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	Quarter	Site	Title	Description
05000445/FIN-2018010-01	2018Q3	Comanche Peak	Failure to Establish Test Program to Verify Residual Heat Removal Suction Valve Capability	The inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for the failure to establish a test program to ensure that residual heat removal suction isolation valves would perform adequately in service.
05000445/FIN-2018010-02	2018Q3	Comanche Peak	Failure to Provide Procedural Guidance for the Failure of a Component Cooling Water Surge Tank Makeup Valve	The inspectors identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to provide procedural guidance for the failure of a component cooling water surge tank makeup valve.
05000416/FIN-2018002-08	2018Q2	Grand Gulf	Performance of Surveillance Testing Following Maintenance on Containment Airlock	The inspectors identified a Green non-cited violation of 10CFRPart50,AppendixB, Criterion XI, Test Control, for the licensees failure to perform surveillance testing of containment airlock seals under appropriate conditions. The licensee failed to appropriately control the sequence of maintenance and testing activities to ensure that surveillance testing was not performed subsequent to maintenance which could affect the validity of surveillance test results.
05000416/FIN-2018002-07	2018Q2	Grand Gulf	Loss of Shutdown Cooling	A self-revealed,Green non-cited violation of Technical Specification 5.4, Procedures,for the licensees failure to follow written procedures was identified when the residual heat removal (RHR) system automatically isolated due to an inadvertent emergency core cooling system (ECCS) actuation. While the plant was shut down with the RHR system in decay heat removal mode, maintenance personnel inadvertently opened an incorrect valve during a transmitter calibration activity, which caused a false low reactor pressure vessel (RPV) water level signal, an ECCS actuation, and a loss of decay heat removal for approximately 31 minutes
05000416/FIN-2018002-06	2018Q2	Grand Gulf	Improper Evaluation and Resolution of Intermediate Range MonitorNoise Leads to Manual Reactor Shutdown	A self-revealed, Green non-cited violation of 10CFRPart50, Appendix B, Criterion XVI, Corrective Action, was identified for the failure of the licensee to identify and correct a condition adverse to quality. Specifically, the licensee failed to implement appropriate corrective actions related to intermediate range monitor (IRM) nuclear instrument (NI) electronic noise spiking. The failure to implement adequate corrective actions over the course of at least 5 years resulted in a plant shutdown due to declaration of multiple IRM channels inoperable while in Mode 2.
05000416/FIN-2018002-05	2018Q2	Grand Gulf	Failure to Follow Procedure Requirements Resulting in Unplanned Dose	A self-revealed, Green non-cited violation of Technical Specification 5.4.1 was identified when an individual alarmed a personnel contamination monitor upon exit from the radiologically controlled area. Specifically, the licensee failed to follow procedures to establish a decontamination plan or procedure, conduct a specific pre-job brief addressing appropriate contamination risk, and receive approval by radiation protection supervision prior to conducting decontamination activities on thereactor pressure vessel(RPV) O-rings
05000416/FIN-2018002-04	2018Q2	Grand Gulf	High Radiation Area Boundary Violation	A self-revealed, Green non-cited violation of Technical Specification 5.7.1 was identified when an individual received a dose rate alarm when the individual failed to comply with established radiological barriers and protective measures and entered a high radiation area. Specifically, an individual leaned over a high radiation area barricade rope, thereby entering the high radiation area. The individuals radiation work permit (RWP) did not permit entry into a high radiation area.
05000416/FIN-2018002-03	2018Q2	Grand Gulf	Failure to Adequately Test NUS Temperature Switch	A self-revealed,Green non-cited violationof 10CFRPart50, AppendixB, CriterionIII, Design Control, was identified when the reactor core isolation cooling (RCIC) system automatically isolated due to an inadvertent high temperature input from the leakage detection system. Specifically, the licensee failed to fully test a modification that installed a new type of temperature switches, and the system inappropriately isolated the RCIC system when a loss and subsequent restoration of power occurred.
05000416/FIN-2018002-02	2018Q2	Grand Gulf	Failure to Follow ASME Requirements for Maintaining Inservice Inspection (ISI) Cycles and Perform ASME Required Inservice Inspections within the Scheduled ISI Cycle	The inspector identified 15 examples of a Green non-cited violation (NCV)of 10 CFR 50.55(a)(g)(4)(ii), which requires that inservice examination of components classified as American Society of Mechanical Engineers (ASME), Section XI, Code Class 1, Class 2, and Class 3 be conducted during successive 120-month inspection intervals, and requires compliance with the requirements of the latest edition and addenda of the ASME Code (and all its paragraphs) applicable to the specific interval, including maintaining the 120-month inspection interval in accordance with the ASME Code, Section XI, Paragraph IWA-2430. Specifically, the licensee inappropriately adjusted its second inservice inspection 120-month cycle, and failed to perform VT-3 and MT examinations of 15 class 1, class 2, and class 3 components, including the high pressure core spray pump attachment weld and reinforcing band before the third inservice inspection cycle expired on November 30, 2017, as required by 10CFR50.55(a)(g)(4)(ii).
05000416/FIN-2018002-01	2018Q2	Grand Gulf	Failure to Institute Effective Corrective Action to Preclude Repetition	An NRC-identified,Green non-cited violation of 10CFRPart50, AppendixB, CriterionXVI, Corrective Action, was identified when the licensee failed to institute effective corrective actions to preclude repetition of a significant condition adverse to quality. Specifically, the licensee left a secondary containment personnel hatch in an open configuration for approximately 30 minutes while performing a roof inspection, which rendered secondary containment inoperable. This issue had also previously occurred in 2016, but corrective actions to prevent it from occurring again were ineffective.
05000529/FIN-2017001-01	2017Q1	Palo Verde	Failure to establish station procedure instructions for denial work authorizations	The inspectors identified a Green, non-cited violation of Technical Specification 5.4.1.a, Procedures, for the failure to establish procedure instructions for work authorization denials or deferrals. Specifically, this led to a 60 day extended unavailability of the diverse auxiliary feedwater actuation system when corrective maintenance was inappropriately deferred by the operations department. Failure to provide adequate procedural guidance in the event of a denied work authorization, a circumstance anticipated to occur, is a performance deficiency. The performance deficiency is more than minor, because it affected the equipment performance attribute of the Mitigating Systems Cornerstone objective to ensure the availability and reliability of equipment that responds to an initiating event. Specifically, because the corrective maintenance was not performed in a timely manner, both trains of the diverse auxiliary feedwater actuation system remained in bypass for an additional 60 days whereby the system was not capable of performing its required safety function. The inspectors evaluated the significance of the finding using Inspection Manual Chapter 0609, Appendix A, Significance Determination Process for Findings at Power, Exhibit 2, Mitigating Systems Screening Questions, Section A, Question 2, which required a detailed risk evaluation because the finding involved a loss of system safety function. A Region IV senior reactor analyst performed a detailed risk assessment of the finding and determined that the finding was of very low safety significance (Green). The inspectors determined that the finding had a cross-cutting aspect in the human performance area of Work Management. The work process includes the identification and management of risk commensurate to the work and the need for coordination with different groups or job activities. Specifically, the Unit Operations Managers decision to deny the work authorization was based on conservative but faulty assumptions, and if other work groups with greater specific technical knowledge had been involved, the corrective maintenance likely would have proceeded (H.5)
05000275/FIN-2016302-01	2016Q4	Diablo Canyon	Licensee-Identified Violation	A severity level IV violation that was identified by the licensee has been reviewed by the examiners. Corrective actions taken or planned by the licensee have been entered into the licensees corrective action program. The violation and corrective action tracking number are listed in Section 4OA7 of this report. Title 10 CFR 50.9(a) requires, in part, that information provided to the Commission by a licensee shall be complete and accurate in all material aspects. On October 4, 2016, the NRC gave approval to the licensee to administer a written examination to initial operating license applicants on October 14, 2016. The approval was made based on content of the written examination provided to the NRC on October 4, 2016. In this version of the written examination, Question 55 had been revised based on NRC comments so that it had only one correct answer. The previous draft revision of the question had two plausible correct answers. The written examination was administered on October 14, 2016. During licensee review of the exam, the licensee identified that the version of Question 55 on the administered written examination wa not the version that was approved on October 4, 2016. The licensee notified the NRC of the issue on November 7, 2016, and completed an extent of condition review that showed that this was the only written examination question inconsistent with the questions approved on October 4, 2016. The violation was of very low safety significance because the performance deficiency did not contribute to the NRC making any incorrect regulatory decisions regarding issuance of operating licenses.
05000529/FIN-2016004-01	2016Q4	Palo Verde	Inadequate monitoring of MSIV nitrogen pre-charge pressure	The inspectors reviewed a self-revealing non-cited violation of Technical Specification 3.7.2 for exceeding the Condition A completion time for an inoperable main steam isolation valve (MSIV) single actuator train and not immediately declaring the affected main steam isolation valve inoperable in accordance with Condition E. Specifically, the Unit 2 main steam isolation valve 171 actuator A was inoperable from July 30, 2016, to August 9, 2016, when a known nitrogen leak was not adequately monitored. The licensees inadequate monitoring allowed the nitrogen pre-charge pressure in the actuator to decrease to below the minimum acceptable limit for operability. The licensee restored the pre-charge pressure and entered this issue into their corrective action program as Condition Report 16-12740. The failure to perform adequate monitoring for a degraded condition as required by procedure 40DP-9OP26, Operations Condition Reporting Process and Operability Determination/Functional Assessment, was a performance deficiency. The performance deficiency was more-than-minor and therefore a finding because it affected the equipment performance attribute of the Mitigating Systems Cornerstone to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically the failure to adequately monitor a known nitrogen leak resulted in depressurizing one of two hydraulic accumulators thereby reducing the reliability of the system to initiate a fast closure of MSIV 171 upon receipt of a main steam isolation signal. The inspectors performed the initial significance determination using NRC Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, Issue Date: June 9, 2012. The finding required a detailed risk evaluation since it represented a loss of function for a single train for greater than the Technical Specification allowed outage time. A Region IV senior reactor analyst determined the finding was of very low safety significance (Green) since the MSIV remained capable of performing its safety function with the alternate actuator. The finding has a cross-cutting aspect in the area of human performance associated with the teamwork component. Specifically, the licensee failed to coordinate activities across organizational boundaries in that the operations personnel did not obtain engineering input to ensure that additional monitoring requirements for the nitrogen pre-charge leak were adequate to verify continued MSIV 171 operability (H.4).
05000313/FIN-2016003-04	2016Q3	Arkansas Nuclear	EA-16-143, Enforcement Discretion for Tornado-Generated Missile Protection Noncompliances	Appendix A to 10 CFR 50, General Design Criteria for Nuclear Power Plants, Criterion 2, Design Bases for Protection Against Natural Phenomena, states, in part, that SSCs important to safety shall be designed to withstand the effects of natural phenomena, such as tornadoes. Criterion 4, Environmental and Dynamic Effects Design Basis, states, in part, that SSCs important to safety shall be appropriately protected against dynamic effects including missiles which may result from events and conditions outside the nuclear power unit. As part of their response to external flood boundary degradation, the licensee performed a review of external hazard protection at the site, which included protection against tornado-generated missiles required by the current licensing basis for each unit. During the review, on four separate occasions, the licensee identified plant areas containing safety-related SSCs that could be susceptible to tornado missiles: Unit 1 Upper South Electrical Penetration Room Unit 1 Cable Spreading Room Unit 1 Controlled Access Area Unit 1 Vital Switchgear In each case, the licensee identified low-probability scenarios where one or more tornado-generated missiles could penetrate doors, walls, and other building features that were not fully qualified, and subsequently damage equipment that was important to safety inside the affected rooms. Details about the date of discovery, affected SSCs, condition report numbers, compensatory actions taken by the licensee, notifications made to the NRC, and affected technical specification actions for each susceptible area are listed in Attachment 3 of this report. Relevant Enforcement Discretion Policy On June 10, 2015, the NRC issued Enforcement Guidance Memorandum (EGM) 15-002, Enforcement Discretion for Tornado-Generated Missile Protection Noncompliance. (ML15111A269) The EGM referenced a bounding generic risk analysis performed by the NRC staff that concluded that tornado missile vulnerabilities pose a low risk significance to operating nuclear plants. Because of this, the EGM described the conditions under which the NRC staff may exercise enforcement discretion for noncompliances with the current licensing basis for tornado-generated missile protection. Specifically, if the licensee could not meet the technical specification required actions within the required completion time, the EGM allows the staff to exercise enforcement discretion provided the licensee implements initial compensatory measures prior to the expiration of the time allowed by the limiting condition for operation. The compensatory actions should provide additional protection such that the likelihood of tornado missile effects are lessened. The EGM then requires the licensee to implement more comprehensive compensatory measures within approximately 60 days of issue discovery. The compensatory measures must remain in place until permanent repairs are completed, or until the NRC dispositions the non-compliance in accordance with a method acceptable to the NRC such that discretion is no longer needed. In addition, the issue must be entered into the licensees corrective action program. Because EGM 15-002 listed Arkansas Nuclear One as a Group A plant, enforcement discretion will expire on June 10, 2018. However, the EGM did not provide for enforcement discretion for any related underlying technical violations; the EGM specifically requires that any associated underlying technical violations be assessed through the enforcement process. Licensee Actions For each of the examples listed above, the licensee declared the affected systems inoperable and complied with the applicable technical specification action statement(s), initiated a condition report, invoked the enforcement discretion guidance, implemented prompt compensatory measures, and returned the SSCs to an operable status. The licensee instituted compensatory measures intended to reduce the likelihood of tornado missile effects that included developing actions to be taken if a tornado watch is predicted or issued for the area to ensure the operability or restore redundant equipment during severe weather, and actions to be taken if a tornado warning is issued, including pre-staging operators in safe, strategic locations to promptly implement mitigative actions, and verifying the readiness of equipment and procedures dedicated to the Diverse and Flexible Coping Strategy (FLEX). Other specific compensatory actions for the individual areas are listed in Attachment 3. NRC Actions The inspectors review addressed the material issues in the plant, and whether the measures were implemented in accordance with the guidance in EGM 15-002. The inspectors also evaluated whether the measures would function as intended and were properly controlled. The inspectors verified through inspection that the EGM 15-002 criteria were met in each case. Therefore, the staff determined that it was appropriate to exercise enforcement discretion and not take enforcement action for the technical specification requirements listed in Attachment 3 of this report, provided the noncompliances are resolved by June 10, 2018 (EA-16-143). The inspectors did not fully review the underlying circumstances that resulted in the technical specification violations. As stated in EGM 15-002, violations of other requirements which may have contributed to the technical specification violations will be evaluated independently of EGM implementation. The inspectors will verify restoration of compliance and assess the underlying circumstances in a follow-up inspection tracked under Licensee Event Reports 05000313/2016-002-00 and 05000313/2016-003-00, and any updates or additional licensee event reports that the licensee issues.
05000275/FIN-2016301-01	2016Q2	Diablo Canyon	Insufficient Procedural Direction Contained Within Procedure EOP E-2, Faulted Steam Generator Isolation	The examiners identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings. Specifically, Procedure EOP E-2, Faulted Steam Generator Isolation, does not contain sufficient procedural direction for isolating auxiliary feedwater flow to a faulted steam generator in the event that auxiliary feedwater control valves cannot be closed from the control room. Procedure EOP E-2, Appendix HH, Isolated Faulted Steam Generator, Step 1.d, and its associated column, Response Not Obtained, does not ensure that a faulted steam generator would remain isolated under all conditions. The Response Not Obtained column permits operators to either locally close auxiliary feedwater control valves OR secure the auxiliary feedwater pump feeding the faulted steam generator. However, due to the absence of pull-to-lock or hard stop switches for the auxiliary feedwater pumps, the possibility exists for an automatic restart of an auxiliary feedwater pump and a re-initiation of feedwater to a faulted steam generator. The failure to ensure that Procedure EOP E-2 contained sufficient direction to isolate a faulted steam generator when auxiliary feedwater flow control valves cannot be closed from the control room was a performance deficiency. This performance deficiency was of more than minor safety significance because it was associated with the procedure quality attribute of the Barrier Integrity cornerstone (reactor coolant system and containment) and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the re-initiation of feedwater to an isolated, faulted steam generator has the potential to adversely affect the reactor coolant system barrier by causing an additional unintended cooldown of the reactor coolant system, increased potential for pressurized thermal shock, and thermal stress to the steam generator u-tubes. Additionally, the containment barrier would be affected by the reinitiation of feedwater to a steam line break within containment. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, the team determined that the finding required a detailed risk evaluation due to the potential to affect the reactor coolant system boundary. A senior reactor analyst performed a bounding detailed risk evaluation and estimated the maximum increase in core damage frequency to be 5.9E-8/year, and therefore the finding was determined to be of very low safety significance (Green). This increase in core damage frequency was mitigated by the low probability of multiple equipment failures in the auxiliary feedwater system when combined with the low initiating event frequency of a faulted steam generator. Because the violation was of very low safety significance (Green) and the issue was entered into the licensees corrective action program as Notification 50847218, this violation is being treated as a non-cited violation consistent with Section 2.3.2.a of the Enforcement Policy: NCV 05000275/2016301; 05000323/2016301-01, Insufficient Procedural Direction Contained Within E-2, Faulted Steam Generator Isolation. This finding has a crosscutting aspect in the area of human performance associated with resources because the organization did not ensure procedures are available and adequate to support nuclear safety (H.1).
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.
05000369/FIN-2016002-02	2016Q2	McGuire	Failure to Ensure Containment Equipment Hatch Was Properly Closed During Fuel Movements	An NRC-identified Green NCV of Technical Specification (TS) 5.4.1.d, Procedures, was identified for the licensees failure to adequately implement the commitments in Selective Licensee Commitment (SLC) 16.9.25, Refueling Operations Containment Equipment Hatch, which required the containment equipment hatch to be closed during the movement of non-recently irradiated fuel inside containment. Specifically, during reactor vessel fuel reload activities, the inspectors identified that the equipment hatch was left partially open due to the failure to properly tighten the bolts evenly around the hatch resulting in direct communication of the containment atmosphere with the environment. The licensee took immediate corrective action to suspend fuel movements and properly tighten the equipment hatch bolts prior to resuming fuel movements and entered the issue into their corrective action program as ARs 02018605 and 02018701. The PD was more than minor because it impacted the configuration control attribute of the barrier integrity cornerstone and affected the cornerstone objective to provide reasonable assurance that containment protects the public from radionuclide releases caused by accidents or events. Additionally, if left uncorrected, the PD would have the potential to lead to a more significant safety concern. Specifically, the radiological barrier functionality of the containment equipment hatch was degraded due to the gap opening which could have allowed direct access of radiological releases from the containment atmosphere to the outside environment during a potential fuel handling accident inside containment. The inspectors screened the finding in accordance with IMC 0609, Appendix G, Shutdown Operations Significance Determination Process, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings. Because the finding degraded the ability to close or isolate the containment, it required review using IMC 0609, Appendix H, Containment Integrity Significance Determination Process. While the containment boundary function was considered degraded, the incident occurred eight days after the beginning of the refueling outage when short lived volatile radioisotopes had decayed sufficiently such that the potential radiological releases to the public would not likely contribute to the large early release frequency (LERF). Based on this, the finding was screened as having very low safety significance (Green). The cause of the PD was directly related to the cross-cutting aspect of procedure adherence in the cross-cutting area of human performance because the licensee failed to follow containment equipment hatch closing procedures which explicitly required performing a visual inspection that the containment equipment hatch was sealed and secured with metal-to-metal contact with the containment hatch flange and had no visual gaps.
05000369/FIN-2016002-01	2016Q2	Mcguire McGuire	Failure to Perform General Visual Examinations of Containment Moisture Barriers Associated with Containment Liner Leak Chase Test Connections	An NRC-identified Green non-cited violation (NCV) of 10 CFR Part 50.55a, Codes and Standards, was identified for the licensees failure to perform general visual examinations of moisture barrier material in the reactor containment leak-chase channel test connections in accordance with the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME BPV Code), Section XI, Subsection IWE. The licensee performed the required examinations in Unit 1 during the March 2016 refueling outage and initiated corrective actions to revise the Containment Inservice Inspection (ISI) Plan. The licensee also planned to perform similar examinations in Unit 2 prior to the end of the first containment ISI period. Additionally, the licensee performed a containment operability determination to justify continuous operation of the Unit 1 and Unit 2 containment based on the results of all visual examinations, extent of condition activities, and the results of containment integrated leak rate tests. The licensee entered this issue into their corrective action program as action request (AR) 02038505. The failure to conduct the required visual examination of moisture barrier material in accordance with the ASME BPV Code was a performance deficiency (PD). The PD was of more than minor significance per IMC-0612, Appendix B, Issue Screening, because the current Containment ISI Plan did not adequately implement the ASME BPV Code requirements for the examination of moisture barriers, and if left uncorrected, it had the potential to lead to a more significant concern. The finding was of very low safety significance (Green) per IMC-0609 because it did not represent an actual open pathway in the physical integrity of the reactor containment and did not involve an actual reduction in function of hydrogen igniters in the reactor containment. The finding had a cross-cutting aspect of resolution in the problem identification and resolution cross-cutting area because the licensee did not take effective corrective actions to implement the ASME BPV code requirements in the Containment ISI Plan when a reasonable opportunity was available through the review of NRC Information Notice (IN) 2014-07.
05000482/FIN-2016001-01	2016Q1	Wolf Creek	Failure to Adequately Establish and Adjust Preventive Maintenance Activities for Control Room Air Conditioning Unit SGK04A Sensing lines and Fittings	The inspectors identified a Green 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 adequately develop a preventive maintenance replacement task and schedule for control room air conditioning unit SGK04A refrigerant sensing lines and fittings. The licensees immediate actions included securing and declaring the SGK04A system inoperable, completing corrective maintenance to eliminate the refrigerant leak, and confirming that the impacted preventive maintenance frequency was adequately established. The licensee entered this condition into the corrective action program as Condition Reports 101862 and 101867. 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 utilized Inspection Manual Chapter 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 structures, systems, and components (SSC) 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 it 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 finding has a cross-cutting aspect in the area of human performance, resources, because leaders did not ensure that personnel, equipment, procedures, and other resources are available and adequate to support nuclear safety. Specifically, leaders did not ensure procedures and other resource materials were available to support successful work performance when setting preventive maintenance activity base dates, which resulted in the licensee failing to adequately develop and adjust preventive maintenance activities associated with control room air conditioning unit SGK04A refrigerant sensing lines and fittings (H.1).
05000458/FIN-2016009-01	2016Q1	River Bend	Failure to Follow Procedure While Installing Jumpers for Shutdown Cooling	The team reviewed a self-revealing, non-cited violation of Technical Specification 5.4, Procedures, for the licensees failure to correctly implement Procedure SOP-0031, Residual Heat Removal System, Revision 326. SOP-0031, Attachment 5, Step 5.4.1, required that a retractable sheathed banana jumper be used when bypassing the 135-psi SDC isolation. Instead, the licensee used a standard banana jumper, which resulted in a short circuit and inadvertent closure of Valves E12MOV-F008, Shutdown Cooling Suction Valve, and E12MOV-F053A, Shutdown Cooling Injection Valve. This caused a loss of decay heat removal. This issue was entered into the licensees corrective action program as Condition Report CR-RBS-2016-0210. Corrective actions included revising Procedure SOP-0031 to include actions to de-energize the applicable valves while bypassing the 135-psi shutdown cooling isolation. The failure to use the correct jumpers as specified in Procedure SOP-0031 was a performance deficiency. This performance deficiency is more than minor, and therefore a finding, because it is associated with the human performance attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the shorting of contacts resulting from the use of incorrect jumpers caused a loss of shutdown cooling and decay heat removal. The team evaluated the finding using NRC Inspection Manual Chapter 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Screening and Characterization of Findings. When applying Exhibit 2 - Initiating Events Screening Questions, the team determined the loss of residual heat removal event did not occur when the refuel cavity was flooded, and therefore it required a risk evaluation using the Appendix G, Attachment 3, Phase 2 Significance Determination Process Template for Boiling Water Reactors during Shutdown. The analyst determined that a modified but still conservative Phase 2 quantitative estimate in combination with qualitative and deterministic insights led to a final conclusion that the finding was of very low safety significance (Green). The finding has a field presence cross-cutting aspect within the human performance area because the licensee failed to promptly correct deviations from standards and expectations. Specifically, the licensee failed to correct deviations from standards and expectations during the performance of the pre-job brief and ensure proper communication and oversight is maintained in the control room during risk significant evolutions (H.2).
05000458/FIN-2016009-04	2016Q1	River Bend	Failure to Adequately Assess Risk During Motor Generator Set Unavailability	The team identified a non-cited violation of 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, for the licensees failure to adequately assess the increase in risk that may result from proposed maintenance activities. Specifically, the team identified that since 2012, the licensee failed to adequately assess the risk of simultaneously powering both reactor protection system buses from the alternate power sources, which resulted in an increased risk of a reactor scram due to grid instabilities. This issue was entered into the licensees corrective action program as Condition Report CR-RBS-2016-3176. Corrective actions included revising Procedure SOP-0079, Reactor Protection System, to include precautions to address the increased risk associated with supplying both reactor protection system buses from the alternate power source. The team determined that the licensees failure to adequately assess the increase in risk associated with simultaneously powering both reactor protection system buses from the alternate power sources was a performance deficiency. The performance deficiency is more than minor, and therefore a finding, because it is associated with the design control attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the performance deficiency resulted in an increased risk of a reactor scram due to grid instabilities. The team performed an initial screening of the finding in accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using Inspection Manual Chapter 0609, Appendix A, Exhibit 1, Initiating Events Screening Questions, a detailed risk evaluation was required since the finding resulted in a reactor scram and main steam isolation valve closure. The finding was evaluated using Inspection Manual Chapter 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, Flowchart 1, Assessment of Risk Deficit, dated May 19, 2005, to assess the significance of the finding. A senior reactor analyst estimated the incremental core damage probability deficit to be 2.0E-7 and the incremental large early release probability deficit to be 4.0E-8. Since this incremental core damage probability deficit was less than 1E-6 and the incremental large early release probability deficit was less than 1E-7, the analyst used Flowchart 1 to determine the finding was of very low safety significance (Green). This finding has a conservative bias cross-cutting aspect within human performance area because the licensee determined that powering both reactor protection system buses from the alternate source instead of the motor generator sets was safe even though the motor generator sets are the preferred source and provide protection against grid perturbations (H.14).
05000458/FIN-2016009-03	2016Q1	River Bend	Failure to Implement Corrective Actions to Prevent the Recurrence of a Reactor Scram Due to Grid Disturbances	The team reviewed a self-revealing, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to establish measures to assure that corrective action is taken to preclude repetition of a significant condition adverse to quality. Specifically, following a November 27, 2015, reactor scram, the licensee failed to implement corrective actions associated with the alternate power lineup of the reactor protection system buses to preclude repetition of a significant condition adverse to quality during the January 9, 2016, reactor scram. This issue was entered into the licensees corrective action program as Condition Report CR-RBS-2016-0180. Corrective actions included supplying reactor protection system bus A from the normal power source on January 12, 2016. The failure to assure corrective actions are promptly taken for a significant condition adverse to quality to preclude repetition of a reactor scram associated with both buses being affected by a switchyard voltage transient was a performance deficiency. This performance deficiency is more than minor, and therefore a finding, because it is associated with the human performance attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the licensees failed to implement corrective actions to address grid instabilities following the November 27, 2015, reactor scram to preclude the January 9, 2016, reactor scram. The team performed an initial screening of the finding in accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using Inspection Manual Chapter 0609, Appendix A, the team determined that this finding is of very low safety significance (Green) because it did not involve the loss of mitigation equipment or a support system. This finding has an evaluation cross-cutting aspect within the problem identification and resolution area because the licensee failed to thoroughly evaluate the cause of the November 27, 2015, reactor scram and ensure that the resolution addresses causes and extent of conditions commensurate with their safety significance (P.2).
05000445/FIN-2016001-01	2016Q1	Comanche Peak	Failure to Adequately Evaluate Operability for a Degraded Condition	The inspectors identified seven examples of a non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, associated with the licensees failure to perform adequate operability assessments for a degraded or nonconforming condition. Specifically, when vacuum breakers installed in the service water system failed to actuate during surveillance testing, the licensee completed an operability evaluation that relied on judgement, and was contrary to the station design analysis. In particular, the licensee concluded that the vacuum breakers were not required to support operability of the service water system. Following questions from inspectors, the licensee determined that this judgement was not correct and performed a new evaluation to establish operational parameters necessary to ensure operability of the service water system with a failed vacuum breaker. The licensee entered this issue into corrective action program as Condition Report CR-2015-008334. The failure to properly assess and document the basis for operability for a degraded or nonconforming condition was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, service water vacuum breakers failing to open resulted in a condition where structures, systems, and components necessary to mitigate the effects of a column separation event may not have functioned as required. Using Inspection Manual Chapter (IMC) 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, inspectors determined that this finding was of very low safety significance (Green) because the finding (1) was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality, (2) did not represent a loss of system and/or function, (3) did not represent an actual loss of function of at least a single train for longer than its allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time, and (4) does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant for greater than 24 hours in accordance with the licensees maintenance rule program. The inspectors determined that this finding does not have a cross-cutting aspect because the most significant contributor of this finding occurred more than three years ago, and is not indicative of current licensee performance.
05000458/FIN-2016009-02	2016Q1	River Bend	Failure to Establish Adequate Procedural Guidance	The team reviewed a self-revealing, non-cited violation of Technical Specification 5.4, Procedures, for three examples of the licensees failure to establish sufficient procedural guidance. Specifically, the licensees operations and radiation protection procedures did not provide sufficient direction to plant personnel to expeditiously establish a reactor vessel vent path, restore from a loss of shutdown cooling, and perform time sensitive entries into radiologically controlled areas. This issue was entered into the licensees corrective action program as Condition Reports CR-RBS-2016-0210, CR-RBS-2016-0370, and CR-HQN-2016-0132. Corrective actions included revising the applicable procedures. The failure to establish adequate procedural guidance in accordance with Regulatory Guide 1.33 was a performance deficiency. Specifically, Procedures GOP-0002, Power Decrease/Plant Shutdown, Revision 72, and AOP-0051, Loss of Decay Heat Removal, Revision 313, failed to provide adequate direction to operations personnel to expeditiously establish a reactor vessel vent path and recover shutdown cooling following an isolation. Additionally, Procedure EN-RP-101, Access Control for Radiologically Controlled Areas, Revision 11, failed to provide adequate guidance to perform time sensitive entries into radiologically controlled areas. This performance deficiency is more than minor, and therefore a finding, because it is associated with the procedure quality 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. Specifically, the licensee failed to ensure that adequate procedural direction was provided to operations personnel following a loss of shutdown cooling. This resulted in a delay in the restoration of shutdown cooling and plant heatup. The team performed an initial screening of the finding in accordance with NRC Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process. Using Inspection Manual Chapter 0609, Appendix G, Attachment 1, Exhibit 3, Mitigating Systems Screening Questions, the team determined that the finding is of very low safety significance (Green) because it: (1) affected the design or qualification of a mitigating structure, system, or component, and (2) the structure, system, or component maintained its operability and functionality. A cross-cutting aspect is not being assigned to this finding due to the timing of the performance deficiency not being indicative of current licensee performance.
05000498/FIN-2015004-02	2015Q4	South Texas	Failure to Maintain the Emergency Plan Up to Date With the Safety Evaluation Report	The inspectors identified a non-cited violation of 10 CFR 50.54(q)(2) for failure to maintain the emergency plan in accordance with the approved safety evaluation report. Specifically, the licensee failed to meet 10 CFR 50.47(b)(2) requirements for timely augmentation of response capabilities, in accordance with the approved safety evaluation report. Following an update to the safety evaluation report in 1993, the licensee failed to update the emergency response organization staff augmentation time requirements to commence at the time of an emergency declaration vice from the time of an emergency notification. To restore compliance, the licensee updated the emergency plan in accordance with the current safety evaluation report. Failure to maintain the site emergency plan in accordance with the approved safety evaluation report, dated May 20, 1993, was a performance deficiency. Specifically, the licensee failed to update the ERO staff augmentation time requirements to commence at the time of an emergency declaration, as required by the NRC safety evaluation report. This performance deficiency is more than minor because it is associated with the procedure quality attribute of the Emergency Preparedness Cornerstone and adversely affected the cornerstone objective to ensure that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. This finding was evaluated using Inspection Manual Chapter 0609, Appendix B, Emergency Preparedness Significance Determination Process (SDP), dated September 22, 2015, and was determined to be of very low safety significance (Green) per Table 5.2-1, Significance Examples 50.47(b)(2), because the staffing processes do not meet the threshold of routinely not capable of ensuring timely augmentation of the on shift emergency response staff to the extent that more than one required ERO functional area (in accordance with E-plan commitments) would not be filled. No cross-cutting aspect is assigned because the performance deficiency is not indicative of present performance.
05000498/FIN-2015004-01	2015Q4	South Texas	Failure to Track and Incorporate Actual Plant Data into Simulator Operability Testing	The inspectors identified a finding, associated with simulator operability testing, for the failure of the licensee to track and incorporate actual plant data into their cyclic operability tests, as required by American National Standards Institute-3.5-2009, Nuclear Power Plant Simulators for Use in Operator Training and Examination. With the exception of one transient, the licensee exclusively used engineering analysis from the RETRAN code as baseline data without reference to plant events that may have been related to the required transient tests. This issue was entered into the licensees corrective action program as Condition Report 15-21463. The failure to track and incorporate plant events into baseline data for simulator operability testing is a performance deficiency. It is more than minor and, therefore, a finding because it is associated with the human performance attribute of the Mitigating Systems Cornerstone and negatively affected the objective to ensure the reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, if simulator performance is not being compared to the most relevant baseline data from the plant, the reliability of the simulator performance is reduced. Using Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 worksheets, and the corresponding Appendix I, Licensed Operator Requalification SDP (block 14), the finding was determined to have very low safety significance (Green) because it is a Simulator testing, maintenance, or modification deficiency. This finding has a cross-cutting aspect in the procedure adherence component of the human performance cross-cutting area because the licensee failed to ensure that individuals follow processes, procedures, and work instructions in that the American National Standards Institute-3.5-2009 guidance for selecting baseline data for simulator testing was not followed.
05000446/FIN-2015003-03	2015Q3	Comanche Peak	Notice of Enforcement Discretion 15-4-02 for One Inoperable Train of Emergency Core Cooling Systems	Unit 2 Train B Safety Injection System Inoperable for Longer Than Allowed by Technical Specifications and Notice of Enforcement Discretion 15-4-02 Introduction. The inspectors opened an unresolved item associated with a potential noncompliance with Technical Specification 3.5.2 that occurred on July 10, 2015. Notice of Enforcement Discretion 15-4-02 was granted by the NRC staff agreeing not to enforce compliance with the technical specification completion time for an additional 25 hours. On July 7, 2015, a potential through wall leak from pipe segment SI-2-070 in the Unit 2, train B Safety Injection (SI) pump room was discovered during routine system walkdowns by a licensee engineer. Approximately 1-2 cups of boric acid accumulation was identified on the floor underneath valve 2SI-0055 (SIP 2-02 Suction Test Connection). The pipe insulation was removed to identify the source of the leakage, which was determined to be from a socket weld connection between the six inch suction piping for SI Pump 2-02 and the 34 inch vent piping to 2SI-0055. At 1:04 p.m. on July 7, 2015, the licensee declared unit 2, train B, Emergency Core Cooling System (ECCS) inoperable and entered Technical Specification 3.5.2, Condition B, for one or more (ECCS) trains inoperable for reasons other than one inoperable centrifugal charging pump, and at least 100 percent of the ECCS flow equivalent to a single operable ECCS train available. Required Action B.1 of Technical Specification 3.5.2 required restoration of the train(s) to an operable status within 72 hours. Further, Technical Specification 3.5.2 required that if Required Action B.1 could not be met within 72 hours, unit 2 would be required to enter Technical Specification 3.5.2 Condition C, Required Actions C.1 and C.2, and be in Mode 3 in 6 hours and Mode 4 in 12 hours. The licensees initial assessment determined the likely cause of the socket weld failure to be vibration induced fatigue failure. An attempted repair utilizing ASME Code Case N-666 was conducted on July 8, 2015. During the welding activity a small pinhole leak A-16 developed in the vent piping. The licensee then initiated alternate repair activities including a freeze seal on the affected piping, installation of a new vent line and valve (to facilitate post-repair filling and venting of the SI piping), and repair of the affected weld. The licensee requested a notice of enforcement discretion and an additional 25 hours to restore safety injection pump 2-02, such that the completion time of Required Action B.1 would expire at 2:04 p.m. on July 11, 2015. A notice of enforcement discretion was granted by the NRC staff at 9:20 a.m. on July 10, 2015. Consistent with NRC policy, the NRC agreed not to enforce compliance with the specific technical specifications in this instance, but will further review the cause(s) that created the apparent need for enforcement discretion to determine if there is a performance deficiency, if the issue is more than minor, or if there is a violation of requirements. This issue will be tracked as an unresolved item. (URI 05000446/2015003-03, Notice of Enforcement Discretion 15-4-02 for One Inoperable Train of Emergency Core Cooling Systems)
05000445/FIN-2015008-05	2015Q3	Comanche Peak	Failure to Perform Adequate Operability Assessments associated with Failures of Service Water System Vacuum Breaker during Surveillance Tests	The team identified an unresolved issue associated with the failures of the vacuum service water breakers that remained in service. During these failures, the licensee had documented the surveillance failures as degraded conditions and concluded that they did not have an impact on the operability of the service water system. The team reviewed the licensees operability assessments associated with surveillance tests where at least one of the service water system vacuum breakers failed to meet acceptance standards. During these failures, maintenance personnel mechanically agitated the vacuum breakers in order to get them to operate but did not replace the vacuum breakers until a future date. The inspectors noted that design basis calculations indicate that the larger of the two vacuum breakers (check valve) was required in order to protect the EDG jacket service water coolers and concluded that the licensee did not have appropriate justification to conclude that the service water system remained operable with a failed vacuum breaker if it was the larger breaker. During the inspection period the team was not able to determine which vacuum breakers were found in a degraded condition, therefore more information is required to determine if a non-compliance exists. Specifically, since September 2010, the licensee issued twenty six operability evaluations associated with failed surveillance test on vacuum breakers in the service water system where operators used incorrect information when assessing operability, which failed to establish a reasonable expectation of operability. This issue does not represent an immediate safety concern because at the time of discovery, there were no failed vacuum breakers in service. The licensee entered the finding into corrective action program as Condition Report CR-2015-008334. This issue will remain unresolved until the NRC is provided sufficient information regarding the particulars associated with the check valve/vacuum breaker failures in order to determine if a non-compliance exists. Specifically, the team requires information associated with the specific valve(s) that failed the length of time that the failed valve remained in service prior to replacement; whether the opposite train diesel generator was ever inoperable during the period the failed valve remained in service. (URI 05000445/2015008-05; 05000446/2015008-05, Failure to Perform Adequate Operability Assessments associated with Failures of Service Water System Vacuum Breaker during Surveillance Tests)
05000445/FIN-2015002-03	2015Q2	Comanche Peak	Failure to Update the UFSAR for Restrictions Associated with Shared System Operations of Component Cooling Water	The inspectors identified a non-cited violation of 10 CFR 50.71(e), Maintenance of Records, Making Reports, associated with the licensees failure to update the Final Safety Analysis Report. Specifically, the licensee failed to update the Final Safety Analysis Report to include information detailing restrictions associated with shared system operations of the non-safeguards component cooling water loads between units. This issue does not represent an immediate safety concern because, at the time of identification, the component cooling water systems were not cross connected. The licensee entered this issue into the corrective action program for resolution as Condition Report CR-2014-007235. The licensees failure to update the Final Safety Analysis Report to reflect restrictions associated with shared system operations of the non-safeguards component cooling water loads was a performance deficiency. Because this performance deficiency had the potential to impact the NRCs ability to perform its regulatory function, inspectors evaluated the performance deficiency using traditional enforcement. Using Inspection Manual Chapter 0612, Power Reactor Inspection Reports, dated January 24, 2013, Appendix B, Issue Screening, and Appendix E, Examples of Minor Issues, the Reactor Oversight Program aspect of this performance deficiency was determined to be minor. Using the NRC Enforcement Policy, dated January 28, 2013, the performance deficiency was determined to be a Severity Level IV violation in accordance with Section 6.1.d.3, because the lack of upto-date information in the Final Safety Analysis Report had not resulted in any unacceptable changes to the facility or procedures. Inspectors determined that cross-cutting was not applicable to this finding because it was strictly a traditional enforcement issue.
05000445/FIN-2015002-02	2015Q2	Comanche Peak	Failure to Critique Weaknesses in Radiation Protection Practices	The NRC identified two examples of licensee failures to correct deficiencies occurring during the June 10, 2015, emergency preparedness exercise as required by 10 CFR 50.47(b)(14). Specifically, the licensee failed to identify that a lack of radiological briefings for plant repair teams and a lack of habitability assessments in the Operations Support Center were deficiencies requiring corrective action. This issue was entered into the licensees corrective action program as Condition Report CR 2015-005496. The failure to correct deficiencies occurring during an emergency preparedness exercise is a performance deficiency within the licensees ability to foresee and correct. The performance deficiency is more than minor because the issue is associated with the emergency response organization readiness and performance cornerstone attributes (training) and adversely affected the cornerstone objective. The performance deficiency affects the cornerstone objective because the licensee cannot assure that adequate measures will be taken to protect the health and safety of the public when deficiencies are not corrected. The finding was evaluated using Manual Chapter 0609, Appendix B, Emergency Preparedness Significance Determination Process, dated September 23, 2014, and determined to be of very low safety significance (Green) because the performance deficiency was a failure to comply with NRC requirements and was not a degraded or lost planning standard function. The planning standard was not degraded or lost because the deficiency was not associated with a risk-significant planning standard function and the licensee identified other deficiencies that occurred during the June 10, 2014, exercise. The finding has been assigned a cross-cutting aspect of Identification in the Problem Identification and Resolution cross-cutting area because the licensee failed to identify issues completely and accurately (P.1)
05000528/FIN-2015301-01	2015Q2	Palo Verde	Licensee-Identified Violation	The following violation of very low safety significance (Green) and Severity Level IV was identified by the licensee and is a violation of NRC requirements, which meets the criteria of the NRC Enforcement Policy, for being dispositioned as a Non-Cited Violation. Title 10 CFR 55.49, Integrity of Examinations and Tests, requires, in part, that facility licensees shall not engage in any activity that compromises the integrity of any application, test, or examination required by this part. Contrary to the above, on April 14, 2015, the licensee engaged in an activity that compromised the integrity of the examination. Specifically after administrative JPMs had been administered to the applicants by the examination team, the licensee, upon performing their examination security walk-down, neglected to secure a three-ring binder that contained two reactor operator and two senior reactor operator administrative JPMs that were to be performed the next day. All four JPMs were left unattended and unsecured until 5:00 a.m. on April 15, 2015, when they were discovered as part of the licensee examination security preparation procedure. The four compromised JPMs were replaced by new administrative JPMs as required by NUREG-1021. The failure to meet 10 CFR 55.49 was evaluated through the traditional enforcement process because it impacted the ability of the NRC to perform its regulatory oversight function. This resulted in assignment of a Severity Level IV violation because it involved a non-willful compromise of examination integrity and is consistent with Section 6.4.d of the NRC Enforcement Policy. The associated performance deficiency was screened as Green because there was not an actual effect on the equitable and consistent administration of any examination required by 10 CFR 55.59, Integrity of Examinations and Tests. The licensee entered this issue into their corrective action program as PVAR 4645293.
05000445/FIN-2015002-01	2015Q2	Comanche Peak	Failure to Adequately Assess Risk and Implement Risk Management Actions for Proposed Maintenance	The inspectors identified a non-cited violation of 10 CFR 50.65(a)(4), Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, for the licensees failure to adequately assess risk and implement required risk management actions for a planned maintenance activity. Specifically, the licensee failed to evaluate the risk associated with the use of a non-seismically qualified crane when moving loads over an operable train of service water during installation of a temporary modification in 2014. This issue did not represent an immediate safety concern because, at the time of identification, the maintenance activity was no longer in progress. The licensee entered this issue into the corrective action program for resolution as Condition Report CR-2015-001203. The failure to adequately assess the risk and implement required risk management actions for proposed maintenance activities was a performance deficiency. This performance deficiency was more than minor, and therefore a finding, because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affected the associated objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, dated May 19, 2005, Flowchart 2, Assessment of Risk Management Actions, the inspectors determined the need to calculate the risk deficit to determine the significance of this issue. Based on a review of the licensees risk model it was determined that the incremental core damage probability associated with this finding was less than 1 x 10-6; therefore, this finding is determined to have very low safety significance (Green). The finding has a human performance cross-cutting aspect associated with consistent processes because the licensee failed to use a consistent, systematic approach to evaluate risk for planned maintenance activities (H.13)
05000445/FIN-2015002-04	2015Q2	Comanche Peak	Failure to Evaluate and Appropriately Approve Design Changes	The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, associated with the licensees failure to ensure that design changes were subject to design control measures commensurate with those applied to the original design and were approved by the designated responsible organization. Specifically, the licensee changed required embedment depths for safety-related concrete expansion anchors associated with manhole covers but failed to re-perform the design calculation to demonstrate that the new embedment depth was sufficient for tornado loading. The licensee performed an operability determination which established a reasonable expectation for operability pending final resolution of the issue. This issue was entered into the licensees corrective action program as Condition Report CR-2015-003152. The licensees failure to ensure that changes to the facility were subject to design control measures commensurate with those applied to the original design, and were approved by the designated responsible organization was a performance deficiency. This performance deficiency was more than minor, and therefore a finding, because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affected the associated objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee changed required embedment depths for safety-related concrete expansion anchors associated with manhole covers but failed to re-perform the design calculation to demonstrate that the new embedment depth was sufficient for tornado loading. Using Inspection Manual Chapter (IMC) 0609, Appendix A, The Significance Determination Process (SDP) for Findings AtPower, dated June 19, 2012, inspectors determined that this finding was of very low safety significance (Green) because the finding: (1) was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality, (2) did not represent a loss of system and/or function, (3) did not represent an actual loss of function of at least a single train for longer than its allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time, and (4) does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety significant for greater than 24 hours in accordance with the licensees maintenance rule program. The inspectors determined that this finding does not have a cross-cutting aspect because the most significant contributor of this finding occurred more than three years ago and does not reflect current licensee performance.
05000445/FIN-2015002-05	2015Q2	Comanche Peak	Licensee-Identified Violation	A licensee identified violation of Unit 1 and Unit 2 License Condition 2.G Fire Protection, was identified for the licensees failure to correctly implement the fire protection program with regard to unfused direct current ammeter circuits that could result in a secondary fire due to multiple fire induced faults. Inspectors determined that this issue required a detailed risk evaluation. A senior reactor analyst performed a detailed risk evaluation and the bounding change to the core damage frequency was less than 2E-8/year (Green). The dominant core damage sequences involved a control room fire initiating event in the train A or B direct current ammeter cabinets, a secondary cable fire in a cable tray associated with one train of safety related equipment (two hot shorts required), and having the alternate train of safety related equipment out of service for maintenance. The low fire frequency and the train separation and protection that are required by the fire protection program helped to minimize the significance.
05000445/FIN-2015002-06	2015Q2	Comanche Peak	Licensee-Identified Violation	Title 10 CFR 50.65(a)(3) requires, in part, that performance and condition monitoring activities and associated goals shall be evaluated every refueling cycle and that adjustments should be made to ensure that the objective of minimizing failures is appropriately balanced against the objective of minimizing unavailability. Contrary to the above, on July 11, 2013, the licensee evaluated performance and condition monitoring activities and failed to make necessary adjustments to performance monitoring criteria. Specifically, the licensee had implemented a revision to the plant probabilistic risk analysis in 2012. This revision resulted in changes to reliability and availability assumptions that were identified as needing to be incorporated into the maintenance rule performance criteria for several risk significant systems. The licensee did not implement these changes in a timely manner, and failed to recognize that during the periodic assessment completed on July 11, 2013. In 2014, the licensee hired a new maintenance rule coordinator, who recognized the failure to incorporate the risk analysis recommendations and took action to review the affected performance criteria. The violation is more than minor because it affected the equipment performance attribute of the Mitigating Systems Cornerstone and impacted the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Attachment 04, Initial Characterization of Findings, and Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) For Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, the finding was determined to be of very low safety significance (Green) because the finding was not a deficiency affecting the design or qualification of a mitigating SSC, and the SSCs maintained their operability. The violation was entered into the licensees corrective action program as CR 2015-005304.
05000445/FIN-2015002-07	2015Q2	Comanche Peak	Licensee-Identified Violation	Title 10 CFR 50.54(q)(2) requires, in part, that licensees shall follow and maintain the effectiveness of an emergency plan that meets the planning standards of 10 CFR 50.47(b). Title 10 CFR 50.47(b)(4) requires, in part, that a standard emergency classification and action level scheme is in use by the licensee. The licensees emergency plan provides for the ability to classify an alert due to a seismic event based on an alarm condition on their seismic monitoring system panel in conjunction with other indications. Contrary to the above, during four separate periods between May 16, 2012 and October 1, 2014, the licensee failed to maintain the ability to classify an alert due to a seismic event. The licensees emergency action level HA1.1, an alert due to a seismic event, required the receipt of an alarm from the seismic monitoring system. The licensee had implemented proceduralized compensatory measures when the system was unavailable that consisted of an engineering evaluation to determine whether the event met the emergency action level criteria. The licensee determined these measures would not be sufficient to allow the emergency director to classify the event within fifteen minutes. The licensee discovered this during a review of industry operating experience and submitted a notification report for a loss of major assessment capability. The violation is more than minor because it affected the ERO Performance attribute of the Emergency Preparedness cornerstone and impacted the cornerstone objective to ensure that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Using Inspection Manual Chapter 0609, Attachment 04, Initial Characterization of Findings, and Inspection Manual Chapter 0609, Appendix B, Emergency Preparedness Significance Determination Process, the inspector determined that the violation is of very low safety significance (Green) because the finding represented a failure to comply with planning standard (b)(4), and, using table 5.4-1, was screened as a Green finding because an emergency action level initiating condition was rendered ineffective such that an Alert would be declared in a degraded manner for a seismic event, but no Site Area Emergency or General Emergency initiating conditions were affected. The violation was entered into the licensees corrective action program as CR-2015-003129.
05000323/FIN-2015001-01	2015Q1	Diablo Canyon	Failure to Provide Adequate Design Review of EDG 2-3	The inspectors documented a self-revealing violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to ensure that design control measures shall provide for verifying or checking the adequacy of design by the performance of design reviews and design control measures shall be applied to items such as maintenance, and repair; and delineation of acceptance criteria for inspections and tests. Specifically, the licensee failed to incorporate a design modification to remove a terminal block cover during the original installation of emergency diesel generator 2-3. This modification was identified as a corrective action following previous trips of emergency diesel generators 1-2 and 1-1. The licensees failure to identify that a terminal block cover was removed from emergency diesel generator 2-3 as corrective actions following previous trips of emergency diesel generators 1-2 and 1-1, and to incorporate this modification into the design and installation of emergency diesel generator 2-3, was a performance deficiency. This performance deficiency was more than minor because it is associated with the design control attribute of the Mitigating Systems cornerstone objective and adversely affected the 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 performance deficiency adversely affected the diesel generators capability to operate loaded for the technical specification required time. Using Inspection Manual Chapter 0609, Attachment 04, Initial Characterization of Findings, and Appendix A, The Significance Determination Process (SDP) For Findings At-Power, dated July 1, 2012, the inspectors determined that the finding could not be screened as Green, or very low safety significance, due to loss of a function of a single train for greater than its technical specification outage time. As a result, a detailed risk evaluation was performed by a senior risk analyst. The detailed risk evaluation determined that the finding was Green, or very low safety significance. This finding did not have a cross-cutting aspect because the most significant contributor did not reflect current licensee performance.
05000275/FIN-2014004-05	2014Q3	Diablo Canyon	Notice of Enforcement Discretion 14-4-001 for a Loss of Both Required Offsite Power Circuits	On August 10, 2014, at 6:56 a.m., emergency diesel generator 2-2 was removed from service for a planned maintenance outage. During the maintenance, a diesel fuel oil inlet to fuel header capscrew was discovered broken. An extent of condition review was performed and a similar capscrew was discovered to have an ultrasonic test indication on diesel generator 2-3. Diesel generator 2-3 was declared inoperable August 14, 2014, at 4:31 p.m., and DCPP Unit 2 entered Technical Specification 3.8.1, Condition E, Required Action E.1, to ensure at least two diesel generators were operable. The capscrew on diesel generator 2-3 was replaced, but during preparations to return the diesel generator to service, a separate, non-related failure of the engine driven fuel oil booster pump shaft seal occurred. As required by Technical Specification 3.8.1, Condition H, Action H.1, operators shut the unit down and placed the unit in Mode 3, Hot Standby. Technical Specification 3.8.1, Condition H, Required Action H.2 also required the unit to be in Mode 5 in 36 hours. Enforcement discretion was requested by the licensee to permit additional time to make repairs and restore diesel generator 2-3 to operable status before entry into Mode 5 within 36 hours, as required. An additional 3 hours was requested to restore diesel generator 2-3 such that the completion time of Required Action H.2 would expire at 9:31 a.m. on August 16, 2014. A notice of enforcement discretion (NOED) was granted by the NRC staff at 2:45 p.m. on August 14, 2014. The condition causing the need for this NOED was corrected by the licensee with the restoration of diesel generator 2-3 to operable status, allowing the licensee Unit 2 to exit Technical Specification 3.8.1, Required Action H.2, and the NOED on August 14, 2014, at 6:00 p.m. On August 15, 2014, emergency diesel generator 2-2 was restored to operable status at 2:21 p.m. on August 17, 2014. Consistent with NRC policy, the NRC agreed not to enforce compliance with the specific technical specifications in this instance, but will further review the cause(s) that created the apparent need for enforcement discretion to determine if there is a performance deficiency, if the issue is more than minor, or if there is a violation of requirements. This issue will be tracked as an unresolved item (URI) 05000275/2014004-05; Notice of Enforcement Discretion 14-4-001 for a Loss of Both Required Offsite Power Circuits.
05000275/FIN-2014004-04	2014Q3	Diablo Canyon	Inadequate Procedure Results in Unnecessary Main Steam Safety Valve Lift	The inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensee failure to prescribe a procedure appropriate to the circumstances with respect to safetyrelated atmospheric dump valves and main steam safety valves. Specifically, control of atmospheric steam dump valves was not appropriate for a rapid plant shutdown resulting in unnecessary lifting of a spring-loaded main steam safety valve. The inspectors determined that the licensees failure to ensure appropriate procedures to properly control steam generator pressure and prevent unnecessary lifting of main steam safety valves was a performance deficiency. This performance deficiency was determined to be more than minor because it affected the Mitigating Systems cornerstone attribute of procedural quality and the objective of ensuring the availability and reliability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Attachment 04, Initial Characterization of Findings, and Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the finding was determined to be of very low safety significance (Green) because the finding was a deficiency affecting the design or qualification of a mitigating structure, system, or component that did not affect operability or functionality. The inspectors concluded that this finding affected the cross-cutting aspect of human performance associated with avoiding complacency, because the licensee failed to recognize during rapid load reductions the inherent risk of lifting a main steam safety valve and did not recognize or plan with adequate procedures, for a condition with a potential latent problem.
05000275/FIN-2014004-02	2014Q3	Diablo Canyon	Inadequate Maintenance Procedure Resulted in Improper Configuration of Safety Related Equipment	The inspectors reviewed a Green self-revealing, non-cited violation of Technical Specification 5.4.1.a, Procedures, for failure to implement properly preplanned maintenance procedures affecting the performance of safety-related equipment. Specifically, inspectors reviewed the licensee performance associated with surveillance and maintenance activities and identified two examples of improper configuration of safety-related equipment returned to service, because of inadequate preplanned maintenance procedures. The failure to implement properly preplanned maintenance procedures affecting the performance of safety-related equipment is a performance deficiency. The inspectors determined that the finding was more than minor because it is associated with the procedure quality attribute of the Mitigating System Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesired consequences. Specifically, the restriction of airflow caused by inadvertent closure of ventilation registers following the damper inspection resulted in the undesired consequences of higher ambient 480 volt switchgear room temperatures. In addition, the misconfiguration of the source range N-32 nuclear instrumentation impacted the functioning of the P-6 permissive and prevented it from performing properly during Unit 2 reactor startup such that operator action was necessary to prevent damage to the detector. Using Inspection Manual Chapter 0609, Attachment 04, Initial Characterization of Findings, and Appendix A, Exhibit 2, Mitigating Systems Screening Questions, this finding was determined to be of very low safety significance (Green) because the finding was a deficiency affecting the design or qualification of mitigating structures, systems, and components that did not affect operability or functionality. The inspectors concluded that this finding affected the cross-cutting aspect of human performance associated with documentation, because the licensee did not ensure plant activities are governed with comprehensive maintenance procedures which are complete, accurate, and up to date to ensure work processes did not affect the performance of safety-related equipment.
05000275/FIN-2014004-03	2014Q3	Diablo Canyon	Failure to Provide Adequate Procedural Guidance Resulting in a Loss of Unit 1 230 kV Off-site Power	The inspectors reviewed a Green self-revealing finding for the licensees failure to provide appropriate acceptance criteria to ensure work activities were satisfactorily accomplished. Specifically, the licensee failed to provide acceptance criteria for torqueing or verification of acceptable torqueing during the re-assembly of the load tap changer in Work Order 64006965, Reinhausen Tap Changer Overhaul, for the re-termination of the Unit 1 startup transformer load tap changer diverter switch flex lead terminations. The licensee documented this issue in Notification 50578636. The licensee replaced the load tap changer and revised the procedure as part of their corrective actions. The licensees failure to provide appropriate acceptance criteria in Work Order 64006965 for the re-termination of the Unit 1 Startup Transformer load tap changer diverter switch flex lead terminations was a performance deficiency. Specifically, the work order did not provide acceptance criteria for torqueing or verification of acceptable torqueing during the re-assembly of the load tap changer diverter switch flex lead terminations. This performance deficiency was more than minor because it is associated with the procedure quality attribute of the Initiating Events cornerstone objective and adversely affected the objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Using Inspection Manual Chapter 0609, Attachment 04, Initial Characterization of Findings, and Appendix A, Exhibit 1, Initiating Events Screening Questions, this finding was determined to be of very low safety significance (Green) because, it did not result in a reactor trip or a loss of mitigating equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. This finding has a human performance cross-cutting aspect associated with work management, specifically in that the licensee did not implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority.
05000275/FIN-2014004-01	2014Q3	Diablo Canyon	Failure to Document Degraded Conditions in the Corrective Action Process	The inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and drawings, in that the licensee did not enter degraded conditions into the corrective action process. The inspectors identified two examples. Specifically, on May 12-13, 2014, the licensee experienced high temperatures in the 480 volt vital bus rooms and did not initiate a notification to document the unexpected condition. Second, on May 20, 2014, the licensee failed to document that a 480 volt vital bus room ventilation system register louvers was found closed. The failure to enter problems into the corrective action process on the 480 volt busses was a performance deficiency. The performance deficiency was more than minor because it was associated with the human performance 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, and is therefore a finding. Using Inspection Manual Chapter 0609, Attachment 04, Initial Characterization of Findings, and Appendix A, Exhibit 2, Mitigating Systems Screening Questions, this finding was determined to be of very low safety significance (Green) because, it was not a design or qualification deficiency, was not a loss of the system or function, and did not represent an actual loss of function of a single train for greater than its technical specification allowed outage time. The inspectors determined this finding has a human performance cross-cutting aspect associated with challenging the unknown attribute, specifically in that licensee personnel did not maintain a questioning attitude to resolve unexpected conditions.
05000313/FIN-2013012-04	2014Q1	Arkansas Nuclear	Failure to Follow the Materials Handling Program during the Unit 1 Generator Stator Move	The inspectors reviewed a self-revealing apparent violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, which states, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures or drawings. The licensee did not follow the requirements specified in Procedure EN-MA-119, Material Handling Program, in that, the licensee did not perform an adequate review of the subcontractors lifting rig design calculation and the licensee failed to conduct a load test of the lifting rig prior to use. The licensee initiated Condition Report CR-ANO-C-2013-00888 to capture this issue in the corrective action program. The licensees corrective actions included repairing damage to the Unit 1 turbine deck, fire main system, and electrical system. In addition, changes were made to various procedures including Procedure EN-DC-114, Project Management, to provide guidance on review of calculations, quality requirements, and standards associated with third party reviews. The inspectors determined that the finding was more than minor because it was associated with the procedural control attribute of the initiating event cornerstone, and adversely affected the cornerstones objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown, as well as power operations. The stator drop affected offsite power to Unit 1, resulting in a loss of offsite power for approximately 6 days and a loss of the alternate AC diesel generator. The inspectors used Inspection Manual Chapter 0609, Attachment 0609.04, Initial Characterization of Findings, dated June 19, 2012, to evaluate the significance of the finding. Since the plant was shutdown, the inspectors were directed to Inspection Manual Chapter 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Operational Checklists for Both PWRs and BWRs, Checklist 4, dated May 25, 2004. Using Appendix G, Attachment 1, Checklist 4, the inspectors concluded that this finding represented a degradation of the licensees ability to add reactor coolant system inventory when needed since a loss of offsite power occurred and therefore, this finding required a Phase 3 analysis. A shutdown risk model was developed by modifying the at-power Arkansas Nuclear One Unit 1 Standardized Plant Analysis Risk Model, Revision 8.19. The NRC risk analyst assessed the significance of shutdown events by calculating an instantaneous conditional core damage probability. The results were dominated by two sequences. The largest risk contributor (approximately 97 percent) was based on a failure of the emergency diesel generators without recovery. The second largest risk contributor was the failure to recover decay heat removal. The result of the analysis was an instantaneous conditional core damage probability of 3.8E-4; therefore, this finding was preliminarily determined to have high safety significance (Red). This finding had a cross-cutting aspect in the area of human performance associated with field presence, because the licensee did not ensure adequate supervisory and management oversight of work activities, including contractors and supplemental personnel. Specifically, the licensee did not provide a sufficient level of oversight in that, the requirements in Procedure EN-MA-119, for design approval and load testing of the temporary hoisting assembly, were not followed (H.2) (Section 4OA3.9).
05000368/FIN-2013012-05	2014Q1	Arkansas Nuclear	Failure to Follow the Materials Handling Program during the Unit 1 Generator Stator Move	The inspectors reviewed a self-revealing apparent violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, which states, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures or drawings. The licensee did not follow the requirements specified in Procedure EN-MA-119, Material Handling Program, in that, the licensee did not perform an adequate review of the subcontractors lifting rig design calculation and the licensee failed to conduct a load test of the lifting rig prior to use. The licensee initiated Condition Report CR-ANO-C-2013-00888 to capture this issue in the corrective action program. The licensees corrective actions included repairing damage to the Unit 1 turbine deck, fire main system, and electrical system. In addition, changes were made to various procedures including Procedure EN-DC-114, Project Management, to provide guidance on review of calculations, quality requirements, and standards associated with third party reviews. The inspectors determined that this finding was more than minor because it was associated with the procedural control attribute of the initiating event cornerstone, and adversely affected the cornerstones objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown, as well as power operations. The stator drop caused a reactor trip on Unit 2 and damage to the fire main system which resulted in water intrusion into the electrical equipment causing a loss of startup transformer 3. This resulted in the loss of power to various loads, including reactor coolant pumps, instrument air compressors, and the safety-related Train B vital electrical bus. The inspectors used Inspection Manual Chapter 0609, Attachment 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, to evaluate the significance of the finding. Since this was an initiating event, the inspectors used Exhibit 1 of Appendix A and determined that Section C, Support System Initiators, was impacted because the finding involved the loss of an electrical bus and a loss of instrument air. The inspectors determined that Section E, External Event Initiators, of Exhibit 1 should also be applied because the finding impacted the frequency of internal flooding. Since Sections C and E were impacted, a detailed risk evaluation was required. The NRC risk analyst used the Arkansas Nuclear One, Unit 2 Standardized Plant Analysis Risk Model, Revision 8.21, and hand calculation methods to quantify the risk. The model was modified to include additional breakers and switching options, and to provide credit for recovery of emergency diesel generators during transient sequences. Additionally, the analyst performed additional runs of the risk model to account for consequential loss of offsite power risks that were not modeled directly under the special initiator. The largest risk contributor (approximately 96 percent) was a loss of all feedwater to the steam generators, with a failure of once-through cooling. The result of the analysis was a conditional core damage probability of 2.8E-5; therefore, this finding was preliminarily determined to have substantial safety significance (Yellow). This finding had a cross-cutting aspect in the area of human performance associated with field presence, because the licensee did not ensure adequate supervisory and management oversight of work activities, including contractors and supplemental personnel. Specifically, the licensee did not provide a sufficient level of oversight in that, the requirements in Procedure EN-MA-119, for design approval and load testing of the temporary hoisting assembly, were not followed (H.2) (Section 4OA3.9).
05000458/FIN-2014301-01	2014Q1	River Bend	Failure of the Plant Referenced Simulator to Demonstrate Expected Plant Response with Four Examples	Title 10 CFR Part 55.46(c)(1), Plant-Referenced Simulators, states, in part, A plant referenced simulator used for the administration of the operating test...must demonstrate expected plant response to operator input and to normal, transient, and emergency conditions to which the simulator has been designed. Contrary to this, Operators were unable to open the main steam isolation valves because the River Bend Station simulator did not correctly model the differential pressure across the main steam isolation valves. Because of this, the job performance measure had to be rejected and another developed. This modeling deficiency was entered into the licensees corrective action program as Condition Report CR-RBS-2014-965. On multiple occasions, the River Bend Station simulator randomly initiated a main turbine runback when plant conditions did not warrant this action. After unsuccessful attempts were made to resolve this modeling deficiency, the applicants were briefed to ignore this event should it occur. This modeling deficiency was entered into the licensees corrective action program as Condition Reports CR-RBS-2014-965 and CR-RBS-2014-1496. The River Bend Station simulator initiated a control rod drift during a scenario where plant conditions did not support this response. After identification, the licensee entered the issue into the licensees corrective action program as Condition Report CR-RBS-2014-1496. These failures of the plant-referenced simulator to demonstrate expected plant response during conditions to which the simulator has been designed to respond was a performance deficiency. The finding was more than minor because it is associated with the human performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring availability, reliability, and capability of systems needed to respond to initiating events to prevent undesired consequences. Specifically, the incorrect simulator response could adversely affect the operating crews ability to assess plant conditions and take actions in accordance with approved procedures. In accordance with NRC Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheets, and the associated Appendix I, Licensed Operator Requalification Significance Determination Process (SDP), Block 15, the finding was determined to be of very low safety significance because the deficient simulator performance did not negatively impact operator performance in the actual plant during a reportable event. Following the operating test, it was discovered the modeling deficiencies were introduced as part of a simulator upgrade more than ten years ago and therefore, are not considered to be a reflection of current performance. The hardware failure associated with the main steam line pressure gauge was determined to have no actual operator impact and was not a generic training issue. Therefore, this finding has no cross-cutting aspect associated with it.
05000498/FIN-2013301-01	2013Q4	South Texas	Failure to Maintain Licensed Operator Examination Integrity	A self-revealing Green noncited violation of 10 CFR Part 55.49, Integrity of Examinations and Tests, was identified for the failure of operations training personnel to ensure the integrity of an operating test scheduled for administration for an initial licensing examination scheduled for the week of September 30, 2013. This failure resulted in a potential compromise of examination integrity, but did not lead to an actual compromise of the administered examination. This finding was more than minor because it would have affected examination integrity had it not been detected. However, because no actual compromise of examination integrity occurred, the finding was determined to have very low safety significance. This finding had a cross-cutting aspect in the area of human performance associated with work practices because the licensee did not properly self- and peer check to ensure a potential compromise of examination materials would not occur (H.4(a)) (Section 4OA5.5).
05000482/FIN-2013003-05	2013Q2	Wolf Creek	Failure to Properly Manage Reactivity Changes when Swapping Turbine Steam Admission Modes from Full to Partial Arc	Inspectors identified a Green non-cited violation of Technical Specification 5.4.1.a for the failure to follow Conduct of Operations and Reactivity Management procedures. The inspectors reviewed an unplanned 11 percent power increase during a shift in turbine control modes, and identified that pre-job briefings did not adequately discuss expected plant response, operators did not take action to limit the power increase when an unexpected response was observed, and management was not adequately involved in decision making prior to continuing power ascension before the details of an apparent turbine control malfunction were fully understood. This issue was entered into the licensees corrective action program under Condition Report 68711. Failure to provide contingency actions for a greater than anticipated reactor transient in the pre-job reactivity brief, and continuing with power ascension without understanding the cause of the unexpected turbine control system behavior is a performance deficiency. The performance deficiency is more than minor because it affected the human performance attributes of the Initiating Events cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Using Inspection Manual Chapter 0609 Appendix A, Checklist 1, Initiating Events Screening Questions, and the inspectors determined that the finding was of very low safety significance (Green) because the finding did not result in a reactor trip coincident with the loss of mitigation equipment. The inspectors determined that this finding had a cross-cutting aspect in the area of human performance area of work practices because the licensee failed to communicate human error prevention techniques, such as holding pre-job briefings, self and peer checking, and proper documentation of activities such that work activities were performed safely. In addition, personnel proceeded in the face of uncertainty or unexpected circumstances. Specifically, in the first example control room operators pre-job reactivity brief was not appropriate commensurate with the risk of the assigned task; in the second example station personnel proceeded in the face of uncertainty.
05000482/FIN-2013003-01	2013Q2	Wolf Creek	Failure to Follow Station Procedures	The inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, which states, in part, activities affecting quality shall be prescribed by procedures of a type appropriate to the circumstances and accomplished in accordance with these procedures. Contrary to the above, the licensee failed to ensure procedures related to the boric acid corrosion control program were adequate and properly implemented. Specifically, prior to February 19, 2013, the licensee failed to: (1) resolve discrepancies within the boric acid corrosion control program procedure; (2) resolve discrepancies between the boric acid corrosion control program procedure and the boric acid leak management procedure; and (3) failed to track and resolve leakage for locations where health physics had installed drip catch containments, to review the Health Physics Drip Bag Log as part of the quarterly outside containment walkdown, and to add component locations to the program. Further, the licensee failed to periodically assess the effectiveness of the program on a refueling frequency. The violation was entered into the licensees corrective action program as Condition Report 65212. The inspectors determined that the failure to recognize discrepancies between boric acid control procedures and the failure to follow boric acid program procedures was a performance deficiency. The performance deficiency was more than minor because it affected the Initiating Events Cornerstone attribute of procedure quality and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations, and if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern. Specifically, failure to resolve discrepancies within procedures or track and resolve leak locations where health physics had installed drip catch containments had the potential to mischaracterize leaks or allow leaks to corrode safety-related systems. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the finding was determined to be of very low safety significance (Green), because the finding was a procedure quality problem that did not represent a loss of system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The finding had a cross-cutting aspect in the area of human performance associated with the work practices component because the licensee failed to ensure supervisory and management oversight of work activities, including procedure appropriateness and compliance, such that nuclear safety is supported.
05000482/FIN-2013003-02	2013Q2	Wolf Creek	Failure to Identify Leakage at Refuel Pool Cavity	The inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, which states, in part, Measures shall be established to assure that conditions adverse to quality are promptly identified and corrected. Contrary to the above, the licensee failed to identify and correct a condition adverse to quality in a timely manner. Specifically, prior to February 19, 2013, the licensee failed to document the large area of boric acid leakage and corroded steel plates on the south primary shield wall of the containment refueling pool. The violation was entered into the licensees corrective action program as Condition Report 64213. The inspectors determined that the failure to promptly identify and evaluate a condition adverse to quality was a performance deficiency. The performance deficiency was more than minor because it affected the Initiating Events Cornerstone attribute of human performance and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations, and if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern. Specifically, failure to implement corrective actions could result in increased leakage and further degradation of the safety system. Using Manual Chapter 0609.04, Phase 1 Initial Screening and Characterization of Findings, the inspectors determined that this finding was of very low safety significance (Green), because it was not a design or qualification deficiency, did not represent a loss of system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The finding had a cross-cutting aspect in the area of human performance associated with the work practices component because the licensee failed to define and effectively communicate expectations regarding procedural compliance and that personnel follow procedures.
05000482/FIN-2013003-03	2013Q2	Wolf Creek	Diesel Generator Pressure Switch Failed Due to Instrument Line Pressure Oscillations	A self-revealing non-cited violation of 10 CFR Part 50 Appendix B, Criterion XVI, Corrective Action, was identified on March 13, 2013. Specifically, the licensee replaced a jacket water pressure transmitter ten times, but failed to correct pressure oscillations that caused a fatigue failure of a pressure switch diaphragm, which rendered emergency diesel generator B inoperable. The inspectors concluded that the licensee ineffectively focused on correcting the apparent source of the pressure oscillations, but failed to evaluate the effects of the pressure cycles on components exposed to the same oscillations. This issue was entered into the licensees corrective action program as Condition Report 65624. Failure to analyze the effects of pressure oscillations in the emergency diesel jacket water system on interfacing system components is a performance deficiency. The performance deficiency is more than minor because it affected the equipment performance attribute of the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609 Appendix A, Significance Determination Process for Findings At Power, and determined that the finding screens as very low safety significance (Green) because the finding does not meet any criteria outlined in the Exhibit 2, Section A. Specifically the finding did not represent a loss of system safety function and did not exceed its technical specification allowed outage time of 72 hours. The inspectors determined that the finding had a cross-cutting aspect in the area of problem identification and resolution evaluations because the licensee failed to ensure that issues that potentially affect nuclear safety are fully evaluated and addressed in a timely manner.

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