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	Quarter	Site	Title	Description
05000275/FIN-2018008-05	2018Q2	Diablo Canyon	Minor Violation	Performance Deficiency: Failure to use the site corrective action program to track, trend, correct, and prevent recurrence of failures and deficiencies in the physical protection program, as required by Title 10 of the Code of Federal Regulations 73.55, Requirements for physical protection of licensed activities in nuclear power reactors against radiological sabotage. On April 17, 2018, during a plant tour, inspectors identified a deficiency associated with the physical protection program and brought it to the attention of control room operators. On May 1, 2018, inspectors asked licensee personnel for a copy of the SAPN documenting the deficiency. None had been initiated. Further, the deficiency had not been logged in the security logs as required. The failure to log the issue was itself a loggable event. The licensee documented the deficiency and the failure to initially document it in SAPN 50978291. Screening: The performance deficiency was minor because it would not have led to a more significant security concern and did not adversely affect the security cornerstone objective. Enforcement: This failure to comply with 10 CFR 73.55 constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy.
05000482/FIN-2018007-02	2018Q2	Wolf Creek	Minor Violation	Performance Deficiency: Failure to promptly identify and correct known-defective switches in inservice safety-related breakers, or to control nonconforming breakers accepted into warehouse stores, as required by 10 CFR 50 Appendix B Criteria XV and XVI. In February 2008, the licensee received a notification from GE Hitachi of reduced reliability of some safety-related circuit breakers due to defective cutoff switches internal to the breakers. The licensee incorrectly screened this information as not applicable to the Wolf Creek Generating Station. In August 2011, after licensee engineers received the information again from industry peers, the licensee screened the information as applicable. The licensee then added steps to its overhaul and pre-install test procedures to check for the defective subcomponent. These steps were performed during subsequent regularly scheduled overhaul or pre-install tests, with the last affected switches being replaced in June 2014 and the last potentially susceptible safety-related breaker being inspected in March 2015. The team determined that because the station had information on the defect in February 2008, but did not correct the condition until 2014 and did not confirm that it was corrected until 2015, the licensee had failed to promptly identify and correct a condition adverse to quality. Further, the licensee failed to inspect or place administrative controls on potentially affected spare breakers that had been accepted into warehouse stores, though the added steps in the pre-install procedure likely would have prevented a defective component from being installed. However, by failing to segregate the potentially affected components until they were inspected, the licensee failed to comply with quality assurance requirements for control of nonconforming components. On June 26, 2018, the licensee put a hold on four potentially affected breakers that were in warehouse stores. The licensee documented this performance deficiency in CR 124693. Screening: The performance deficiency was minor because the licensee did not experience an inservice failure as a result of the defect during the 6 years they remained in service and had a procedure in place that would likely have prevented a defective spare from being issued for installation. Therefore, there was no adverse effect on the mitigating systems cornerstone objective and there was no potential to create a more significant safety concern. Enforcement: This failure to comply with 10 CFR 50 Appendix B Criteria XV and XVI constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy.
05000275/FIN-2018008-04	2018Q2	Diablo Canyon	Licensee-Identified Violation	This violation of very low safety-significant was identified by the licensee and has been entered into the licensee corrective action program. This is being treated as a non-cited violation (NCV), consistent with Section 2.3.2 of the Enforcement Policy. 7 Violation: Title 10 CFR Part 50, Appendix B, Criterion III, requires that measures shall include provisions to assure that appropriate quality standards are specified and included in design documents, and that deviations from such standards are controlled. Contrary to the above, from approximately February 2004 until August 2017, the licensee did not assure that appropriate quality standards were specified and included in design documents, and that deviations from such standards were controlled. Specifically, the licensee had classified the seat o-ring used in Crosby and Lonergan pressure relief valves (e.g., RV-354 and RV-355) servicing safety-related back-up air/nitrogen applications as non-safety related when they should have been classified as safety-related. Consequently, the o-rings were procured as commercial grade (non-safety related), not dedicated as safety-related and installed in safety-related equipment. Significance/Severity Level: This violation was more than minor because it had the potential to lead to a more significant safety concern if left uncorrected. Specifically, the use of non-qualified seat o-rings had the potential to cause excessive leakage past the seat, adversely affecting the fixed air/nitrogen volume required to operate safety-related equipment during a loss of normal air/nitrogen. Using IMC 0609, Appendix A, dated June 19, 2012, the team determined that this violation was of very low safety significance (Green) because it was a deficiency affecting the design or qualification of a structure, system or component, and operability was maintained. Corrective Action Reference(s): SAPNs 50935776 and 50970247
05000482/FIN-2018007-01	2018Q2	Wolf Creek	Failure to Provide Adequate Work Instructions for Preventive Maintenance on Safety-Related Equipment	The team reviewed a Green, self-revealed non-cited violation of Technical Specification 5.4.1.a to establish, implement, and maintain written procedures recommended by Regulatory Guide 1.33, Appendix A, Revision 2. Specifically, work instructions for the preventive maintenance for the train B Class 1E electrical equipment A/C unit SGK05B, lacked adequate guidance for preventive maintenance and calibration of the associated thermostat. This resulted in the loss of cooling failure of the A/C unit SGK05B,on February 12, 2018.
05000275/FIN-2018008-01	2018Q2	Diablo Canyon	Emergency Diesel Generator Mission Time for Operability Evaluations	The team identified an unresolved item (URI) related to diesel generator (DG) mission time for operability evaluations. On December 3, 2016, an operator discovered during rounds that the air inlet boot seal on DG 1-2 had degraded, and subsequently, an inspection of the other diesel generators (DGs) revealed that the DG 2-2 boot seal was also degraded. The licensee performed an operability evaluation and concluded that the DGs were operable based on a mission time of 24 hours. The licensee then performed a past operability evaluation, concluding that the DGs had remained able to perform their safety function for this stated 24-hour mission time despite the deficiency; therefore no licensee event report was required by 10 CFR 50.73. The team requested information related to the basis of the 24-hour mission time. The licensee provided a non-controlled reference document, Engineered Safety Feature (ESF) Equipment Mission Time, to the licensees operability determination Procedure OM7.ID12. The document listed the mission time for the DGs as 7 days (24 hours, 6 hours). The 6 and 24 hour values depend on the particular accident sequence and electrical power recovery time, and were from a letter sent to the NRC related to the licensees Individual Plant Examination of External Events (IPEEE), which is a plant-specific probabilistic risk assessment (PRA). The 7-day value is related to the required diesel fuel oil storage volume as discussed in Technical Specification Bases 3.8.3. The document also states that the licensee has no defined post-accident operation / mission times because such times are not mandated by regulation or recommended by NRC guidance. The team noted, however, that IPEEEs do not typically evaluate accidents past 24 hours, and furthermore, IMC 0326, Operability Determinations and Functionality, states that the use of PRA or probabilities of occurrence of accidents or external events is not consistent with the assumption that the event occurs, and is not acceptable for making operability decisions. Additionally, Procedure OM7.ID12 defines mission time as the duration of structure, system, or component (SSC) operation that is credited in the current licensing bases for the SSC to perform its specified safety function; however, as documented above by the licensee, there is no design or licensing basis mission time for the DGs. The licensees definition of mission time is essentially the same as described in IMC 0326. The inspectors performed a brief review of documents related to mission times. Technical Specification Limiting Condition for Operation 3.8.3, Diesel Fuel Oil, Lube Oil, Starting Air, and Turbocharger Air Assist, requires verification of diesel fuel oil level to satisfy a 7-day fuel oil storage requirement. Additionally, NUREG-1407 discusses an Electric Power Research Institute approach that defines and evaluates the capacity of those components required to bring the plant to a stable condition (either hot or cold shutdown), and maintain that condition for at least 72 hours. Also, the ESF equipment mission time document referenced several 30-day mission times for SSCs that would require emergency power from either offsite power, if available, or the DGs. The team also performed a search of previous NRC findings at the DCPP, Unit 1 and 2, and found one reference to a 7-day mission time for the DGs in NRC Pilot Engineering Inspection Report 2006005. The inspectors also reviewed NEI 97-04, Design Bases Program Guidelines, Revised Appendix B, Guidance and Examples for Identifying 10 CFR 50.2 Design Bases. The Appendix describes how the 10 CFR 50.2 design bases of a facility are a subset of the current licensing basis and are required pursuant to 10 CFR 50.34(a)(3)(ii) and (b) and 10 CFR 50.71(e), to be included in the updated Final Safety Analysis Report (FSAR). Title 10 CFR 50.2 design bases consist of design bases functions and design bases values. Design bases values are the values or ranges of values of controlling parameters established as reference bounds for design to meet design bases functional requirements. In other words, the 10 CFR 50.2 design bases include the bounding conditions under which SSCs must perform their design bases functions and may be derived from normal operation, or any accident or events for which SSCs are required to function. Because 10 CFR 50.71(e), IMC 0326, and Procedure OM7ID.12 indicated that DG mission time should be part of the design and licensing bases, and documented in the FSAR, but a DG mission time design and licensing basis does not appear to exist at DCPP, Units 1 and 2, the inspectors could not determine that an appropriate mission time was used for a past operability determination. Therefore, the team could not conclude that the licensee had not missed a 10 CFR 50.73 event report because of a potentially incorrect assumption about DG mission time. This is applicable to both units. Planned Closure Action(s): In order to resolve this issue, the NRC needs to determine whether or not the basis for the 24-hour DG mission time is appropriate by determining which standard or standards apply to mission time at DCPP, Units 1 and 2. Licensee Action(s): Because the licensees position is that the DG mission time is not a part of their current licensing or design basis, they maintain that the 24-hour mission time used in the past operability determination was adequate to provide reasonable assurance of operability and, therefore, no event report was required. However, prior to this inspection and because of other uncertainties in determining mission times, the licensee generated Notification 50832335 to reassess the mission times associated with the ESF equipment. The intent is to develop the bases for ESF equipment mission time in a controlled document. However, this effort is not yet complete and, as such, the mission time for the DGs has not been evaluated under this notification. Corrective Action Reference(s): Notifications 50832335, 50882125, 50882140, and 50882498.
05000275/FIN-2018008-03	2018Q2	Diablo Canyon	Failure to Promptly Identify and Correct Emergency Diesel Generator 1-1 Cardox System Inoperability	An NRC-identified, Green, non-cited violation (NCV) of the licensees fire protection license condition occurred when licensee personnel failed to identify a trouble light lit on the Emergency Diesel Generator (DG) 1-1 cardox fire protection system panel. The light, which had been lit for 2 weeks before being identified by the NRC, indicated a condition that would have prevented the automatic fire suppression system from effectively suppressing a fire in the DG 1-1 room.
05000275/FIN-2018008-02	2018Q2	Diablo Canyon	Failure to Identify Diesel Generator Air Inlet Boot Seal Critical Characteristics	A self-revealed, Green, non-cited violation (NCV) of Title 10, Code of Federal Regulations(CFR) Part 50, Appendix B, Criteria VII and XV, occurred when the licensee failed to ensure materials intended for installation in safety-related applications conformed to procurement requirements or, if they did not, were adequately controlled and evaluated.
05000298/FIN-2018011-04	2018Q2	Cooper	Incorrect Classification of Potential Safety-Related Components	An NRC-identified, Green, Non-cited Violation of Title 10, Code of Federal Regulations Part 50, Appendix B, Criterion III, Design Control, occurred for failure to assure that applicable regulatory requirements and the design basis were correctly translated into specifications, drawings, procedures, and instructions. Specifically, the inspectors identified three examples of the licensees failure to properly classify potential safety-related components in the emergency diesel generator ventilation system and RHR service water booster pump room cooling systems.
05000298/FIN-2018011-03	2018Q2	Cooper	Inadequate Design Basis Calculation for the EDG Rooms Temperature Distribution	An NRC-identified, Green, Non-cited Violation of Title 10, Code of Federal Regulations Part 50, Appendix B, Criterion III, Design Control, occurred for the licensees failure to ensure design control measures provide for verifying or checking the adequacy of design of the emergency diesel generator room ventilation system by use of alternate or simplified calculation methods, or by a suitable testing program. Specifically, the licensee incorrectly extrapolated the results of the test program, which led to an incorrect room temperature profile. Additionally, the design calculation did not assume potential failures of the CO2 dampers.
05000298/FIN-2018011-02	2018Q2	Cooper	Failure to Ensure Adequate Design Control Measures are in Place Associated with RHR Service Water Booster Pump Room Cooling	An NRC-identified, Green, Non-cited Violation of Title 10, Code of Federal Regulations Part 50, Appendix B, Criterion III, Design Control, occurred for failure to assure that applicable regulatory requirements and the design basis were correctly translated into specifications, drawings, procedures, and instructions. Specifically, the licensee failed to incorporate malfunctions of the residual heat removal (RHR) service water booster pump (SWBP) room cooling temperature switch, which could cause environmental changes leading to functional degradation of system performance, into the design basis to verify the necessary protection system action be retained.
05000298/FIN-2018011-01	2018Q2	Cooper	Failure to Correct Extent of Condition of Surge Suppression Varistor Failures	An NRC-identified, Green, Non-cited Violation of Title 10, Code of Federal Regulations Part 50, Appendix B, Criterion XVI, Corrective Action, occurred when the licensee failed to correct conditions adverse to quality associated with the corrective actions identified in Condition Report RCR 2002-1665 to verify that installed surge suppressor varistors were appropriately sized and that design information was correctly reflected in controlled drawings for the reactor protection system, diesel generator control circuits, and high pressure coolant injection control circuits.
05000275/FIN-2018008-06	2018Q2	Diablo Canyon	Minor Violation	Performance Deficiency: Failure to install safety-related pressure transmitters (PTs) in accordance with engineering design documents, without documented authorization and prior approval for deviation from that design. Unit 2 Steam Generator pressure transmitters PT-544A and PT-534A were not installed per design. The design called for mounting the PTs on independent unistruts but, contrary to this, the transmitters were installed on a common unistrut. Though the new mounting configurations were documented and analyzed in SAPNs 50881613 and 50881415, Work Order 68039185 which installed the PTs did not record the deviation from originally designed mounting configuration. The licensee attributed the failure to install per original design to human error and initiated SAPN 50976632 to evaluate it. Screening: The performance deficiency is minor in that the current configuration was evaluated not to affect the seismic or structural qualification. Enforcement: This failure to comply with 10 CFR Part 50, Domestic licensing of production and utilization facilities, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, Criterion III, Design Control, constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy.
05000382/FIN-2017010-01	2017Q3	Waterford	Failure to Evaluate Departures from Approved Methodologies for Reactor Vessel Fluence	The inspectors identified a Severity Level IV, non-cited violation of 10 CFR 50.59, Changes, Tests, and Experiments, Section (c)(1), which states, in part, that a licensee may make changes in the facility as described in the updated safety analysis report without obtaining a license amendment pursuant to 10 CFR 50.90 only if: (i) a change to the technical specifications incorporated in the license is not required, and (ii) the change, test, or experiment does not meet any of the criteria in paragraph (c)(2). Title 10 CFR 50.59, Section (c)(2)(viii), states, in part, that a licensee shall obtain a license amendment pursuant to Section 50.90 prior to implementing a proposed change, test, or experiment if the change, test, or experiment would result in a departure from a method of evaluation described in the updated safety analysis report used in establishing the design bases or in the safety analyses. Specifically, since January 2017, the licensee revised updated final safety analysis report Section 4.3.3.3 to reflect RAPTOR-M3G as the current licensing basis fluence method without first obtaining a license amendment. This finding was entered into the licensees corrective action program as Condition Report CR-WF3-2017-04748. The inspectors determined that the failure to evaluate proposed changes to determine if prior NRC review was required in accordance with 10 CFR 50.59 was a performance deficiency. Using NRC Inspection Manual Chapter 0612, Appendix B, Issue Screening, the inspectors determined that this performance deficiency had minor safety significance. In accordance with the NRC Enforcement Manual, violations of 10 CFR 50.59 are not processed through the Reactor Oversight Process significance determination process because this violation potentially impacted the ability of the NRC to perform its regulatory oversight function. Therefore, this violation was processed through traditional enforcement examples of Section 6.1 of the NRC Enforcement Policy. This violation was more than minor because there was a reasonable likelihood that the change would require NRC review and approval prior to implementation, similar to the more than minor example of a change in requirements in the NRC Enforcement Manual, Appendix E, Minor Violations Examples, dated September 9, 2013. Since the violation was associated with a performance deficiency of minor significance, the traditional enforcement violation was determined to be a Severity Level IV violation, consistent with the example in paragraph 6.1.d(2) of the NRC Enforcement Policy.
05000275/FIN-2017002-02	2017Q2	Diablo Canyon	Failure to Conduct Required Biennial Medical Examinations Within Two Years	SL -IV. The inspectors identified a Severity Level IV, non -cited violation of 10 CFR 55.21, Medical Examination, for the licensees failure to ensure that a medical examination by a physician to determine satisfaction of 10 CFR 55.33(a)(1) requirements was conducted every 2 years for two licensed senior operators. Specifically, one licensed senior operator exceeded the two- year medical examination requirement by approximately 16 months between November 27, 2015, and April 6, 2017. A second licensed senior operator exceeded the 2 -year medical examination requirement by 4 months between November 19, 2016, and April 6, 2017. As a corrective action, the licensee has conducted the required medical examination for one senior operator and initiated a license termination request for the other senior operator. This issue was entered into the licensees corrective action program as Notification 50912407. The failure of the facility licensee to conduct required biennial medical examinations for two licensed senior operators was a performance deficiency. This issue was evaluated using the traditional enforcement process because it negatively impacted the NRCs ability to perform its regulatory oversight function. Specifically, the failure to comply with medical testing requirements for two operators compromised the facility licensees ability to assure conformance to medical standards, detect non -conforming medical conditions, and report non-conformances to the NRC. This performance deficiency was determined to be Severity Level IV because it fits the Severity Level IV example of Enforcement Policy Section 6.4.d.1, Violation Examples: Licensed Reactor Operators. This section states, Severity Level IV violations involve, for example ... (b) an individual operator who did not meet the American National Standards Institute/American Nuclear Society (ANSI/ANS) 3.4, Medical Certification and Monitoring of Personnel Requiring Operator Licenses for Nuclear Power Plants, Section 5, Health Requirements and Disqualifying Condit ions, as certified on NRC Form 396, Certification of Medical Examination by Facility Licensee, required by 10 CFR 55.23, Certification, but who did not perform the functions of a licensed operator or senior operator while having a disqualifying medical condition. No cross -cutting aspect was assigned because the violation was processed using traditional enforcement.
05000275/FIN-2017002-03	2017Q2	Diablo Canyon	Failure to Report a Permanent Medical Condition Within 30 Days	SL -IV. The inspectors identified a Severity Level IV, non -cited violation of 10 CFR 55.25, Incapacitation Because of Disability or Illness, for the licensees failure to notify the NRC within 30 days of a change to one licensed senior operators medical condition. Specifically, the licensed senior operator developed a permanent medical condition which caused him to permanently leave the site on December 1, 2014, and transition into a long- term disability program on April 23, 2015. The licensee did not notify the NRC of this change in medical condition. As a corrective action, the licensee initiated a license termination request for the affected operator, effective April 6, 2017. This issue was entered into the licensees corrective action program as Notification 50912407. The failure of the facility licensee to notify the NRC within 30 days of a change in a licensed senior operators medical condition was a performance deficiency. This issue was evaluated using the traditional enforcement process because it negatively impacted the NRCs ability to perform its regulatory oversight function. Specifically, the failure to report 4 changes in a licensed senior operators medical condition prevented the NRC from taking action to issue either a license amendment or termination, as appropriate. This performance deficiency was determined to be Severity Level IV because it fits the Severity Level IV example of Enforcement Policy Section 6.4.d.1, Violation Examples: Licensed Reactor Operators. This section states, Severity Level IV violations involve, for example (b) an individual operator who did not meet the American National Standards Institute/American Nuclear Society (ANSI/ANS) 3.4, Medical Certification and Monitoring of Personnel Requiring Operator Licenses for Nuclear Power Plants, Section 5, Health Requirements and Disqualifying Conditions, as certified on NRC Form 396, Certification of Medical Examination by Facility Licensee, required by 10 CFR 55.23, Certification, but who did not perform the functions of a licensed operator or senior operator while having a disqualifying medical condition. No cross -cutting aspect was assigned because the violation was processed using traditional enforcement
05000275/FIN-2017002-01	2017Q2	Diablo Canyon	Inadequate Expansion Scope of Risk - Informed Welds	Green . The inspectors identified a non -cited violation of the licensees risk -informed inservice inspection program (which is their alternative to portions of the ASME Code, Section XI inservice inspection program approved in accordance with 10 CFR 50.55a(z)) for the failure to properly expand the scope of additional welds to inspect. Specifically, a rejectable flaw on a pipe weld in the pressurizer spray line was identified during refueling outage 1R19 while performing an ultrasonic examination. The licensee expanded the inspection scope by four additional welds, but failed to select those assigned with the same degradation. For immediate corrective actions, the licensee identified and intended to inspect four additional welds assigned to the same degradation mechanism as required by the risk -informed inservice inspection program. This issue was entered into the licensees corrective action program as Notification 50920222. The licensees failure to properly expand the weld examination scope as required by the risk -informed inservice inspection program was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance 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 undesirable consequences. Specifically, the failure to select additional welds that were susceptible to the same degradation mechanism as weld WIB -378 placed the plant at an increased risk due to the potential of having an active degradation mechanism that could affect additional components. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP ) for Findings At-Power, dated June 19, 2012, the inspector s determined the finding screened as having very low significance (Green) because: (1) it was not a design deficiency; (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 technical specification allowed outage time; and (4) did not result in the loss of a high safety -significant non -technical specification train. This finding had a cross -cutting aspect in the area of human performance associated with 3 change management because leaders failed to use a systematic process for evaluating and implementing the change to a risk -informed inservice inspection program. The implementing procedure failed to include the reference to degradation mechanism allowing for a misinterpretation of weld expansion requirements once a flaw was identified in a weld WIB -378 (H.3).
05000275/FIN-2017002-04	2017Q2	Diablo Canyon	Failure to Follow Procedures Results in Partial Loss of Cooling Flow to Shutdown Cooling	Green . The inspectors reviewed a self -revealing, non- cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because PG&E personnel failed to follow the requirements of AD7.ID14, Assessment of Integrated Risk, Revision 11. Specifically, PG&E personnel failed to obtain shift manager permission, conduct a protected equipment briefing, and document shift manager approval prior to performing work on protected equipment. This resulted in a loss of flow of cooling water to one of two in- service shutdown cooling residual heat removal heat exchangers and subsequent perturbation in reactor coolant system temperature during refueling outage 1R20. The inspectors determined that PG&E s failure to follow AD7.ID14, Assessment of Integrated Risk, Section 5.14 Performing Work on Posted Protected Equipment, was a performance deficiency within PG&Es ability to foresee and correct. This performance deficiency was considered to be more than minor because it impacted the configuration control attribute of the Mitigating Systems cornerstone and its objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the loss of cooling flow to the RHR heat exchanger while in shutdown cooling mode resulted in a perturbation in RCS temperature of approximately 8 degrees Fahrenheit. The finding was evaluated in accordance with IMC 0609, Appendix G, Shutdown Operations Significance Determination Process, and determined to be of very low safety significance (Green) since it did not represent a loss of system safety function of at least a single train for greater than four hours. The finding had a cross- cutting aspect in the area of human performance associated with conservative bias because PG&E personnel did not use decision- making practices that emphasize prudent choices over those that are simply allowable. Specifically, despite being authorized to close component cooling water cross connect valves by the work control process, PG&E personnel did not question the impact of their actions on shutdown cooling (H.14 ).
05000482/FIN-2017007-01	2017Q2	Wolf Creek	Failure to Maintain Effectiveness of the Emergency Plan upon Loss of Containment High Radiation Monitoring	The inspectors identified a Green non-cited violation of 10 CFR 50.54(q)(2) which requires that a holder of a nuclear power plant operating license follow and maintain the effectiveness of an emergency plan that meets the requirements in Appendix E of this part and the risk significant planning standards of 10 CFR 50.47(b). Specifically, from March 7, 2017, to July 12, 2017, Wolf Creek Generating Stations response to the inoperability of containment high radiation monitors failed to restore capability to classify emergency action levels during a loss-of-coolant accident or main-steam-line-break accident. In response to this issue, the licensee provided additional radiation survey monitoring measures and correlations to monitor radiation in the containment building. This finding was entered into the licensees corrective action program as Condition Report CR-114274. The inspectors determined that the failure to maintain the effectiveness of the emergency action level schemes by providing adequate preplanned methods and compensatory measures for the loss of the containment high range radiation monitors in accordance with 50.54 (q)(2) was a performance deficiency. This finding was determined to be more thanminor because it was associated with emergency response organization performance attribute of the Emergency Preparedness cornerstone and adversely affected the cornerstone objective. Specifically, the failure to maintain the effectiveness using appropriate compensatory measures adversely affected the objective of ensuring the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The finding was determined to be of very low safety significance (Green) because (1) emergency action level schemes were rendered ineffective such that any Site Area Emergency would not be declared for a particular off-normal event, but because of other emergency action levels, an appropriate declaration could be made in a degraded manner; and, (2) the emergency action level classification process would result in an over-classification causing an unnecessary emergency declaration. This finding had a cross-cutting aspect in the area of human performance associated with conservative bias because the licensee failed use decision making-practices that emphasized prudent choices over those that are simply allowable. (H.14)(Section 1R21N)
05000482/FIN-2017007-02	2017Q2	Wolf Creek	Licensee-Identified Violation	Technical Specification 3.3.3, Post Accident Monitoring Instrumentation, required two channels of containment area radiation (high range) detectors to be operable when the unit is in Modes 1, 2, or 3. It also required, for one or more functions with two required channels inoperable, that one of the required channels be restored to Operable within 7 days or initiate action in accordance with Technical Specification 5.6.6. Specification 5.6.6 required that a Post Accident Monitoring Instrumentation Report be submitted within 14 days that outlined the preplanned alternate method of monitoring, the cause of inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to Operable status. Contrary to the above, from 1997 to March 2017, the licensee failed to restore at least one channel of containment high range radiation monitors to operable status, initiate preplanned alternate methods of monitoring the appropriate parameter, or prepare and submit a Post Accident Monitoring Instrumentation Report within 14 days pursuant to Technical Specification 5.6.6. The violation was more than minor because it was associated with the Facilities and Equipment attribute of the Emergency Preparedness Cornerstone and adversely affected the cornerstone objective of ensuring 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. The inspectors determined the significance using Inspection Manual Chapter 0609, Attachment 04, Initial Characterization of Findings, and Inspection Manual Chapter0609, Appendix B, Emergency Preparedness Significance Determination Process, Section 5.4 for failure to comply with risk significant planning standard 10 CFR 50.47(b)(4). The finding was determined to be of very low safety significance (Green) because (1) emergency action level schemes were rendered ineffective such that any Site Area Emergency would not be declared for a particular off-normal event, but because of other emergency action levels, an appropriate declaration could be made in a degraded manner; and, (2) the emergency action level classification process would result in an over-classification causing an unnecessary emergency declaration. The violation was entered into the licensees corrective action program as Condition Reports CR-111440, CR-111536, and CR-113217.
05000382/FIN-2016008-01	2016Q4	Waterford	Failure to Control Nonconforming Parts	The team identified a Green non-cited violation of 10 CFR Part 50,Appendix B, Criterion XV, Nonconforming Materials, Parts, or Components, which occurred when the licensee failed to dedicate commercial-grade relays for use insafety-related applications. After receiving information from a vendor that more than124 relays potentially installed in safety-related applications did not conform to quality assurance standards, the licensee failed to take appropriate steps to accept these Commercial-grade relays as basic components. After discussion with the team, the licensee documented this condition in Condition Report CR-WF3-2016-07710 and initiated actions to ensure compliance with quality assurance requirements. The failure to dedicate commercial-grade relays used asor intended for use asbasic components (in safety-related applications) as required by plant procedures and by10 CFR Part 21 was a performance deficiency. This performance deficiency wasmore-than-minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the objective of ensuring the availability,reliability, and capability of systems that respond to initiating events. Using Inspection Manual Chapter 0609, Appendix A, dated June 19, 2012, the team determined that this finding was of very low safety significance (Green) because it was a deficiency affecting the design or qualification of a structure, system, or component, and operability was maintained.The finding has a conservative bias cross-cutting aspect in the human performance cross cutting area because licensee personnel improperly rationalized the adequacy of the nonconforming components to perform their safety-related functions (H.14). Because this performance deficiency was also a violation that impacted the regulatory process, in that the licensee accepted a change to plant design without appropriate evaluation and notification, it was also evaluated for traditional enforcement. The team determined that the violation was Severity Level IV because it was similar to several examples in Section 6.5.d of the NRC Enforcement Policy.
05000313/FIN-2016008-02	2016Q4	Arkansas Nuclear	Failure to Incorporate NRC Safety Guide 9 Criteria into Surveillance Procedures	Green. The team identified Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, which states, A test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. Additionally, Test results shall be documented and evaluated to assure that test requirements have been satisfied. Specifically, as of December 2, 2016, Units 1 and 2 emergency diesel generator surveillance procedures failed to incorporate the applicable voltage and frequency limits of NRC Safety Guide 9, and did not consistently document or evaluate results to assure test requirements have been satisfied. In response to this issue, the licensee provided the team test results which demonstrated that an immediate safety concern was not present. This finding was entered into the licensees corrective action program as Condition Reports CR-ANO-1-2016-4785 and CR-ANO-2-2016-4257. The team determined that the failure to incorporate the acceptance limits of NRC Safety Guide 9 into surveillance test procedures for emergency diesel generators and assure that test requirements have been satisfied in accordance with 10 CFR Part 50, Appendix B, Criterion XI, Test Control, was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the procedure quality attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences, and would have the potential to lead to a more significant safety concern. Specifically, the failure to incorporate appropriate acceptance criteria in test procedures and assure that the criteria have been satisfied had the potential to lead to a worse condition, if left uncorrected. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. This finding did not have a cross-cutting aspect because the performance deficiency did not reflect current licensee performance.
05000313/FIN-2016008-03	2016Q4	Arkansas Nuclear	Failure to Monitor Startup Transformers 1, 2, and 3 Voltage Regulator/Tap Changer Function	Green. The team identified a Green finding for the failure to meet the surveillance standards of IEEE 308-1971, Criteria for Class 1E Electric Systems for Nuclear Power Generating Stations, Section 5.2.3, Preferred Power Supply. Specifically, from 2001 to December 2, 2016, the licensee failed to monitor the operation of the voltage regulator/load tap changer functions on startup transformers 1, 2, and 3. In response to this issue, the licensee provided reasonable assurance that the voltage regulator/load tap changer was operating properly based on review of plant computer voltage plot data following an Arkansas Nuclear One, Unit 1 trip that occurred on December 14, 2015. This finding was entered into the licensees corrective action program as Condition Reports CR-ANO-C-2016-4777, CR-ANO-C-2016-4879, and CR-ANO-C-2016-5015. The team determined that the failure to monitor startup transformers 1, 2, and 3 voltage regulator/load tap changers to the extent that they are shown to be ready to perform their intended function, in accordance with IEEE Standard 308-1971, was a performance deficiency. The finding was determined to be more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. Specifically, the failure to monitor the adequacy of the voltage supplied from startup transformers 1, 2, and 3 voltage regulator/load tap changer did not ensure that offsite power would be available to perform its necessary functions to provide power to the safety-related mitigation equipment. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. This finding did not have a cross-cutting aspect because the performance deficiency did not reflect current licensee performance.
05000313/FIN-2016008-04	2016Q4	Arkansas Nuclear	Failure to Perform an Adequate Emergency Feedwater Pump Suction Transfer Design Calculation or Testing (EA 2017-017)	Green. The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part that, design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Specifically, prior to December 22, 2016, the licensee failed to verify the adequacy of the emergency feedwater suction transfer procedure by determining if the qualified condensate storage tank will be completely empty of water, possibly causing an air ingestion failure of the Unit 1 emergency feedwater pumps, prior to transferring to the credited safety-related alternate suction source. In response to this issue, the licensee resolved the immediate safety concern by revising the emergency feedwater pump operating procedure, removing the steps that were the cause of the concern. This finding was entered into the licensees corrective action program as Condition Reports CR-ANO-1-2016-5166, CR-ANO-1-2016-5725, and CR-ANO-1-2017-0040. The team determined that the failure to verify the adequacy of the design of the Unit 1 emergency feedwater suction from the qualified condensate storage tank to alternate sources of water by performance of design review, by use of calculational methods, or by performance of a suitable testing program in accordance with 10 CFR Part 50, Appendix B, Criterion III, Design Control, was a performance deficiency. This finding was more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the reliability, availability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to have adequate measures in place to ensure an acceptable design analysis or a suitable test program would verify that the process of transferring emergency feedwater suction from the qualified storage tank to the alternate sources ensures the capability of the Unit 1 emergency feedwater system to perform its safety function. In accordance with Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, dated June 19, 2012, the team determined this finding affected the secondary short term heat removal function of the Mitigating Systems Cornerstone. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the finding represented a loss of the emergency feedwater system and function. Therefore, a detailed risk evaluation was necessary. The senior reactor analyst determined that the change in core damage frequency of this finding was 7 x 10-7 per year, therefore the significance was of very low safety significance (Green). This finding did not have a cross-cutting aspect because the performance deficiency did not reflect current licensee performance.
05000313/FIN-2016008-05	2016Q4	Arkansas Nuclear	Failure to Ensure Safety Systems Would Survive Sustained Degraded Voltage Conditions	Green. The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Specifically, from December 17, 1979, to December 2, 2016, the licensee did not verify that the design of the protective devices for the loads required at the beginning of a loss-of-coolant accident were adequate to prevent tripping these devices under degraded voltage conditions, which would render the affected loads non-functional. In response to this issue, the licensee performed a preliminary analysis to determine that the protective overload devices would not cause safety equipment to fail at degraded voltages allowed by technical specifications. This finding was entered into the licensees corrective action program as Condition Reports CR-ANO-C-2016-5027 and CR-ANO-C-2016-5191. The team determined that the failure to ensure that safety-related electrical components would not fail during the allowable time duration of a degraded voltage condition (in accordance with NRC Multi-Plant Action B-23, Position 1.C) was a performance deficiency. The finding was determined to be more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. Specifically, the failure to ensure that the protective devices for the loads required at the beginning of a Loss of Control Accident would not fail under degraded voltage conditions did not ensure that these loads would be available to perform their mitigating functions. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. The team determined that this finding did not have a cross-cutting aspect because the most significant contributor did not reflect current licensee performance.
05000382/FIN-2016008-02	2016Q4	Waterford	Failure to Perform Operability Determinations for Nonconforming Conditions	The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B,Criterion V, Instructions, Procedures, and Drawings, that occurred when the licensee failed on two occasions to perform an operability determination for a nonconforming condition affecting numerous safety-related components. Following receipt of information from a vendor that more than 124 relays potentially installed in safety-related applications did not conform to quality requirements, licensee personnel failed to perform an operability evaluation. Later, during a Part 21 evaluation for the potential defect, the evaluator noted that an operability determination was needed, but failed to initiate the appropriate processes. After discussion with the team, the licensee documented this condition in Condition Report CR-WF3-2016-07710, declared the affected components operable, but degraded, and initiated actions to restore full qualification.Failures to perform an operability determination following identification of a nonconforming condition as required by station procedures were two examples of a performance deficiency.This performance deficiency was more-than-minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the objective of ensuring the availability, reliability, and capability of systems that respond to initiating events. Using Inspection Manual Chapter 0609, Appendix A, dated June 19, 2012,the team determined that this finding was of very low safety significance (Green) because it did not represent the actual loss of function of any system or train. The finding has an identification cross-cutting aspect in the problem identification and resolution cross-cuttingarea because licensee personnel failed to recognize a nonconforming condition as a condition adverse to quality (P.1).
05000382/FIN-2016008-03	2016Q4	Waterford	Failure to Include Appropriate Quantitative Acceptance Criteria for the Reconstituted Feedwater/Emergency Feedwater Monitoring Plan Associated with Steam Generator Replacement Induced Vibration	The team identified a finding of very low safety significance and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to include appropriate quantitative accept an cecriteria for determining that important activities have been satisfactorily accomplished. Specifically, the licensees reconstituted feedwater and emergency feedwater system monitoring plan, which was created to monitor both systems vibrations following the sites steam generators replacement, did not include a range for acceptable vibration levels for all The team identified a finding of very low safety significance and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to include appropriate quantitative acceptance criteria for determining that important activities have been satisfactorily accomplished. Specifically, the licensees reconstituted feedwater and emergency feedwater system monitoring plan, which was created to monitor both systems vibrations following the sites steam generators replacement, did not include a range for acceptable vibration levels for all
05000313/FIN-2016008-01	2016Q4	Arkansas Nuclear	Failure to Verify the Adequacy of Motor Operated Valve Thermal Overload Devices	Green. The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Specifically, prior to December 2, 2016, the licensee failed to use appropriate assumptions in thermal overload device calculations and failed to establish a suitable periodic test program for safety-related Unit 1 motor operated valve thermal overload device trip setpoints, as discussed in Regulatory Guide 1.106, Regulatory Position C.2. In response to this issue, the licensee demonstrated reasonable assurance of operability by using the results of the 18-month high pressure injection system valve testing which required multiple stroking of block valves to obtain various flows without tripping the thermal overload devices. This finding was entered into the licensees corrective action program as Condition Reports CR-ANO-C-2016-5017 and CR-ANO-1-2016-5130. The team determined that the failure to meet the intent of Regulatory Guide 1.106, Regulatory Position C.2 was a performance deficiency. The finding was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. Specifically, the failure to verify the adequacy of the design and perform suitable testing for thermal overload device setpoint drift did not ensure that the safety-related motor operated valves would be available to throttle the associated system flows during a design basis accident. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. This finding had a cross-cutting aspect in the area of problem identification and resolution associated with evaluations because the licensee failed to thoroughly evaluate issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, the licensee failed to thoroughly evaluate Condition Report CR-ANO-1-2016-0778 which documented NRC inspector concerns associated with design and testing of motor operated valve thermal overload devices (P.2).
05000313/FIN-2016008-06	2016Q4	Arkansas Nuclear	Readiness to Cope with External Flooding	Green. The team identified three examples of a Green non-cited violation of 10 CFR Part 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. Specifically, prior to December 2, 2016, Unit 1 Operating Procedure OP 1203.025, Natural Emergencies, Revision 60 and Unit 2 Operating Procedure OP 2203.008 Natural Emergencies, Revision 42 failed to ensure all actions required to establish external flood protection, as specified by flood protection design basis engineering report CALC-ANOC-CS-00003, Revision 00 were implemented. This issue was entered into the licensees corrective action program as Condition Report CR-ANO-2-2016-4265. The licensees failure to prescribe procedures appropriate to the circumstances for combating emergencies or other significant acts of nature such as flooding was a performance deficiency. The performance deficiency was more than minor because it was associated with the procedure quality attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences, and would have the potential to lead to a more significant safety concern. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it does not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding, or severe weather initiating event. This finding had a cross-cutting aspect in the area of problem identification and resolution associated with identification because the licensee failed to identify issues, completely, accurately, and in a timely manner in accordance with the corrective action program. Specifically, the licensee failed to identify these deficiencies during a review of these same procedures as part of actions to close significant performance deficiencies as documented in Arkansas Nuclear One Area Action Plan FP-6 (P.1).
05000382/FIN-2016008-04	2016Q4	Waterford	Departure from Approved Method to Determine Steam Generator Internal Loads During Main Steam Line Break	The team identified a Severity Level IV non-cited violation of 10 CFR 50.59(c)(2),Changes, Tests, and Experiments, for the licensees failure to obtain a license amendment prior to implementing a proposed change, test, or experiment that would result in a departure from a method of evaluation described in the final safety analysis report (as updated) used in establishing the design bases or in the safety analyses. Specifically, the licensee departed from their approved CEFLASH-4A methodology to determine steam generator internal differential loads caused by a main steam line break to an unapproved TRANFLOW methodology. In response to this issue, the licensee entered the issue into the corrective action program as Condition Report CR-WF3-2016-07639 and initiated actions to prepare a new evaluation under current regulatory guidelines or to submit a license amendment request to the NRC.The licensees failure to obtain a license amendment prior to implementing a change that resulted in a departure from a method of evaluation described in the final safety analysis report (as updated) used in establishing the design bases or in the safety analyses, as required by 10 CFR 50.59(c)(2) was a violation. In accordance with the NRC Enforcement Manual, violations of 10 CFR 50.59 are not processed through the Reactor Oversight Process significance determination process because this violation potentially impacted the ability of the NRC to perform its regulatory oversight function. Therefore, this violation was processed through traditional enforcement examples of Section 6.1 of the NRC Enforcement Policy. This violation was more-than-minor because there was a reasonable likelihood that the change would require NRC review and approval prior to implementation, similar to the more-than-minor example of a change in requirements in the NRC Enforcement Manual,Appendix E, Minor Violations Examples, dated September 9, 2013. In accordance with the NRC Enforcement Policy, the significance determination process was used to inform the significance of the failure to obtain a license amendment prior to implementing a proposed change. The departure from the original CEFLASH-4A method to the TRANFLOW method to determine differential loads on steam generator internal structures following a main steam line break event was associated with the design control attribute of the Barrier Integrity Cornerstone 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. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012,Exhibit 1, Initiating Events Screening Questions, the issue screened as having very low safety significance (Green) because the issue would not result in the complete or partial loss of a support system that contributes to the likelihood of an initiating event, or result in the steam generators violating accident leakage performance criterion. Since the violation was determined to be Green in the significance determination process, the traditional enforcement violation was determined to be a Severity Level IV violation, consistent with the example in paragraph 6.1.d(2) of the NRC Enforcement Policy. Traditional enforcement violations are not assessed for cross-cutting aspects.
05000445/FIN-2016007-01	2016Q3	Comanche Peak	Failure to Update Final Safety Analysis Report Section 8.3.1.1.11	The inspectors identified a Severity Level IV violation of 10 CFR 50.71(e) which states, in part, that the licensee shall update periodically the final safety analysis report originally submitted as part of the application for the license, to assure that the information included in the report contains the latest information developed. The submittal shall include the effects of all changes made in the facility or procedures as described in the final safety analysis report, or all safety analyses and evaluation performed by the licensee either in support of approved license amendments or in support of conclusions that changes did not require a license amendment in accordance with 10 CFR 50.59 (c)(2). Specifically, from October 9, 2012, to September 29, 2016, the licensee did not include the effects of changes to the K300 voltage relay setpoint or the safety evaluation in submittals to the Final Safety Analysis Report, Section 8.3.1.1.11, that supported the conclusion that the changes did not require a license amendment. In response to this issue, the licensee planned a corrective action to initiate a licensing document change request to update the final safety analysis report. This finding was entered into the licensees corrective action program as Condition Report CR-2016-008177. The inspectors determined that the licensees failure to initiate a Licensing Document Change Request, in accordance with Procedure STA-116, Maintenance of CPNPP Licensing Basis Documents, Operating License conditions and Technical Specifications, Revision 14, Instruction 6.1, to update the Final Safety Analysis Report, Section 8.3.1.1.11, for the setpoint revision of the K300 voltage relays was a performance deficiency. In accordance with Inspection Manual Chapter 0612, Appendix B, Issue Screening, dated September 7, 2012, this was determined to be a minor performance deficiency. This violation was evaluated using the traditional enforcement process because it had the potential for impacting the NRCs ability to perform its regulatory oversight function. The reactor oversight processs significance determination process does not consider violations that impact the NRCs regulatory oversight function. This violation was determined to be a Severity Level IV violation, consistent with the example in paragraph 6.1.d.3 of the NRC Enforcement Policy, dated August 1, 2016. Specifically, the licensee failed to update the final safety analysis report as required by 10 CFR 50.71(e), but the lack of up-to-date information has not resulted in any unacceptable change to the facility or procedures. The inspectors determined that this violation did not have a cross-cutting aspect because traditional enforcement violations are not assessed for cross-cutting aspects.
05000416/FIN-2016002-01	2016Q2	Grand Gulf	Failure to Maintain Secondary Containment Operable during Roof Inspections	The inspectors identified a Green, non-cited violation of Technical Specification Surveillance Requirement 3.0.1, for the failure to meet Surveillance Requirement 3.6.4.1.1 and declare Limiting Condition for Operation 3.6.4.1 not met. Specifically, the licensee did not maintain the enclosure building hatch penetration in the closed position as required by Surveillance Requirement 3.6.4.1.1, which resulted in secondary containment being inoperable. The licensee restored compliance by closing the hatch following the surveillance, and put corrective actions in place to control the enclosure building hatch penetration in a closed position except for entry and exit for the inspection. This finding was entered into the licensees corrective action program as Condition Report CR-GGN-1-2016-03707. The failure to declare that Limiting Condition for Operation 3.6.4.1 was not met when the enclosure building hatch was maintained in the open position was a performance deficiency. This performance deficiency is more than minor, and therefore a finding, because it is associated with the configuration control attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (secondary containment) protect the public from radionuclide releases caused by accidents or events. Specifically, on April 7, 2016, the licensee did not maintain the enclosure building hatch penetration in the closed position as required by SR 3.6.4.1.1, which resulted in secondary containment being inoperable. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings AtPower, and Inspection Manual Chapter 0609, Appendix A, Exhibit 3, Barrier Integrity Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because the finding only represented a degradation of the radiological barrier function provided for the control room, or auxiliary building, or spent fuel pool, or standby gas treatment (SBGT) system (BWR). This finding has a cross-cutting aspect in the area of human performance associated with documentation, in that, the organization failed to create and maintain complete, accurate and up-to-date documentation. Specifically, Work Order 52671695 for implementing the roof inspection was not complete and accurate with regards to the impact on operability of secondary containment when leaving the enclosure building hatch penetration open during inspection activities.
05000416/FIN-2016002-02	2016Q2	Grand Gulf	Failure to Provide Detailed Work Instructions Resulted in a Reactor Scram	The inspectors reviewed a Green, self-revealed finding of Procedure EN-WM-105, Planning, Revision 16, for the failure to ensure Work Order 397549 provided detailed instructions for performing troubleshooting on the B phase of the main transformer. Specifically, Work Order 397549 did not contain detailed instructions for performing troubleshooting on the B phase of the main transformer, which resulted in an incorrect current transformer ratio and subsequent reactor scram. The licensees corrective actions were to incorporate more detailed instructions to the work order, repair the improper wiring, and restore the main transformer prior to transitioning from Mode 3 to Mode 1. Inspectors did not identify a violation of regulatory requirements associated with this finding. This finding was entered into the licensees corrective action program as Condition Report CR-GGN-1-2016-02950. The failure to ensure Work Order 397549 provided detailed instructions for performing troubleshooting on the B phase of the main transformer in accordance with Procedure EN-WM-105 was a performance deficiency. This performance deficiency is more than minor, and therefore a finding, because it is associated with the procedure quality 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, on March 29, 2016, the licensee failed to ensure Work Order 397549 provided detailed instructions for performing troubleshooting on the B phase of the main transformer, which resulted in an incorrect current transformer wiring ratio and subsequent reactor scram. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, and Inspection Manual Chapter 0609, Appendix A, Exhibit 1, Initiating Events Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because the finding did result in a reactor trip, but did not result in the 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 cross-cutting aspect in the area of human performance associated with field presence, in that, senior managers failed to ensure supervisory and management oversight of work activities, including contractors and supplemental personnel. Specifically, while performing Work Order 397549, the licensee did not have contractor oversight established, and the contract workers performed troubleshooting without detailed instructions to ensure work was performed properly.
05000285/FIN-2016001-01	2016Q1	Fort Calhoun	Implementing a Procedure Change for Alternative Shutdown Cooling that would have Required NRC Approval	The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.59, Changes, Tests, and Experiments, for the failure to recognize that a change to the facility as described in the Updated Safety Analysis Report would require prior NRC review and approval. Specifically, the 10 CFR 50.59 evaluation revised a site procedure, without NRC approval, to substitute automatic flow control of shutdown cooling flow and temperature with manual control using the low pressure safety injection loop injection valves. The licensees corrective actions included revising the affected procedure to reflect the original automatic flow control. The licensee entered this issue in the corrective action program as Condition Report 2013-15342. The licensees failure to implement the requirements of 10 CFR 50.59 and adequately evaluate changes to determine if prior NRC approval is required was a performance deficiency. Because this violation had the potential to impact the NRCs ability to perform its regulatory function, the inspectors evaluated the violation using traditional enforcement. In accordance with Section 2.1.3.E.6 of the NRC Enforcement Manual, the team evaluated this finding using the significance determination process to assess its significance. The inspectors 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, dated July 1, 2012. Using Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated July 1, 2012, the finding was determined to have very low safety significance (Green) because it: (1) was not a deficiency affecting the design or 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 technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and (4) did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees Maintenance Rule program. Therefore, in accordance with Section 6.1.d.2 of the NRC Enforcement Policy, the inspectors characterized this performance deficiency as a Severity Level IV violation. The inspectors determined that a cross-cutting aspect was not applicable because the issue involving the failure to perform an adequate 10 CFR 50.59 evaluation was strictly associated with a traditional enforcement violation.
05000285/FIN-2016001-02	2016Q1	Fort Calhoun	Licensee-Identified Violation	Technical Specification 2.6(1) requires containment integrity to be maintained unless the reactor is in a cold or refueling shutdown condition. If containment integrity is not maintained and the reactor does not meet these cold or refueling shutdown conditions, then containment integrity must be restored within one hour or the reactor is required to be in hot shutdown within the next six hours. From November 22, 2013, through June 27, 2014, a test connection cap was left off of a containment penetration which constituted a loss of containment integrity. Upon discovery of this condition on June 27, 2014, the licensee entered Technical Specification 2.6(1) and Abnormal Operating Procedure 12 for loss of containment integrity. The cap was re-installed and containment integrity was restored within one hour. The violation is more than minor because it is associated with the configuration control attribute of the Barrier Integrity Cornerstone. Failure to install the containment penetration cap following local leak rate testing on November 22, 2013, resulted in a loss of containment integrity until it was discovered missing on June 27, 2014. This adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (i.e., containment) protect the public from radionuclide releases caused by accidents or events. The violation was reviewed by a Senior Reactor Analyst and was determined to be of very low safety significance because the test connection fitting was a 14-inch diameter opening. Inspection Manual Chapter 0609, Significance Determination Process, Appendix H, identifies that small lines (less than 1 to 2 inches in diameter) would not generally contribute to large early release frequency. Therefore, this finding screens to Green. The licensee entered the issue into their corrective action program as Condition Report 2014-07958.
05000445/FIN-2015005-02	2015Q4	Comanche Peak	Failure to Take Appropriate Maintenance Rule Corrective Actions for the 6.9 kV AC System	The inspectors identified a non-cited violation of 10 CFR Part 50.65(a)(1), for the failure to establish goals that provide reasonable assurance that the 6.9 kV electrical distribution system is capable of fulfilling its intended functions. Specifically, the 6.9 kV electrical distribution system had been in maintenance rule (a)(1) status since 2009 due to the failure of breakers to close on demand. Subsequently, in 2013 and 2015 there were additional breaker failures, which exceeded the established performance criteria, and were due to causes not previously evaluated. These additional failures were determined to be due to inadequate maintenance, but the licensee did not re-evaluate the established goals and revise the corrective actions to address these additional failures. The licensee implemented corrective actions to re-evaluate the goals and corrective actions for the 6.9 kV AC system. The licensee entered this issue into the corrective action program as Condition Report CR-2015-009077. The licensees failure to evaluate existing goals and corrective actions for a system that did not meet established performance goals 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 and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to take appropriate corrective actions adversely affected the reliability of a system scoped in the plant's maintenance rule program. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012, the finding was determined to be 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 finding has a human performance cross-cutting aspect associated with procedure adherence, in that, the licensee failed to follow maintenance rule implementing procedures. (H.8).
05000445/FIN-2015005-04	2015Q4	Comanche Peak	Failure to Identify Conditions Adverse to Quality	The inspectors identified two examples of a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to identify conditions adverse to quality. Specifically, in two separate instances involving extent of condition reviews for grease on 6.9 kV breaker stabs and degraded piping in the Unit 1 service water system, the licensee failed to identify conditions adverse to quality that were reasonably within their ability to identify. As a result, the licensee failed to; 1) identify 24 additional breakers that were in a degraded condition due to grease on secondary stabs, and 2) identify a section of service water piping that was below the ASME minimum wall thickness. The licensee implemented immediate corrective actions by entering the issues into the corrective action program for resolution and performed an operability determination for the identified degraded conditions. The licensee entered these issues into the corrective action program as Condition Reports CR-2015-009992 and CR-2015-010120. The licensees failure to identify conditions adverse to quality for quality related systems 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 and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to identify degraded conditions could affect the reliability or availability of multiple safety related systems. Using Inspection Manual Chapter 0609, Attachment 04, Initial Characterization of Findings, and Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 1, Initiating Events Screening Questions, dated June 19, 2012, the finding was determined to be of very low safety significance (Green) because the finding is a deficiency affecting the design or qualification of a mitigating SSC, but the SSC maintained its operability. The finding has a problem identification and resolution cross-cutting aspect associated with evaluation, in that, the licensee failed to thoroughly evaluate issues to ensure that resolutions address extent of conditions. Specifically, the licensee failed to adequately consider the extent of the degraded conditions on similar safety related components (P.2).
05000445/FIN-2015005-05	2015Q4	Comanche Peak	Failure to Follow Procedure When Disabling A Hazard Barrier	The inspectors identified a finding associated with the licensees failure to follow procedural requirements for disabling a hazard barrier. Specifically, Station Procedure STA 696, Hazard Barrier Controls, Revision 2, requires that appropriate temporary barriers be prescribed when a hazard barrier is impaired. However, in support of an auxiliary, safeguards and fuel building negative pressure test, the licensee failed to follow Procedure STA 696 and incorrectly credited alternate doors to protect safety-related equipment from the effects of a high-energy line break when disabling the primary hazard barrier. The licensee implemented corrective actions to correctly assess the activity and implemented appropriate risk management actions. The licensee entered the finding into corrective action program as Condition Report CR-2015-005583. The licensees failure to follow station procedures when crediting temporary hazard barriers 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, opening the high energy line break door without an appropriate temporary barrier in place removed a credited barrier for safety-related electrical equipment. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012, the finding was determined to be 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 safetysignificant 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 would have occurred more than three years ago, and is not reflective of current licensee performance.
05000445/FIN-2015005-07	2015Q4	Comanche Peak	Licensee-Identified Violation	Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality shall be accomplished in accordance with documented instructions, procedures, or drawings, of a type appropriate to the circumstances. Station Procedure STI-442.01, Operability Determination and Functionality Assessment Program, Revision 3, an Appendix B quality related procedure, provides instructions for evaluating the operability of safety-related components. Procedure STI-442.01, Step 6.1, requires, in part, that when a potential degraded or nonconforming condition is identified, the shift manager should ensure the operability determination process is initiated to determine the operability of the structure, system or component. Contrary to the above, on July 26, 2015, when a potential degraded or nonconforming condition was identified, the shift manager failed to ensure the operability determination process was initiated to determine the operability of the structure, system or component. Specifically, the licensee failed to adequately assess and demonstrate the operability of Unit 1 train B containment spray system when a degraded condition was identified. 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: did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic event, and (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 violation was entered into the licensees corrective action program as Condition Report CR-2015-006785.
05000445/FIN-2015005-08	2015Q4	Comanche Peak	Licensee-Identified Violation	Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality shall be accomplished in accordance with documented instructions, procedures, or drawings, of a type appropriate to the circumstances. Station Procedure STI-442.01, Operability Determination and Functionality Assessment Program, Revision 3, an Appendix B quality related procedure, provides instructions for evaluating the operability of safety-related components. Procedure STI-442.01, Step 6.1, requires, in part, that when a potential degraded or nonconforming condition is identified, the shift manager should ensure the operability determination process is initiated to determine the operability of the structure, system or component. Contrary to the above, on October 14, 2015, when a potential degraded or nonconforming condition was identified, the shift manager failed to ensure the operability determination process was initiated to determine the operability of the structure, system or component. Specifically, the licensee failed to enter the operability determination process, as required by Station Procedure STI-442.01, step 6.1, when a degraded or nonconforming condition was identified associated with incorrectly performed visual examination required by the ASME code. 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: did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic event, and (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 safetysignificant for greater than 24 hours in accordance with the licensees maintenance rule program. The violation was entered into the licensees corrective action program as Condition Report CR-2015-009586.
05000445/FIN-2015005-09	2015Q4	Comanche Peak	Licensee-Identified Violation	Title 10 CFR Part 50, Appendix B, Criterion XI, "Test Control," states, in part, A test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. Contrary to the above, on October 20, 2015, the licensee failed to incorporate adequate acceptance limits in a quality-related written procedure that demonstrates components will perform satisfactorily. Specifically, the licensee failed to use appropriate acceptance limits for integrated testing of the station emergency diesel generators. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012, the finding was determined to be 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 safetysignificant for greater than 24 hours in accordance with the licensees maintenance rule program. The violation was entered into the licensees corrective action program as Condition Report CR-2015-009990.
05000445/FIN-2015005-01	2015Q4	Comanche Peak	Incorrect Visual Resolution Requirements in Augmented Dissimilar Metal Weld Visual Examination Procedures	The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion IX, Control of Special Processes, because the licensee failed to assure that visual examination activities for the reactor vessel dissimilar metal nozzle welds and bottommounted instrumentation nozzles were accomplished in accordance with the visual acuity requirements of ASME Code Case N-722-1. In response to the issue, for Unit 2, the licensee scheduled reexamination of the welds prior to the end of the outage, and, for Unit 1, performed a reasonable degradation evaluation to determine that reexamination of the welds could be delayed to the next outage. This finding was entered into the corrective action program as Condition Report 2015-009586. The inspectors determined that the failure to assure visual examination activities were accomplished in accordance with the visual acuity requirements of ASME Code Case N-722-1 was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Specifically, routinely performing examinations with incorrect visual acuity requirements of N-722-1 has the potential to lead to missed opportunities to identify and correct relevant indications in reactor coolant system pressure boundaries. In accordance with Inspection Manual Chapter MC 0609, Attachment 4, Significance Determination Process Initial Characterization, the inspectors determined that this finding affected the Initiating Events cornerstone as a primary system LOCA initiator contributor. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 1, Initiating Events Screening Questions, the finding screened as having very low safety significance (Green) because after a reasonable assessment of degradation, the finding did not result in exceeding the RCS leak rate for a small LOCA and did not affect other systems used to mitigate a LOCA. The finding does not have a crosscutting aspect because the most significant contributor is not reflective of current licensee performance.
05000445/FIN-2015005-03	2015Q4	Comanche Peak	Inadequate Compensatory Measures for Seismic Monitoring System Maintenance	The inspectors identified a non-cited violation of 10 CFR 50.54(q)(2) for a failure to meet planning standard 10 CFR 50.47(b)(4) during periodic outages of the seismic monitoring system. Specifically, during planned maintenance on the seismic monitoring system, inspectors determined that the system would not be able to perform its function of alerting control room staff of an entry condition into the emergency action levels for a seismic event, and the specified compensatory measures were not adequate. The licensee implemented correction actions to establish viable compensatory measures for periods when the seismic monitoring system is unavailable. The licensee entered these issues into corrective action program as Condition Report CR-2016-000091. The licensees failure to maintain the effectiveness of their emergency plan was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, 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-2016-000091. The inspectors determined that this finding has a problem identification and resolution crosscutting aspect associated with resolution, because the licensee failed to take appropriate corrective action after they recognized the inadequacy of their compensatory measures (P.3).
05000445/FIN-2015005-06	2015Q4	Comanche Peak	Failure to Barricade High Radiation Areas	The inspector identified a non-cited violation (NCV) of Technical Specification 5.7.1.a, with two examples, associated with not barricading High Radiation Areas (HRAs) with dose rates not exceeding 1.0 rem/hour at 30 centimeters from the radiation source. Specifically, access to the HRA containment trashracks and access to the HRA reactor cavity before flood up were not barricaded to prevent entry. The licensee took immediate corrective action to barricade the associated HRAs to restrict access and entered this issue into the corrective action program as CR-2015-009095 and CR-2015-009303. The failure to barricade high radiation areas in accordance with TS 5.7.1.a was a performance deficiency. The inspector determined that the performance deficiency was more than minor, and therefore a finding, because it impacted the program and process attribute of the Occupational Radiation Safety Cornerstone and adversely affected the cornerstone objective to ensure adequate protection of worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Specifically, not barricading HRAs could lead to inadvertent worker entry into high dose rate areas without knowledge of the radiological conditions. The finding was assessed using IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008, and was determined to be of very low safety significance (Green) because the problem was not an ALARA planning issue; there was no overexposure, nor substantial potential for an overexposure; and the licensees ability to assess dose was not compromised. The finding was associated with a crosscutting aspect of Resolution in Problem Identification and Resolution area. Specifically, the organizations corrective actions to address HRA issues raised by Nuclear Oversight, the NRC and independent assessments in a timely manner commensurate with their safety significance have not been effective (P.3).
05000416/FIN-2015007-03	2015Q4	Grand Gulf	Lack of Coordination of Division III HPCS Switchgear 127N Undervoltage Relays	The following issues were discussed during the inspection; however, the team must review additional information provided by the licensee to determine whether these issues result in a more than minor performance deficiency or a violation of NRC requirements. In accordance with Inspection Manual Chapter 0612, this issue will be characterized as an unresolved item. The incoming offsite power supply circuit breakers for Division III 4160 V switchgear 17AC are equipped with 127N undervoltage relays. According to Drawing E-1009, One Line Meter and Relay Diagram, 4.16kV E.S.F. System Bus 17AC, Revision 9, these relays Trip incoming breaker to bus & start diesel. According to drawing E-0121-005, Summary of Relay Settings (ESF) 4.16 kV Bus 17AC and Diesel Gen 13, Revision 7, these relays are set with a 0 second time delay. This instantaneous time response potentially results in lack of coordination of the 127N undervoltage relays with high voltage system protective relays, switchgear overcurrent relays, and loss of voltage relays, thus preventing the other relays from performing their credited design functions. Protective devices that the 127N undervoltage relays are potentially not coordinated with are as follows: Protective relays associated the main transformer and its output circuit: Lack of coordination between the 127N undervoltage relays and the protective relays associated with the main transformer and its output circuit can result in coincident loss of two alternating current power supplies, contrary to the requirements of General Design Criterion 17. Grand Gulf Updated Final Safety Analysis Report Section 8.3.1.2.1 states: The degree of reliability of the power sources required for safe shutdown is considered very high due to independence and ample redundancy; it equals or exceeds all the requirements of Criterion 17. General Design Criterion 17 states: Provisions shall be included to minimize the probability of losing electric power from any of the remaining supplies as a result of, or coincident with, the loss of power generated by the nuclear power unit... Contrary to this General Design Criterion 17 requirement, a fault on the main transformer or its high voltage connection to the transmission system could cause coincident loss of electric power from both the main generator and the offsite power supply to Division III. The protective relaying on the main transformer and its output circuit is designed to initiate tripping of the main generator and isolation of the area of the fault without causing any cascading failures. However, the 127N undervoltage relay for the Division III 4160 V switchgear would also respond spuriously to the momentary voltage dip caused by the fault and cause loss of the offsite power supply to Division III. Transmission system bus protective relays: Grand Gulf Updated Final Safety Analysis Report Section 8.2, Offsite Power System, Subsection 8.2.2.1, Availability Considerations states: Short circuits on a section of a bus are isolated without interrupting service to any circuit other than that connected to the faulty bus section. Contrary to this requirement, the instantaneous setting of the 127N relays would not coordinate with the transmission bus protective relays and would react to the momentary voltage dip caused by a transmission system bus fault, resulting also in spurious loss of the offsite power supply to Division III. Loss of Voltage Relays: In addition to the 127N undervoltage relays, Division III high pressure core spray 4160 V switchgear 17AC is equipped with 127S1, S2, S3, and S4 loss of voltage relays and 127 1A, 1B, 2A, and 2B degraded voltage relays, which also trip the offsite power supply to 17AC upon actuation. NRC Regulatory Issue Summary 2011-12, Adequacy of Station Electric Distribution System Voltages, Revision 1, describes one of the functions of the degraded voltage relay time delay as follows: The time delay shall override the effect of expected short duration grid disturbances, preserving availability of the offsite power source(s). The same principle is relevant to the other undervoltage relays that automatically trip the switchgear offsite power supply. This conclusion is consistent with Grand Gulf Updated Final Safety Analysis Report Section 8.3.1.1.2.3, which states: Protective devices of Class 1E systems, particularly the ECCS, are set to maintain continuity of power as long as possible short of causing a derangement of the equipment. However, the 127N undervoltage relays do not preserve availability of power as long as possible in the event of harmless transmission grid voltage transients, such as those caused by lightning strikes and normally-cleared faults on transmission lines, because their instantaneous setting miscoordinates with the time delay setting of the Technical Specification credited loss of voltage relays 127S1, S2, S3, and S4. According to Technical Specification Table TR 3.3.8.1-1, the 127S1, S2, S3, and S4 loss of voltage relays have a time delay setting of 2.3 seconds. According to Section 6.17 of calculation JC-Q1P81-90027, Division III Loss of Bus Voltage Setpoint Validation (T/S 3.3.8.1), Revision 2, Spurious segregation from the offsite source is prevented by the time delay function. However, since the non- Technical Specification 127N relays react instantaneously to trip the offsite power source during momentary voltage dips, their 0-second time delay setting invalidates this credited design function of the 2.3-second time delay of the 127S1, S2, S3, and S4 loss of voltage relays. Therefore, spurious segregation from the offsite source is not prevented, and a vulnerability exists for unnecessary loss of offsite power events initiated by, and subsequent to, harmless voltage transients from the transmission system. An actual event of this type occurred on April 2, 2012, as described in Licensee Event Report 2012-003. A lightning strike on a 500 kV transmission circuit resulted in actuation of the instantaneous 127N relay and unnecessary loss of the offsite power supply to the Division III electrical distribution system. Switchgear 17AC offsite power supply circuit breaker overcurrent relays: Grand Gulf Updated Final Safety Analysis Report Section 8.3.1.1.4.2.5.3, HPCS Class 1E Electrical Equipment Circuit Protection states: Emphasis is given in preserving function and limiting loss of Class 1E equipment function in situations of power loss and equipment failure. Contrary to this statement, the instantaneous setting of the 127N undervoltage relays prevents the offsite power supply circuit breaker overcurrent relays from preserving function and limiting loss of Class 1E equipment function in the event of a switchgear bus fault. Switchgear 17AC offsite power supply circuit breakers are equipped with 151B overcurrent relays that, when actuated, trip and lockout the switchgear supply breakers. The purpose of the lockout function is to prevent attempted reenergization of a faulted bus. However, due to the instantaneous response time of the 127N undervoltage relays, the fault would be cleared and the bus deenergized on the undervoltage signal before the overcurrent relays could respond and initiate the bus lockout signal. This would result in automatic starting of the Division III diesel generator, closure of the diesel generator output breaker onto the faulted bus, and the potential for damage to the diesel generator and further damage to the switchgear. Switchgear 17AC feeder circuit breaker overcurrent relays: The circuit breakers for feeders downstream of switchgear 17AC are equipped with 150/151M and 150/151T overcurrent relays that are designed to isolate downstream faults locally. Grand Gulf Updated Final Safety Analysis Report Section 8.3.1.1.2.3, Control Power and Circuit Protection states: A complete analysis of the application and coordination of the protective devices on Class 1E distribution has been conducted. This analysis shows that under design operation of these devices, faults, and undervoltages will be detected and corrected at the lowest level of distribution. Referring to the 150/151M and 150/151T overcurrent relays, Grand Gulf Updated Final Safety Analysis Report Section 8.3.1.1.4.2.5.3 states: Relay settings are coordinated in such a way that interference of service is not communicated to a higher level involving equipment other than that immediately affected by the fault or overload. Contrary to these requirements, the licensee failed to perform a coordination analysis or to ensure that interference of electrical service is limited as described. The voltage dip caused by a fault on a 4160 V circuit downstream of switchgear 17AC would be detected by the 127N relay, which would react instantaneously to trip the 17AC switchgear offsite power supply circuit breaker rather than isolating the fault locally at the downstream circuit breaker. This is contrary to the design criterion that the fault be detected and corrected at the lowest level of distribution and maintain continuity of power to the switchgear. These issues were entered into the licensees corrective action program as Condition Report CR-GGN-2015-4973.
05000529/FIN-2015002-02	2015Q2	Palo Verde	Failure to Take Timely Corrective Actions to Prevent Charging Pump Discharge Bladder Failure	The inspectors reviewed a self-revealing non-cited violation of 10 CFR Part 50 Appendix B, Criterion XVI for the failure to take timely corrective actions associated with failure of the discharge pulsation dampener poppet valves in the positive displacement charging pump. The charging system. is designated as quality related for its function to provide a boration flowpath to the reactor coolant system. Specifically, following the investigation of a degrading discharge dampener bladder on the Unit 2 charging pump E and the discovery that the poppet valve stem was galled and stuck in the poppet valve seat, the licensee incorrectly concluded that routine monthly monitoring and the 5-year bladder replacement maintenance would identify further failures in the other charging system trains. The licensee entered this issue into the corrective action program as Condition Report 15-4230. Failure to take timely corrective actions to replace the charging pump discharge dampener poppet valve assemblies was a performance deficiency. The performance deficiency was more-than-minor and is a finding because it is associated with the equipment performance attribute and directly affected the Initiating Event 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 failure to correct this condition adverse to quality resulted in a reactor coolant system transient and challenged normal plant operations. Using Manual Chapter 0609, Appendix A, "Significance Determination Process (SDP) for Findings At Power," the inspectors determined the finding was of very low safety significance (Green) because the finding did not result in a reactor trip and the loss of mitigating equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. This finding has an evaluation cross-cutting aspect in the area of problem identification and resolution because the organization did not thoroughly evaluate issues to ensure that resolutions address causes and extent of condition commensurate with their safety significance. Specifically, the corrective actions taken in response to the January 2014 poppet galling event included a number of engineering judgements and assumptions regarding both the degradation mechanism, and the internal workings of the sytem components were used to justify not performing additional poppet assembly inspections. These assumptions were known to be incorrect by uninvolved technical experts inside the licensee and vendor organization. Had those assumptions been properly vetted and verified by vendor or other industry experts at the time, the extent-of-condition inspections likely would have been accelerated (P.2).
05000529/FIN-2015002-05	2015Q2	Palo Verde	Failure to Establish Adequate Procedures to Respond to a Total Loss of Charging Event	The inspectors reviewed a self-revealing non-cited violation of Technical Specification 5.4.1.a, through Regulatory Guide 1.33, Revision 2, Appendix A, Section 6.t, February 1978 for the licensees failure to establish adequate procedures for combating emergencies and other significant events regarding a total loss of charging pumps due to gas binding that affected reactor coolant system pressure and level control. On March 20, 2015, after Unit 2 experienced a total loss of charging, operators relied on a normal operating procedure which did not address how to combat a total loss of charging flow due of gas binding from a failed discharge pulsation dampener. The licensee entered this issue into the corrective action program as Condition Report 15-4230. The failure to provide adequate procedures for combating emergencies and other significant events regarding a total loss of charging pumps due to gas binding that affected reactor coolant system pressure control was a performance deficiency. The performance deficiency was more-than-minor and is a finding because it is associated with the procedure quality attribute and directly affected the Initiating Event 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 lack of adequate procedural guidance challenged reactor operators during the loss of charging event. In accordance with Inspection Manual Chapter 0609, Appendix A, "Significance Determination Process (SDP) for Findings AtPower," the performance deficiency was determined to be of very low safety significance (Green) because the finding did not result in a reactor trip and the loss of mitigating equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. This finding did not have a cross-cutting aspect because the most significant contributor did not reflect current licensee performance because the decision to eliminate the abnormal operating procedure and not to train reactor operators was made in 1997.
05000530/FIN-2015002-04	2015Q2	Palo Verde	Notice of Enforcement Discretion of Technical Specification 3.5.3 Emergency Core Cooling System Operating Conditions B and C	(Open) Unresolved Item 05000530/2015002-04, TAC Number MF6276 - NOED Number 15-4-01. Notice of Enforcement Discretion of Technical Specification 3.5.3 Emergency Core Cooling System - Operating Conditions B and C On May 27, 2015, the licensee removed Unit 3 high pressure safety injection train A for planned maintenance. The following morning, during the maintenance, the licensee noted lube oil contamination, and determined that an outboard motor bearing had apparently failed during the last run following maintenance during the last refueling outage which involved disassembling and reassembling the bearing. The licensee identified procedural guidance inadequacies in the reassembly procedure that were the likely cause of the failure. The licensee could not perform required repairs in a controlled manner within the remaining action statement completion time, so on May 29, 2015, the licensee requested a Notice of Enforcement Discretion for a one-time action statement extension of 24 hours to allow time to reassemble and test the replacement bearings prior to restoring operability. The NRC granted that request as NOED 15-4-01. The licensee completed maintenance, testing, and restoration approximately 11 hours into the 24-hour extension window. In accordance with Inspection Manual Chapter 0410, Unresolved Item (URI) 05000530/2015002-04 is opened for NOED 15-4-01, and remains open pending further inspection and disposition in a future inspection report.
05000528/FIN-2015002-01	2015Q2	Palo Verde	Failure to Verify the Design of the Essential Spray Pond System Crosstie Valves	The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, involving the failure to maintain adequate design control measures associated with the ultimate heat sink. Specifically, the essential spray pond crosstie valves did not meet design requirements established in Regulatory Guide 1.117, "Tornado Design Classification," as described in the Updated Final Safety Analysis Report. Consequently, if the crosstie valves were damaged by a tornado, the licensee would not have enough available water inventory to meet the mission time of the essential spray pond system. The licensee has added steps to their emergency operating procedure to instruct operators to open the crosstie valves to achieve and maintain long-term cooling subsequent to a design-basis tornado event, and is evaluating potential plant modifications. The licensee has entered this issue into the corrective action program as Palo Verde Action Request 4633058. The failure to verify the design of the essential spray pond system in accordance with Regulatory Guide 1.117 was a performance deficiency. This performance deficiency was more-than-minor and is a finding because it affected the protection against external factors attribute of the Mitigating Systems Cornerstone objective of ensuring the capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, if the crosstie valves were damaged by a tornado, the licensee would not have enough available water inventory to meet the mission time for one train of the essential spray pond system during accident conditions. The inspectors performed the initial significance determination for the performance deficiency using NRC Inspection Manual 0609, Appendix A, Exhibit 2, "Mitigating System Screening Questions," dated July 1, 2012. The finding required a detailed risk evaluation because it involved the potential loss of a safety system, in that after at least 13 days of spray pond operation, operators were required to open the spray pond cross-connect valve to enable one train of the ultimate heat sink to use both trains of spray pond inventory. A Region IV senior reactor analyst performed a detailed risk evaluation. The design basis accident mission time was 30 days. However, the probabilistic risk assessment mission time was only 24 hours. Since the spray ponds could still perform the probabilistic risk assessment function for the probabilistic risk assessment mission time, this finding was of very low safety significance (Green). The change to the core damage frequency was much less than 1E-7/year. The finding did not contribute to the large early release frequency. Because the most likely cause of the finding does not reflect current licensee performance, no cross-cutting aspect is assigned to this finding.
05000529/FIN-2015002-03	2015Q2	Palo Verde	Failure to Identify and Correct Engineered Safety Features Actuation System Steam Generator Differential Pressure Setpoint Drift	The inspectors reviewed a self-revealing non-cited violation of Technical Specification 3.3.5 condition A.1 for failure to place a failed steam generator differential pressure in bypass or trip. Specifically, on January 11, 2015, after Unit 2 received a steam generator pressure difference setpoint alarm on channel B, operators failed to determine the cause of the alarm. As a result, the auxiliary feedwater actuation signal channel was inoperable for a period of 13 days, which was longer than the technical-specification allowed outage time of one hour, during which time the failed channel would provide a false negative under valid actuation setpoint conditions. The licensee entered this condition in their corrective action program and performed a root cause evaluation under Condition Report Disposition Request 4618033. The failure to provide adequate alarm procedures was a performance deficiency. The performance deficiency was more-than-minor and is 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 control room operators did not have an alarm response procedure for plant monitoring system (RJ) alarm on point SASB22, which resulted in the channel B auxiliary feedwater actuation signal steam generator 2 drifting out of tolerance for a period of 13 days. This exceeded the allowed outage time specified in the technical specifications. The inspectors performed the initial significance determination using NRC Inspection Manual 0609, Appendix A, Exhibit 2, "Mitigating Systems Screening Questions." The finding screened to a detailed risk evaluation because it involved the actual loss of function of at least a single train for greater than its technical specification allowed outage time. A Region IV senior reactor analyst performed a detailed risk evaluation and determined that the change in core damage frequency CDF < 5E -9 corresponds to very low (Green) safety significance. This finding has a cross-cutting aspect in the area of human performance associated with the change management component in that the licensee did not use a systematic process for evaluating and implementing change so that nuclear safety remains the overriding priority. Specifically, the licensee did not use a systematic process to identify and correct the lack of alarm procedures associated with this parameter along with 76 other alarms that have technical specification implications during the design modification process for the plant computer alarm system (H.3).
05000458/FIN-2014004-05	2014Q3	River Bend	Licensee-Identified Violation	Title 10 CFR 20.1501(a) requires that each licensee make, or cause to be made, surveys that may be necessary for the licensee to comply with the regulations in 10 CFR Part 20 and that are reasonable under the circumstances to evaluate the extent of radiation levels, concentrations or quantities of radioactive materials, and the potential radiological hazards that could be present. Pursuant to 10 CFR 20.1003, a survey means an evaluation of the radiological conditions and potential hazards incident to the production, use, transfer, release, disposal, or presence of radioactive material or other sources of radiation. Title 10 CFR 20.1201(c) states, in part, the assigned deep-dose equivalent must be for the part of the body receiving the highest exposure. Contrary to this requirement, the licensee did not make or cause to be made surveys that were necessary for the licensee to comply with the regulations of 10 CFR 20.1201(c). Specifically, licensee representatives did not perform surveys to evaluate the radiation dose gradient in the reactor cavity, caused by placement of the reactor pressure vessel head, during work on March 15 and 16, 2013. The failure to provide dose gradient surveys was identified by the outage control center radiation protection representative while reviewing radiation survey records. Licensee personnel documented the failure to survey for radiation dose gradients in Condition Report CR-RBS-2013-02426 and performed an apparent cause evaluation. During follow-up actions, licensee personnel identified an example in which a worker received 104 millirem of unplanned radiation dose and reported it as an occupational exposure control effectiveness performance indicator occurrence. Using Inspection Manual Chapter 0609, Appendix C, "Occupational Radiation Safety Significance Determination Process," the inspectors determined the violation had very low safety significance because: (1) it was not an as low as is reasonably achievable (ALARA) finding, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised.

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