected the ability to reach and maintain safe shutdown conditions in case of a fire that led to control room evacuation and because the Phase 2 methodology of Inspection Manual Chapter 0609, Appendix F, was not appropriate for this finding, a senior reactor analyst performed a Phase 3 evaluation to determine the risk significance. The analyst determined this finding was of very low risk significance (Green). There is no cross-cutting aspect associated with this finding since the performance deficiency is not reflective of present performance (i.e., the performance deficiency occurred more than 3 years ago).
05000482/FIN-2017008-02	2017Q4	Wolf Creek	Failure to Provide Adequate Emergency Lighting	The team identified a non-cited violation of License Condition 2.C.(5) for failure to provide emergency lighting along alternate routes plant operators are allowed to take during implementation of the procedure for control room evacuation due to fire.The failure to provide 8-hour emergency lights along alternate routes used by operators during control room evacuation due to fire is a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external factors attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team evaluated this finding using Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, dated September 20, 2013, because it affected the ability to reach and maintain safe shutdown conditions in case of a fire. The team assigned the finding to the post-fire safe shutdown category since it impacted the alternate shutdown element. The issue screened to Green because the reactor would be able to achieve and maintain hot shutdown because the operators are required to carry flashlights. Specifically, the team had reasonable assurance that the operators would be able to complete the evacuation procedure using handheld flashlights to access safe shutdown equipment. The finding is assigned a cross-cutting aspect in the area of human performance, associated with training, because the operators are not being trained on the access and egress routes that are provided with 8-hour emergency lights during implementation of the control room evacuation procedure due to fire to ensure the time critical actions can be met (H.9).
05000483/FIN-2007004-01	2007Q3	Callaway	Licensee Practices Allow Protective Action Recommendations for Areas Where Protective Action Guides Are Not Exceeded	The licensee and offsite officials have implemented a protective action scheme in which the licensee makes recommendations for areas 2 miles in radius from the plant, and in -13- Enclosure affected sectors, sections 2 to 5 miles from the plant and sections 5 to 10 miles downwind, where a sector is a wedge-shaped area of the emergency planning zone marked by lines of radius 2212 of arc apart. Offsite officials make and implement protective action decisions in geographical zones which are typically several sectors \\\"wide\\\" and may cross the five and ten mile section boundaries. The NRC identified the licensee\\\'s implementation and understanding of procedure EIP-ZZ-00212, \\\"Protective Action Recommendations,\\\" Revision 21, allows the licensee to generate shelter or evacuation protective action recommendations for members of the public in areas of the emergency planning zone where radiological protective action guides have not been exceeded. Specifically, licensee processes allow protective action recommendations to be made for areas of the emergency planning zone 5 to 10 miles away from the reactor when those areas are not affected by the radiological plume. Federal guidance for the choice of protective actions during an emergency is described in EPA-400-R-92-001, Manual of Protective Action Guides and Protective Actions for Nuclear Incidents (EPA-400). Specifically, a protective action recommendation is warranted when radiation doses are projected between 1 and 5 rem (Total Effective Dose Equivalent) or between 5 and 25 rem (Thyroid Committed Effective Dose Equivalent); guidance further recommends that during the plume phase protective action decisions be based primarily on plant conditions and dose projections, without waiting for confirming environmental measurements. Federal guidance states that protective actions are seldom justified in areas where the protective action guides are not exceeded, based in part on minimizing the overall risk to the public. The licensee and offsite agencies have adopted a prompt protective action scheme based on EPA-400 and NUREG-0654, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants, Revision 1, Supplement 3, \\\"Criteria for Protective Action Recommendations for Severe Accidents.\\\" Upon declaration of a General Emergency classification the licensee recommends a minimum protective action of evacuation in a keyhole area 2 miles in radius in all directions, and evacuation to 5 miles in sectors affected by the wind, unless a radiological evaluation indicates a more extensive recommendation is required (under some conditions a shelter recommendation would be made). The inspectors determined the licensee procedure also provided for retaining protective action recommendations as meteorological conditions change, as discussed in Regulatory Information Summary 2003-12, \\\"Clarification of NRC Guidance for Modifying Protective Actions\\\" (that is, once a protective action is recommended for implementation, no reduction in the protective action recommended for that area is permitted). However, inspectors determined that, with an existing minimum keyhole protective action recommendation, when a wind shift is followed by an increase in radiological release severity, the licensee\\\'s practice is to recommend protective actions be taken between 5 and 10 miles from the plant in every emergency planning zone sector that previously had recommendations for actions between 2 and 5 miles from the reactor, regardless of the prevailing wind direction at the time when the radiological release increases in severity. The inspectors interviewed a group including licensee emergency planners, emergency response organization dose assessment staff, and licensee management, to determine the licensee\\\'s expectations and practices for making protective action recommendations -14- Enclosure under conditions of changing wind directions. The inspectors posed an example situation consisting of an initial two-mile 360 evacuation with three-sector evacuation between 2 and 5 miles downwind, followed by a slowly changing wind direction so that over a period of more than an hour the wind stabilizes at a direction opposite its initial direction (that is, a 180 wind direction change), followed more than an hour later by an increase in core damage severity requiring an extension of protective actions to 10 miles downwind. Licensee emergency planners, dose assessment staff, and management all strongly indicated that in the situation described the licensee would recommend an extension of protective actions to ten miles in all sectors along the 180 arc (that is, in any sector previously recommended for actions between 2 and 5 miles), and the licensee\\\'s recommendation would not be limited to the three affected sectors at the time core damage increased in severity. The inspectors determined the licensee\\\'s practice always result in appropriate protective action recommendations to offsite authorities for areas where there is radiological risk to the public, but under conditions of changing wind direction and release severity, the licensees practices can also result in recommendations to take actions in areas where dose assessment identifies radiological risk does not exist. The inspectors determined that the licensee had not adequately defined when an area of the emergency planning zone was affected by a radiological plume, in that a preexisting protective action recommendation for areas 2 to 5 miles from the plant should not automatically require future extension to areas 5 to 10 miles away in the absence of supporting radiological analysis (which may include appropriate professional judgement when the basis is described and documented). Inspectors determined that not increasing protective actions to include areas with no previous recommendation is not the same as reducing a previously-made recommendation, even for adjacent areas within the same sector; the 2 to 5 mile and 5 to 10 mile areas should be considered as distinct from one another in arriving at protective action recommendations. This issue has been entered into the licensees corrective action system as CAR 200707375. This issue is unresolved pending consultation with the Federal Emergency Management Agency, because the issue involves licensee processes for making protective action recommendations to offsite officials: URI 05000483/2007004-01, Licensee Practices Allow Protective Action Recommendations for Areas Where Protective Action Guides Are Not Exceeded
05000483/FIN-2007005-06	2007Q4	Callaway	Licensee-Identified Violation	10 CFR 50.83 requires, in part, that prior to releasing part of a facility or site for unrestricted use, the licensee shall obtain NRC approval. Contrary to this, in January 2007, the licensee identified that on April 20, 2004, it had sold two parcels of land totaling 0.83 acres near the Highway 94 bridge at Logan Creek to the State of Missouri without obtaining NRC approval of the release of the land for unrestricted use. This was entered in the licensees corrective action program as CAR 200700893. On May 15, 2007, the licensee submitted a request, consistent with 10 CFR 50.83, seeking approval to release these parcels of land. This finding is of very low safety significance because the property meets the definition of non-impacted areas in accordance with 10 CFR 50.2, and the property has no reasonable potential for residual radioactivity in excess of natural background. In accordance with the NRC Enforcement Policy, because the violation impacted the regulatory process, it was not suitable for evaluation under the Significance Determination Process and, therefore, was categorized at Severity Level IV.
05000483/FIN-2008006-01	2008Q1	Callaway	Nonconservative Technical Specifications for Battery INTER-CELL Connection Resistance	The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, because the licensee failed to ensure that Technical Specification Surveillance Requirements for the NK11 and NK14 safety-related batteries established limits that met the design requirements. Specifically, until questioned by the team the licensee failed to determine the required design value needed to assure plant safety as requested in Callaway Action Request 200706561. The licensee determined that 69 micro-ohms should be the actual allowed inter-cell voltage limit to meet the design requirements versus an allowed Technical Specification limit of 150 micro-ohms. The performance deficiency associated with this finding involved the failure to ensure that the NK11 and NK14 safety-related batteries would remain operable if all the inter-cell connections measured 150 micro-ohms as allowed by Technical Specification Surveillance Requirements 3.8.4.2 and 3.8.4.5. This finding was greater than minor because it was associated with the Mitigating Systems cornerstone attribute of maintenance and testing and affects the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, \\\\\\\"Phase 1 Initial Screening and Characterization of Findings,\\\\\\\" the finding was determined to have very low safety significance because it was a design deficiency confirmed not to result in loss of operability. The finding had a cross-cutting aspect in the area of problem identification and resolution associated with operating experience because the licensee failed to evaluate in a timely manner relevant internal and external operating experience P.2(a) (Section 4OA2.e)
05000483/FIN-2008006-02	2008Q1	Callaway	Licensee-Identified Violation	Technical Specification 5.4.1.d requires that AmerenUE maintain a fire protection program. Procedure APA-ZZ-0071, Control of Impairments of Fire Protection Systems and Components, requires personnel to maintain the integrity of plant fire doors. Contrary to this, security officers identified during routine tours on March 6, March 20, July 18, and July 31, 2007, which personnel failed to maintain the integrity of Fire Door 15031. This licensee documented these deficiencies in Callaway Action Requests 200702037, 200702596, 200706810, and 200707100, respectively. This finding is of very low safety significance because the exposed fire area contained no potential damage targets that are unique from those in the exposing fire are
05000483/FIN-2015007-01	2015Q2	Callaway	Failure to Identify and Evaluate All Targets Within the Zone of Influence of Ignition Sources	The inspectors identified a non-cited violation of 10 CFR 50.48(c) and National Fire Protection Association Standard 805 for the licensees failure to address the effects of fire damage to risk-significant circuits impacted by an analyzed fire scenario. Specifically, the licensee failed to identify that a target cable raceway containing circuits that could impact the ability to achieve safe and stable conditions during a fire would be impacted during a fire scenario. The licensee entered this issue into their corrective action program as Callaway Action Request 201503262. The inspectors determined that the failure to identify a fire risk important cable raceway impacted by a fire scenario was a performance deficiency. The performance deficiency was determined to be more than minor because it is associated with the reactor safety Mitigating Systems cornerstone attribute of protection against external factors (i.e., fire) and it 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 was screened in accordance with NRC Inspection Manual Chapter (IMC) 0609, Appendix F, Fire Protection Significance Determination Process, dated September 20, 2013, as the finding affected post-fire safe shutdown. Using IMC 0609, Appendix F, Attachment 1, Fire Protection Significance Determination Process Worksheet, dated September 20, 2013, the finding was screened as a Green finding of very low safety significance in accordance with Step 1.3. The finding did not have a cross-cutting aspect since it was not indicative of current licensee performance since the original ignition source and target walkdowns were performed in 2010.
05000498/FIN-2008008-01	2008Q3	South Texas	Potential Fire Damage to the Fire Suppression Water Supply System	The team identified an unresolved item associated with License Condition 2.E for Unit 1 concerning failure to adequately implement the approved fire protection program. Specifically, the team noted the possibility that the licensees fire protection program may not have ensured a reliable fire protection water supply for the fixed and manual fire suppression systems credited as part of the fire protection program defense-in-depth approach. The fire suppression water supply system has three, 100 percent capacity, diesel-engine driven fire water pumps (PA0121, 0221, 0421) located in the fire pump house. One fire pump is required to supply water for fixed water suppression systems and fire hoses. Each pump will start automatically when low pressure is sensed in the pump discharge header. Each pump can also be started manually via control switches located in the Unit 1 main control room and the fire pump house. All three pumps discharge into a common header that supplies an underground piping ring main system. The team determined that the cables for control room manual pump starting could be damaged by a fire between the control room and the fire pump house. Such damage could result in loss of both automatic pump starting and manual start from the control room. A short to ground on a single cable would prevent the automatic starting of its respective pump. The team also determined that the cables for all three fire pumps are routed in the same cable trays. Therefore, all three cables might be exposed to potential damage by a single fire. Cables N0FP1C1SC, N0FP01C2SB and N0FP01C3SB, are routed together in the same cable trays through nine fire areas (Fire Areas 01, 03, 04, 31, 34, 61, 65, 67, and 70). No water would be available for fire suppression until at least one pump was manually started in the pump house if these cables were all damaged.
05000498/FIN-2008008-02	2008Q3	South Texas	Potential Loss of Centrifugal Charging Pump Suction Due to Fire Damage	The team identified an unresolved item associated with License Condition 2.E concerning a potential failure to adequately implement the approved fire protection program. Specifically, the team identified that the licensees fire protection program may not have ensured that the charging pump relied on for achieving post-fire safe shutdown would not be damaged because of a loss of suction. During a post-fire safe shutdown, the charging pump would be necessary to support the reactivity control and reactor coolant makeup functions by providing the reactor coolant system borated water from the refueling water storage tank. The team identified that fire damage to unprotected cables for either of the motor operated valves in the normal suction path had the potential to cause the associated valve to close, which could damage the running pump. During normal plant operations, the chemical and volume control system operates to allow a continuous feed (charging and seal injection) and bleed (letdown and seal leak-off) for the reactor coolant system. Normally one centrifugal charging pump is in operation. In the event of fire, inventory makeup is intended to be accomplished using a centrifugal charging pump by switching to the refueling water storage tank as a source of borated water. Procedure 0POP04-ZO-0009, Safe Shutdown Fire Response, included steps to swap the suction path from the normal suction source to the refueling water storage tank without securing the running charging pump. However, the team determined that if the charging pump credited for safe shutdown was running at the time of the fire, a spurious closure of either one of the two series-connected volume control tank outlet valves (1-MOV-112B or 1-MOV-113A) could result in a loss of suction and damage to the credited charging pump. The post-fire safe shutdown strategy developed by the fire protection program was intended to assure the availability of only one charging pump in fourteen fire areas. The team identified thirteen of these fire areas (Fire Areas 03, 04, 20, 24, 25, 26, 27, 31, 32, 33, 34, 65 and 67) also contain unprotected cables that had the potential to spuriously close at least one of the volume control tank outlet valves (MOV-112B or MOV-113A) due to fire damage. Instructions in Procedure 0POP04-ZO-0009 for each of these fire areas direct the control room operators to place the control switches for both centrifugal charging pumps to the pull-to-lock position to secure the pumps and prevent potential restarting until their suction is aligned to the refueling water storage tank. The team determined that the assumed success of this action was based on an unverified assumption that circuit damage would not occur prior to 10 minutes after control room operators decided to enter Procedure 0POP04-ZO-0009. Entry into Procedure 0POP04-ZO-0009 is based on satisfying criteria provided on the Conditional Information Page of Procedure 0POP04-ZO-0008, Fire/Explosion. The entry conditions for Procedure 0POP04-ZO-0008 are: (1) Verbal report of a fire or an explosion from a person at the scene, or (2) Alarm on the Fire Protection Computer with confirmation of an actual fire in the affected area by an operator dispatched to the fire area. Step 6.0 of Procedure 0POP04-ZO-0008 states, Station the STA (Shift Technical Advisor) in the affected unit to monitor the Conditional Information Page (CIP) until fire is out or transition to Procedure 0POP04-ZO-0009, Safe Shutdown Fire. The procedures would create a time delay between the start of a fire and the decision to initiate a plant shutdown. The team was concerned that fire damage to unprotected cables could spuriously close a volume control tank outlet valve prior to the control room operators securing the credited centrifugal charging pump in accordance with Procedure 0POP04- ZO-0009.
05000498/FIN-2009003-01	2009Q2	South Texas	Failure to Identify Maintenance Rule A1 Condition	The inspectors identified a noncited violation of 10 CFR 50.65(a)(2) for the licensees failure to effectively monitor the performance of the Unit 2 4160Vac Class 1E system. On August 30, 2007, an undervoltage Agastat relay on the Unit 2 4160Vac Train A bus failed. The inspectors determined that this failure should have been recorded as a maintenance preventable functional failure, which would have caused the system to be placed into the Maintenance Rule A1 category. The reason for not recording this failure as a maintenance preventable functional failure was the improper use of the as-found condition codes. The licensee has captured this event under Condition Report 09-2891. This finding was more than minor because it affected the equipment performance attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the reliability of systems that respond to initiating events to prevent undesirable consequences. Using the Significance Determination Process Phase 1 worksheet, this finding was determined to have very low safety significance because it did not result in the actual loss of safety function of one or more trains and did not screen as risk-significant due to seismic, flooding, or severe weather. This finding had a human performance crosscutting aspect associated with work practices because workers failed to ensure proper documentation of activities (H.4(a)) (Section 1R12)
05000498/FIN-2009003-02	2009Q2	South Texas	Failure to Perform Radiation Surveys	A self-revealing noncited violation of 10 CFR 20.1501(a) was identified for failure to perform a radiological survey to determine the potential radiological hazards present when deposting a high contamination area. On October 25, 2008, decontamination technicians were sent into the reactor containment building to remove the decontamination tent from steam generator eddy current testing which was posted as a high contamination area. The technicians were not informed of the expectation to decontaminate the scaffolding and health physics personnel did not follow-up and perform surveys of the deposted area. Subsequently, carpenters were sent in to remove the scaffolding which was still highly contaminated. The licensee was made aware of the situation when one of the carpenters alarmed the personnel contamination monitor and a whole body count revealed approximately 3 millirem intake. The issue was entered into the licensees corrective action program as Condition Report 08-16599. The failure to perform surveys necessary to support deposting a contamination area is a performance deficiency. The finding was greater than minor because it was associated with the Occupational Radiation Safety cornerstone attribute (exposure control) of program and process and affected the cornerstone objective, in that, failure to conduct a radiation survey resulted in unplanned and unintended dose to personnel. Using the Occupational Radiation Safety Significance Determination Process, the finding was determined to be of very low safety significance because it was not an as low as is reasonably achievable finding, there was no overexposure or substantial potential for an overexposure, and the ability to assess dose was not compromised. The finding was self-revealing because the licensee was alerted to the situation when the worker could not pass the personnel contamination monitor. Additionally, this finding had human performance crosscutting aspects associated with work control, in that, the work planning did not appropriately plan work activities by incorporating risk insights and radiological safety (H.3(a)) (Section 2OS1)
05000498/FIN-2009003-03	2009Q2	South Texas	Failure to Ensure a Reliable Fire Suppression Water Supply system	The inspectors identified a noncited violation of License Condition 2.E, Fire Protection, for failure to ensure that equipment required for post-fire safe shutdown system remains free of fire damage. Specifically, the licensee credited manual actions to mitigate the effects of fire damage in lieu of providing the physical protection required by 10 CFR Part 50, Appendix R, Section III.G for the two series-connected volume control tank outlet valves (motor-operated Valve 112B and motor-operated Valve 113A). Failure to ensure that the volume control tank outlet valves relied upon for achieving post-fire safe shutdown were protected from fire damage was a performance deficiency. This finding is of greater than minor safety significance because it impacted the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to external events (such as fire) to prevent undesirable consequences. Specifically, 13 fire areas contain unprotected cables that had the potential to spuriously close at least one of the volume control tank outlet valves which could result in a loss of suction and damage to the only charging pump credited for post-fire safe shutdown. Based on the senior reactor analyst Phase 3 analysis of the Significance Determination Process, this finding was determined to have very low safety significance
05000498/FIN-2009003-04	2009Q2	South Texas	Potential Loss of Centrifuge Charging Pump Suction Due to Fire Damamge	The inspectors identified a noncited violation of License Condition 2.E, Fire Protection, for the failure to ensure that a fire pump would automatically start upon low pressure in the fire main in the event of a fire in the electrical auxiliary building. The team determined that cables for all three fire pumps were routed together in the same cable trays. As a result, a single fire could result in the failure of all three fire pumps to start automatically or manually from the control room. A fire pump could be started locally to restore the water supply, but the delay would reduce the effectiveness of the fire suppression systems in limiting the growth of a fire and minimizing damage to safety-related equipment. The licensee entered this issue into the corrective action program as Condition Report 08-9589. Failure to ensure that a fire pump would automatically start upon low pressure in the fire main in the event of a fire is a performance deficiency. This finding is more than minor because it is associated with the Protection Against External Events attribute of the Mitigating Systems cornerstone and could affect the availability, reliability, and capability of systems that respond to initiating events (such as fire) to prevent undesirable consequences. Based on the senior reactor analyst Phase 3 analysis of the Significance Determination Process, and Inspection Manual Chapter 0609, this finding was determined to have very low safety significance (Section 4OA5.1)
05000498/FIN-2009003-05	2009Q2	South Texas	Licensee-Identified Violation	Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, provides, in part, that procedures shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished. Contrary to this requirement, on February 17, 2009, the licensee failed to verify that the new oil heater thermostat, with the new oil temperature setpoint per Design Change Package 08-13702-21, had appropriate postmaintenance testing to ensure that the heaters would maintain the new required temperature. The postmaintenance test did not have the technicians verify that the heaters would cycle in the correct band. This was identified in the licensees corrective action program as Condition Report 09-2976. This finding is of very low safety significance because the heater control circuit is only an operability factor when the machine is in a standby condition and the safety function was never lost
05000498/FIN-2009003-06	2009Q2	South Texas	Licensee-Identified Violation	Title 10 CFR 50.65(a)(1) requires, in part, that the licensee shall monitor the performance or condition of structures, systems, or components within the scope of the monitoring program as defined in 10 CFR 50.65 (b) against licenseeestablished goals, in a manner sufficient to provide reasonable assurance that such structures, systems, or components are capable of fulfilling their intended functions. Title 10 CFR 50.65(a)(2) states, in part, that monitoring as specified in 10 CFR 50.65(a)(1) is not required where it has been demonstrated that the performance or condition of a structure, system, or component is being effectively controlled through performance of appropriate preventive maintenance, such that the structure, system, or component remains capable of performing its intended function. Contrary to the above, the licensee failed to demonstrate that performance of the Unit 2 electrical auxiliary building ventilation system was being effectively controlled through the performance of appropriate preventive maintenance, in that after a repetitive maintenance preventable failure of the Unit 2 electrical auxiliary building ventilation Train B smoke purge inlet damper occurred on January 27, the licensee failed to consider placing the system under 10 CFR 50.65 (a)(1) for establishing goals and monitoring against the goals. Using the Significance Determination Process Phase 1 worksheets from Inspection Manual Chapter 0609, the inspectors determined that this violation had very low safety significance (Green) because it did not result in the actual loss of safety function of one or more trains and did not screen as risk-significant due to seismic, flooding, or severe weather. The licensee captured this violation in Condition Report 09-1508
05000498/FIN-2011006-01	2011Q3	South Texas	Failure to Timely Correct Conditions Adverse to Fire Protection	The team identified a noncited violation of License Condition 2.E for the failure to implement and maintain in effect all provisions of the approved fire protection program. Specifically, the team identified two examples of failure to implement timely corrective actions to correct conditions adverse to fire protection. The first example related to making Procedure OPOP04-Z0-0001, Control Room Evacuation, Revision 33, consistent with the post-fire safe shutdown analysis in order to ensure the actions met critical time requirements. The second example related to not correcting a condition that could disable all three fire pumps simultaneously as a result of fire damage. Failure to implement timely corrective actions in two instances for conditions adverse to fire protection is a performance deficiency. Both examples of this finding are of greater than minor significance because they impacted the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events (fire) to prevent undesirable consequences. A senior reactor analyst performed Phase 3 significance determination for both examples. The analyst calculated the risk associated with the first example for the actions taken outside the control room as 2.702E-7. For the second example, the analyst assumed that a fire in Fire Area 67 would damage the electrical control cables for all three fire pumps and require manually starting a fire pump at the fire pump house. However, it was determined that a delay in fire suppression because of the need to use a fire hose would not result in a plant transient, require evacuation of the control room, or result in damage to any systems and components required for post-fire safe shutdown. Therefore, the senior reactor analyst determined that both examples of this finding are of very low safety significance (Green). The licensee entered this deficiency into the corrective action program as Condition Record 11-10905. These examples of the performance deficiency had a crosscutting aspect in the area of human performance associated lvith resources because the licensee did not ensure that resources assigned to correct these deficiencies were adequate to assure nuclear safety. Specifically, the licensee failed to ensure adequate design margins by (1) failing to ensure that operators could perform all necessary manual actions prior to exceeding the regulatory requirements and (2) failing to modify the control circuits for the fire pumps to protect them against fire damage.
05000498/FIN-2014008-01	2014Q3	South Texas	Inadequate Loop Flow Test	The team identified a non-cited violation of Technical Specification 6.8.1.d for the failure to implement and maintain written procedures for fire protection program implementation. Specifically, the licensee failed to have procedures for and to flow test the portions of the underground piping that supplied water to the diesel generator buildings since the initial startup test. The licensee initiated actions to perform the flow testing within two months and entered the deficiency into their corrective action program as Condition Report 14-17098. The failure to conduct flow testing of the entire underground fire protection piping loop was a performance deficiency. This performance deficiency was more than minor because it was associated with the protection against external factors attribute (fire) and adversely affected the mitigating systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to performance test the underground fire protection piping loops supplying the emergency diesel generator buildings for both units did not demonstrate the continued capability to deliver adequate flow and pressure to the fire suppression systems supplying those buildings. The team evaluated the finding using Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, because it affected fire protection defense-in-depth strategies involving fire water supply. Using Appendix F, the team determined that the finding screened to very low safety significance. Specifically, the finding did not prevent the reactor from achieving safe shutdown since only one safe shutdown train would be affected at a time. Since these underground fire protection piping loops had not been flow tested since initial installation and nothing caused the licensee to reevaluate the test, the team determined that this failure did not reflect current performance.
05000498/FIN-2014008-02	2014Q3	South Texas	Licensee-Identified Violation	On October 31, 2013, Licensee Event Report 05000498; 05000499/2013-003 described an unanalyzed condition wherein a fire in non-safety-related control room instruments could cause a secondary fire because of a lack of circuit protection. License Condition 2.E specifies, in part, that the licensee implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report through Amendment No. 55 and the Fire Hazards Analysis Report through Amendment No. 7. Section 1.5.2 of the Fire Hazards Analysis Report states that only one fire originating from a single source is assumed to occur at a time. Contrary to the above, since initial construction, the licensee failed to implement the fire protection program that ensured that only one fire originating from a single source will occur at one time. Specifically, the licensee failed to ensure that direct current ammeter circuits were properly protected to prevent secondary fires from initiating in other areas of the plant during a control room fire event. The performance deficiency was more than minor because, if left uncorrected, it could have become a more risk significant fire concern. Since this involved a control room fire, a senior reactor analyst performed a detailed risk evaluation that concluded the fire had very low risk significance. The licensee documented this issue in their corrective action program as Condition Report 13-11633. This violation is also discussed in Section 4OA3.
05000528/FIN-2010007-01	2010Q1	Palo Verde	Impact of recent operating experience on Metal-Enclosed Bus program	The Metal-Enclosed Bus program was a new program, consistent with the GALL Report, credited with managing the aging affects associated with loosening of bolted bus bar connections and reduced insulation and insulator resistance on bus ducts. Metal-enclosed buses are electrical buses installed on electrically insulated supports enclosed in a metal duct. The parameters monitored include loose connections, embrittlement, cracking, melting, swelling or discoloration of insulation, loss of material of bus enclosure assemblies, hardening of boots and gaskets, and cracking of internal bus supports. The applicant included these non safety-related bus ducts because the ducts supported compliance with the station blackout regulations. The applicant scoped in the following non-segregated bus ducts (all conductors enclosed in a common metal enclosure) listed below:1,2,3 NBNA031,2,3 NBNA041,2,3 NBNA051,2,3 NBNA061,2,3 NBNA081,2,3 NBNA09Secondary of Engineered Safety Features Service Transformer E-NBN-X03 to NBN-A08 and NBN-A09Secondary of Engineered Safety Features Service TransformerE-NBN-X04 to NBN-A05 and NBN-A06Connection between Bus NBN-A04 and Vital Bus E-PBA-S03Connection between Bus NBN-A04 and Vital Bus E-PBB-S04Connection between Bus NBN-A03 and Vital Bus E-PBA-S03Connection between Bus NBN-A03 and Vital Bus E-PBB-S04The team reviewed license renewal program basis documents, aging management review documents, existing and new procedures, and preventive maintenance requirements. The team interviewed the license renewal project personnel and the responsible plant and design engineers. The team walked down all the in-scope non-segregated bus ducts. The applicant had developed Procedure 82DP-OEE01, Electrical Aging Management, Draft B, which included qualitative and quantitative acceptance criteria for the inspection of metal enclosed bus internal assemblies. The procedure established the following requirements to monitor for aging effects: (1) inspect a sample of the accessible bolted connections for evidence of overheating; (2) conduct contact resistance testing of a sample of the accessible splice plates to check for loose connections; (3) inspect each bus section for cracks, corrosion, foreign debris, excessive dust buildup, and evidence of water intrusion; (4) inspect the bus insulation for signs of embrittlement, cracking, melting, swelling, or discoloration; and (5) inspect the internal bus supports for structural integrity and signs of cracking. The procedure used visual inspection of bus enclosure assemblies to monitor for loss of material resulting from corrosion and hardening of boots and gaskets. The applicant will allow no unacceptable indications of cracking and will use thermography to determine if excessive heating occurred. The metal enclosed bus inspections will be completed before the period of extended operation and every 10 years as specified by the GALL Report. The team determined that the applicant had an existing preventive maintenance program to inspect and clean bus ducts. The applicant inspected sections of the metal enclosed buses every other outage and performed thermography on the bus at the transformer connections once every 6 months. At the time of the onsite inspection, from review of the corrective action database, the team determined that the applicant had not found any occurrences of corrosion, loss of material, hardening, foreign debris, excessive dust buildup, water intrusion, or overheating that changed the scope of this aging management program. The team did review onsite operating experience that reinforced the need for these inspections. The previous operating experience related to Noryl insulation failures and ground faults on non-segregated bus ducts of components not within the scope of license renewal that resulted in ground faults on the affected buses. Subsequent to the onsite inspection, on March 7, 2010, Unit 1 tripped when a ground fault occurred on Bus 1E-NAN-A03 connecting the non-class 1E 13.8 kV Bus 1E-NAN-S01 to Breaker 1E-NAN-S03B. Because of heavy rains, water intruded past the cover protecting the non segregated bus. The applicant preliminarily concluded the fault resulted from an improperly installed bus duct cover and a breakdown in theNoryl insulation because of aging mechanisms. Because additional information was needed to determine whether a recent age-related failure would impact the proposed aging management program, this is an unresolved item: Unresolved Item 05000528, 05000529, 05000530/2010007-01, Impact of recent operating experience on Metal-Enclosed Bus program. For the Metal-Enclosed Bus program, although some questions remain related to the impact on this program of recent operating experience, the team concluded that the applicant had performed appropriate evaluations and considered pertinent industry experience and plant operating history to determine the effects of aging on the metal enclosed non-segregated bus ducts. The team concluded that, if implemented as described, the applicant provided guidance to appropriately identify and address aging effects during the period of extended operation
05000528/FIN-2010007-02	2010Q1	Palo Verde	Structures monitoring frequency and scope	The Structures Monitoring program was an existing program, consistent with the GALL Report after enhancement, credited with managing cracking, loss of material, and change in material properties of concrete structures and structural supports. The applicant used guidance contained in Procedure 81 DP-OZZ01, Civil System and Component Monitoring Program, to conduct structural inspections. The applicant will enhance the program to require that inspection personnel meet the requirements of American Concrete Institute 349.3R-96 related to qualifications of inspection personnel. The Structures Monitoring program provided inspection guidelines for concrete elements, structural steel, masonry walls, structural features (e.g., caulking, sealants, roofs, etc.), structural supports, and miscellaneous components such as doors. The Structures Monitoring program included all masonry walls and water-control structures within the scope of license renewal. The Structures Monitoring program also monitored settlement of each major structure and inspected supports for equipment, piping, conduit, cable tray, metal-enclosed bus, heating ventilation and air conditioning, and instrument components. The team reviewed the license renewal program basis documents, aging management review documents, and existing procedures. The team searched the corrective action database for relevant condition reports and evaluated the licensee evaluation of operating experience. The team interviewed system and civil engineers involved with performing the inspections and conducted detailed system walkdowns. The team was concerned that the applicant had established an inspection frequency that was not consistent with the frequencies specified in American Concrete Institute 349.3R-96 and the GALL Report. These standards specified that structures exposed to the elements and inside primary containment should be inspected every5 years and that interior structures should be inspected every 10 years. The applicant\'s Structural Monitoring program specified evaluating a representative sample of structures, systems, and components at each of the three units, such that the equivalent of one complete unit is inspected every 10 years. Based on this process, all areas in the three units will have undergone 100 percent inspection once in a 30-year period. The basis for this frequency was documented in procedure 81 DP-OZZ01, Civil System and Component Monitoring Program, which stated that the relative age of the structures at the time of the Maintenance Rule program implementation, the benign environment associated with the site, and having three identical units allowed for an extended inspection frequency. The team was concerned that the licensee had not completed a complete inspection of these structures to use as a basis for justifying an extended inspection schedule. The team noted that the NRC aging management program audit had raised similar questions in Request for Additional information 82.1.32-1. The applicant planned to evaluate whether they should have performed a complete baseline inspection of the facility structures, as documented in Condition Report Disposition Record 3441953.in response to the team\'s concern, the applicant provided a supplemental response to Request for Additional Information B2.1.32-1, as documented in License Renewal Application Amendment 13, dated April 1, 2010. The applicant revised Commitment 34in Table A.4-1, License Renewal Commitments, to state that prior to the period of extended operation, the Palo Verde Structures Monitoring program will be enhanced to inspect structures within the scope of license renewal at 1O-year intervals. The team concluded that the applicant had provided an appropriate response for internal structures, excluding primary containment. However, the team continued to have concerns with the applicant\'s assessment regarding external structures and structures inside containment. The applicant planned to supplement their response to Request for Additional Information B2.1.32-1 to provide a detailed technical justification for the inspection interval for structures inside primary containment or exposed to a natural environment, including all environmental factors. Because additional information is needed to determine whether the applicant had established an appropriate program to manage the effects of aging for the condition of external structures and containment internal structures during the period of extended operation, this is an unresolved item: Unresolved Item 05000528,05000529,05000530/2010007-02, Structures monitoring frequency and scope. For the Structures Monitoring program, the team determined that, with the exception of(1) issues related structural monitoring periodicity and methodology and (2) the limited completion of condition monitoring structural inspections to date, the applicant had implemented and performed appropriate evaluations of structures. Further, the applicant had considered pertinent industry experience and plant operating history to determine the effects of aging on plant structures and structural commodities. The team concluded that, if implemented as described and following resolution of the above concerns, the applicant will have provided adequate guidance to identify and address aging effects during the period of extended operation
05000528/FIN-2015008-01	2016Q1	Palo Verde	Inadequate Flow Test Procedure	The team identified a Green non-cited violation of License Conditions 2.C.7, 2.C.6, and 2.F for Units 1, 2, and 3, respectively, because the licensee had not established criteria for determining when a fire main loop had degraded and had not properly tested all portions of the fire main loop. Specifically, the licensee had not established a differential pressure that would initiate actions to evaluate the cause for a degradation and the licensee had not determined the flow through individual flow paths in their auxiliary and control buildings. The licensee documented these issues in Condition Reports 15-00513 and 16-00686 and initiated actions to correct the procedure and perform the flow test of the individual loops. The team identified a performance deficiency related to the procedure used to test their fire main loop. Specifically, the licensee had not established criteria for determining a degraded fire main loop and had not properly tested all portions of the fire main loop. This performance deficiency was more than minor because it was associated with the protection against external factors attribute (fire) and adversely affected the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to test the fire main loops inside the control/auxiliary building separately and failure to establish appropriate acceptance criteria affected the ability to demonstrate the continued capability to deliver adequate flow and pressure to the fire suppression systems. The finding was screened in accordance with NRC Inspection Manual Chapter (IMC) 0609, Significance Determination Process, Attachment 4, Initial Characterization of Findings, dated June 19, 2012. The inspectors determined that an IMC 0609, Appendix F, Fire Protection Significance Determination Process, dated September 20, 2013, review was required as the finding affected the ability to reach and maintain safe shutdown conditions in case of a fire. Using IMC 0609, Appendix F, Attachment 1, Fire Protection Significance Determination Process Worksheet, dated September 20, 2013, the finding was screened as a Green finding of very low safety significance in accordance with Task 1.4.7, Fire Water Supply, Question A. The inspectors determined that although the licensee failed to test portions of the fire main system in accordance with code requirements, the inspectors determined that at least 50 percent of required fire water capacity would be available based on the testing is done with only one fire pump in service and there are three available fire pumps. Since these fire main loops inside the control/auxiliary building had not been monitored for pressure changes when flow tested since initial testing and nothing caused the licensee to reevaluate the test, the team determined that this failure did not reflect current performance.
05000528/FIN-2016002-01	2016Q2	Palo Verde	Leakage From Reactor Coolant Pump 2B Discharge Pipe Instrument Nozzle	The inspectors identified an unresolved item for pressure boundary leakage from reactor coolant pump 2B discharge pipe instrument nozzle. On April 10, 2016, during the Unit 1 Refueling Outage 19, the licensee discovered reactor coolant system pressure boundary leakage at instrument nozzle 1JRCETW0121Y on the 2B reactor coolant pump discharge piping. The leakage was discovered during a planned visual inspection of Unit 1 hot and cold leg nozzles. The leak was not detectable by either the reactor coolant system leak rate procedure or the containment radiation monitor trend reviews while the unit was operating. Additionally, the leak had not been visually detected during the previous refueling outage. The leakage was consistent with a small leak characterized by moderate boric acid accumulation at the leakage site. The licensee determined that the cause of the leakage was primary water stress corrosion cracking of the Alloy 600 instrument nozzle. The licensee corrected the leakage using a mechanical nozzle seal assembly repair method utilizing ASME Code Case N-733, Mitigation of Flaws in NPS 2 (DN 50) and Smaller Nozzles and Nozzle Partial Penetration Welds in Vessels and Piping by Use of a Mechanical Connection Modification, Section XI, Division 1. The evaluation of the 2B cold leg RTD nozzle leakage is being evaluated by the licensee as part of Palo Verde Action Request 15-01640-012. The inspectors reviewed the circumstances surrounding the discovery of the leak and observed portions of the repair activity during the refueling outage. Once the licensee completes their evaluation, the inspectors will review and complete an inspection to determine if a performance deficiency exists as a result of the nozzle failure.
05000528/FIN-2016002-02	2016Q2	Palo Verde	Failure to Implement High Radiation Area Controls in an Area with a Dose Rates Greater Than 1 rem per Hour	The inspectors reviewed a Green, self-revealing, non-cited violation of Technical Specification 5.7.2, which was caused by the licensees failure to control a high radiation area with radiation levels greater than 1 rem per hour in the Unit 1 containment. A radiation protection technician received an unexpected dose rate alarm while conducting surveys on piping in the 87-foot elevation of the 2B reactor coolant pump bay area near a high efficiency particulate air unit in containment. Licensee personnel corrected the error by guarding the area, posting the area, and changing the pre-filters in the adjacent portable a high efficiency particulate air units to reduce the dose rates. This issue was entered into the licensees corrective action program as Condition Reports 16-06515 and 16-07479. The inspectors determined that the failure to identify a locked high radiation area through timely surveys and adequate a high efficiency particulate air maintenance procedures that could have revealed changing radiological conditions was a performance deficiency. The performance deficiency was more than minor because it was associated with the Occupational Radiation Safety Cornerstone attribute of program and process (exposure control) and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation because licensee personnel did not implement barriers intended to prevent workers from receiving unexpected dose. Using Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008, the inspectors determined the violation had very low safety significance (Green) because: (1) it was not an as low as is reasonably achievable finding, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. This finding has a cross-cutting aspect in the human performance area, associated with the resources component, because the licensee leaders failed to ensure that personnel, equipment, and procedures were available and adequate to support nuclear safety. Specifically, the licensee failed to ensure that procedures were adequate to ensure radiation levels around portable high efficiency particulate air units were monitored to evaluate changing radiological conditions in a timely manner such that hazards were appropriately controlled (H.1).
05000528/FIN-2016002-03	2016Q2	Palo Verde	Inadequate Engineering and Radiological Controls Resulting in a Unit 1 Containment Building Airborne Radioactivity Event with Unplanned Intakes	The inspectors identified a non-cited violation of 10 CFR 20.1701 due to the licensees failure to implement adequate processes and engineering controls necessary to reduce airborne radioactivity and prevent internal dose to workers in Unit 1. On April 20, 2016, inspectors identified that procedures and instructions for monitoring high efficiency particulate air ventilation filter unit to prevent worker exposures to radiation and airborne radioactivity were being inadequately implemented. On April 21, 2016, the licensees inadequate engineering and radiological controls during a high efficiency particulate air operations caused an airborne radioactivity event in containment, resulting in the evacuation of 41 potentially contaminated workers of whom 8 had measurable intakes of radioactive material. The licensees immediate corrective actions included stopping work in the Unit 1 containment, evacuating workers in containment, assessing workers for external and internal contamination, and investigating the cause and source of the contamination event. This matter was placed in the licensees corrective action program as Condition Reports16-06499 and 16-06578 and the licensee initiated a root cause investigation. The inspectors determined that the failures to implement adequate engineering and radiological controls to reduce airborne radioactivity during a high efficiency particulate air unit operations in accordance with 10 CFR 20.1701 and radiation protection procedures were performance deficiencies. The performance deficiencies were more than minor because they were associated with the Occupational Radiation Safety Cornerstone attribute of program and process (exposure control) and adversely affected the cornerstone objective to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. This was evident by the Unit 1 containment airborne radioactivity event on April 21, 2016, that resulted in at least eight workers with unplanned intakes. Using Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008, the inspectors determined the finding had very low safety significance (Green) because: (1) it was not an as low as is reasonably achievable planning and controls finding, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. The inspectors concluded that the finding has a cross-cutting aspect in the human performance area, associated with the resources component, because the licensee leaders failed to ensure that personnel, equipment, procedures, and other resources were available and adequate to support nuclear safety. Specifically, procedures and radiation exposure permits failed to have adequate instructions for ensuring a high efficiency particulate air filter loading and dose rates were monitored to prevent overloading, and safe handling of loaded a high efficiency particulate air filters (H.1).

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