Semantic search

This page or action requires JavaScript to work. Please enable JavaScript in your browser or use a browser where it is supported, so that functionality can be provided as requested. For further assistance, please have a look at the noscript help page.

Search Clear all entries

	Title	Quarter	Description
05000275/FIN-2018404-02	Licensee-Identified Violation	2018Q3
05000275/FIN-2018404-03	Licensee-Identified Violation	2018Q3
05000275/FIN-2018003-01	Multiple Examples of Scaffolding in Place Greater Than 90 Days Without Required Evaluation	2018Q3	The inspectors identified a Green, non-cited violation of 10 CFR 50, Appendix B, Criterion V, Procedures, because PG&E personnel failed to follow the requirements of AD7.ID5, Scaffold Material Structure. Specifically, 20 instances of scaffold structures installed in the plant were identified that had been in place for greater than 90 days without required 10 CFR 50.59 reviews being completed.
05000275/FIN-2018003-02	4 kV Vital Switchgear Room Ventilation Degraded or Non-Conforming Condition and Associated Compensatory Measure Not Corrected in a Timely Manner	2018Q3	The inspectors identified a Green, non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, because PG&E personnel failed to promptly correct a degraded or non-conforming condition associated with an open operability condition. Specifically, PG&E personnel did not promptly correct a degraded condition associated with an open operability determination and corresponding compensatory measure related to Unit 1 and Unit 2, 4 kV vital switchgear ventilation for a period of over 4 years. This time period included two refueling outages for Unit 1 and three refueling outages for Unit 2.
05000275/FIN-2018404-01	Security	2018Q3	SECURTIY
05000275/FIN-2018008-05	Minor Violation	2018Q2	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.
05000275/FIN-2018008-06	Minor Violation	2018Q2	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.
05000275/FIN-2018010-01	Failure to install conduit to environmentally qualified solenoid valves in a manner to prevent moisture intrusion and accumulation within the solenoid enclosure	2018Q2	The inspectors identified a Green finding and associated Non-cited Violation (NCV) of 10 CFR Part 50.49, Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants, for failure to install conduit to environmentally qualified solenoid valves in a manner to prevent moisture intrusion and accumulation within the solenoid enclosure in accordance with Environmental Qualificationfile IH06, Revision 25.
05000275/FIN-2018008-04	Licensee-Identified Violation	2018Q2	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
05000275/FIN-2018008-01	Emergency Diesel Generator Mission Time for Operability Evaluations	2018Q2	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	Failure to Promptly Identify and Correct Emergency Diesel Generator 1-1 Cardox System Inoperability	2018Q2	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	Failure to Identify Diesel Generator Air Inlet Boot Seal Critical Characteristics	2018Q2	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.
05000323/FIN-2018001-01	Improper Troubleshooting Results in Reactor Trip Signal and Loss of Source Range Nuclear Instrument Power	2018Q1	The inspectors reviewed a Green,self-revealed non-cited violation of Technical Specification 5.4.1.a Procedures, because PG&E personnel failed to follow the requirements of MA1.DC54, Conduct of Maintenance, Revision 15. Specifically, on March 20, 2018, with the reactor in Mode 3 during informal troubleshooting of high background count rate on source range nuclear instrument (NI) NI-32, PG&E personnel caused a short in NI cabinet B resulting in a blown fuse and the loss of power to the cabinet. This resulted in the loss of power to power range NI-42, intermediate range NI-36, source range NI-32, a reactor trip signal, a turbine trip signal, and all associated reactor protection interlocks. Power was automatically removed from the remaining source range NI due to reactor protection interlock P-10, resulting in no safety-related source range NI indication being available for control room operators.
05000323/FIN-2018001-02	Failure to Follow Operating Experience Procedures Results in Inadequate Screen of Operating Experience Report	2018Q1	The inspectors identified a finding of very low safety significance (Green) because PG&E personnel failed to follow the requirements of OM4.ID3, Operating Experience Program, Revision 20. Specifically, PG&E personnel failed to screen relevant operating experience relating to a safety-related centrifugal charging pump (CCP) journal bearing failure due to non-metallic anti-rotation pin shear failure. This operating experience notice was received by PG&E September 2011 and was not screened per OM4.ID3, Operating Experience Program, preventing actions from being identified and implemented that could have eliminated vulnerabilities and prevented a similar event from occurring at DCPP. On November11,2017, CCP 2-1 was declared inoperable and determined to be non-functional due to a damaged journal bearing caused by non-metallic, anti-rotation pin shear failur
05000323/FIN-2017003-01	Inadequate Corrective Actions resulted in a Failure to Comply with Technical Specification 3.4.11 and an Emergency Declaration	2017Q3	The inspectors identified a non- cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the failure of the licensee to identify and correct a condition adverse to quality. Specifically, the licensee failed to implement prompt corrective actions related to a nitrogen leak from a component associated with safety -related pressurizer power -operated relief valve (PORV) , PCV -455C . The nitrogen leak subsequently resulted in the PORV being declared inoperable, as well as the declaration of an Alert emergency action level classification due the Unit 2 containment atmosphere exceeding habitability limits. The licensees failure to implement prompt corrective action to correct excessive nitrogen leakage into the Unit 2 containment was a performance deficiency. The finding was more than minor because it was associated with the equipment performance attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the excessive nitrogen leakage resulted in the inoperability of safety -related PORV PCV -455C for greater than technical specification allowed outage time and atmospheric conditions in Unit 2 containment that were an immediate danger to life and health, prompting an Alert emergency declaration. Using NRC Manual Chapter 0609, Significance Determination Process, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, and Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors 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. The finding was assigned a human performance cross -cutting aspect associated with consistent processes, in that the licensee did not use a systematic approach in properly assessing the potential risk significance of an increasing trend of nitrogen leakage inside containment (H.13) .
05000275/FIN-2017002-03	Failure to Report a Permanent Medical Condition Within 30 Days	2017Q2	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-2017002-02	Failure to Conduct Required Biennial Medical Examinations Within Two Years	2017Q2	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-2017002-01	Inadequate Expansion Scope of Risk - Informed Welds	2017Q2	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	Failure to Follow Procedures Results in Partial Loss of Cooling Flow to Shutdown Cooling	2017Q2	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 ).
05000275/FIN-2017001-01	Licensee-Identified Violation	2017Q1	Technical Specification 3.3.3 Post Accident Monitoring (PAM) Instrumentation, requires at least two channels of both wide range) hot leg reactor coolant system (RCS) temperature and wide range cold leg RCS temperature RTDs to be in service. If this action is not met, TS 3.3.3 requires the restoration of all but one channel to operable status within 7 days. If this action cannot be met, TS 3.3.3 requires the plant to be shutdown to Mode 3 within 6 hours and Mode 4 within 12 hours. Contrary to the above, in October 2015 during performance of an apparent cause evaluation investigating failing wide range RCS RTDs, PG&E discovered that the plant had been operating with all channels of hot leg and cold leg wide range RCS temperature monitoring inoperable for greater than the allowed TS 3.3.3 outage time without complying with the requirement to shut down the plant. Pacific Gas and Electric identified an incorrect insulation configuration, installed in 2010, on the thermal extension piping that houses the wires for the wide range RCS RTDs as the direct cause of the failures. The insulation, as installed, trapped heat inside of the thermal extension piping and overheated the associated wires. Pacific Gas and Electric determined that eight wide range RCS RTDs had ether failed or operated outside of the environmental qualification temperature range, however the required channels remained functional. Pacific Gas and Electric determined the cause of the incorrect installation to be insufficient guidance in the associated work package instructions. The inspectors determined that PG&Es failure to develop adequate work guidance to properly install wide range RCS RTD insulation was a performance deficiency that was within PG&Es ability to foresee and correct. Pacific Gas and Electric entered this issue into their corrective action program (CAP) as Notification 50808493, replaced the eight wide range RTDs, restored the insulation per design requirements, revised the drawings for Unit 1 wide range RTDs to provide adequate level of detail, and revised the work order to include the correct drawing and level of details for proper installation of all wide range RTDs. This performance deficiency is considered more than minor, and considered a finding, because it is associated with the Mitigating Systems cornerstone attribute of equipment performance and affected the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated the finding in accordance with IMC 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations. The inspectors determined that the finding was of very low safety significance (Green) because the deficiency did not represent an actual loss of function of at least a single train for greater than its technical specification allowed outage time.
05000275/FIN-2016008-01	Licensee-Identified Violation	2016Q4	License Conditions 2.C.(5) for Unit 1 and 2.C.(4) for Unit 2 state, in part, that the licensee shall implement and maintain all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the license amendment request, dated June 26, 2013, and as approved in the safety evaluation, dated April 14, 2016. Section 3.2.3 of Attachment A of the license amendment request states, in part: Procedures shall be established for implementation of the fire protection program. In addition to procedures that could be required by other sections of the standard, the procedures to accomplish the following shall be established: Inspection, testing, and maintenance for fire protection systems and features credited by the fire protection program. Equipment Control Guideline 18.1, Fire Suppression Systems/Fire Suppression Water Systems, provided the requirements to demonstrate the operability of the fire suppression water system. Equipment Control Guideline Surveillance Requirement 18.1.10 required the licensee to perform a flow test of the fire suppression water system in accordance with Chapter 5, Section 11, of the Fire Protection Handbook, 14th Edition. Chapter 5, Section 11, of the Fire Protection Handbook, 14th Edition, stated, in part, that tests should be conducted in such a way that the available flow and pressure at high value or hazardous areas can be determined readily. Contrary to the above, prior to October 20, 2016, the licensee failed to implement all provisions of the approved fire protection program. Specifically, during the flow tests, the licensee failed to establish surveillance test procedures that measured pressure values for one of the three fire suppression water subsystems using the methodology in Chapter 5, Section 11 of the Fire Protection Handbook, 14th Edition. The performance deficiency was more than minor because it was associated with the protection against external factors (fire) attribute of the Mitigating Systems Cornerstone and adversely affected the Mitigating Systems Cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was screened in accordance with Inspection Manual Chapter 0609, Appendix F, dated September 20, 2013. The finding was determined to be of very low safety significance (Green) in Task 1.4.7, Fire Water Supply, Question A. Although the licensee did not measure and record fire main pressure, they did measure flow and it was greater than 50 percent of the required capacity. The finding was entered into the licensees corrective action program as Notifications NN 50853684 and NN 50863322.
05000323/FIN-2016004-01	Failure to Follow Maintenance Procedure Resulted in Improper Configuration of Safety Related Equipment	2016Q4	Green. The inspectors identified a non-cited violation of Technical Specification 5.4.1.a,Procedures, for the failure to follow Procedure AD7.ID16, Tool Pouch and Minor Maintenance Program, Revision 2. Specifically, the licensee failed to screen work on the safety-related rupture restraint as acceptable to be worked as tool pouch work or minor maintenance. As a result, a safety-related main steam line rupture restraint (MS-41RR) was not properly returned to service and left in an inoperable condition following maintenance. As corrective actions, the licensee returned MS-41RR to an operable condition and initiated a review of the maintenance database to ensure that work performed on main steam line rupture restraints is completed in accordance with appropriate written inspections. The licensee entered the issue into their corrective action program as Notifications 50872133,50872056, and 50872789. The failure to properly preplan and perform maintenance affecting the performance of safety-related equipment was a performance deficiency. The inspectors determined that the finding was more than minor because it was associated with the configuration control attribute of the Mitigating System Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesired consequences. Specifically, because of not following maintenance procedures, a safety-related main steam rupture restraint was left in a disengaged or inactive configuration such that following a postulated line break, the main steam line would be unrestrained. This resulted in a potential of high-energy pipe impacting structures and components designed to be protected from high-energy pipe whip. Using IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined the finding was of very low safety significance (Green) because the finding did not represent an actual loss of function of a mitigating system. Specifically, the single restraint condition would only affect a very limited range of breaks and no risk significant systems would be adversely impacted. The inspectors concluded that this finding affected the cross cutting area of human performance, documentation, because the licensee did not maintain up to date documentation to ensure work planning on safety related equipment are complete, thorough, accurate, and current such that main steam pipe restraints are maintained within design requirements (H.7).
05000275/FIN-2016403-01	Security	2016Q4
05000275/FIN-2016302-01	Licensee-Identified Violation	2016Q4	A severity level IV violation that was identified by the licensee has been reviewed by the examiners. Corrective actions taken or planned by the licensee have been entered into the licensees corrective action program. The violation and corrective action tracking number are listed in Section 4OA7 of this report. Title 10 CFR 50.9(a) requires, in part, that information provided to the Commission by a licensee shall be complete and accurate in all material aspects. On October 4, 2016, the NRC gave approval to the licensee to administer a written examination to initial operating license applicants on October 14, 2016. The approval was made based on content of the written examination provided to the NRC on October 4, 2016. In this version of the written examination, Question 55 had been revised based on NRC comments so that it had only one correct answer. The previous draft revision of the question had two plausible correct answers. The written examination was administered on October 14, 2016. During licensee review of the exam, the licensee identified that the version of Question 55 on the administered written examination wa not the version that was approved on October 4, 2016. The licensee notified the NRC of the issue on November 7, 2016, and completed an extent of condition review that showed that this was the only written examination question inconsistent with the questions approved on October 4, 2016. The violation was of very low safety significance because the performance deficiency did not contribute to the NRC making any incorrect regulatory decisions regarding issuance of operating licenses.
05000323/FIN-2016010-01	Failure to Establish Adequate Work Instructions for Installation of Namco Snap Lock Limit Switches	2016Q3	The inspectors identified a preliminary White finding associated with an apparent violation of Technical Specification 5.4.1.a, Procedures, for the licensees failure to develop adequate instructions for the installation, adjustment, and testing of NamcoTM Model EA170 snap lock limit switches. Specifically, the licensee failed to provide site-specific instructions for limiting the travel of these external limit switches when installed on safety-related motor operated valves. Consequently, the lever switch actuator for valve RHR-2-8700B, residual heat removal pump 2-2 suction from the refueling water storage tank, was installed such that the limit switch was operated repeatedly in an over-travel condition resulting in a sheared internal roll pin that ultimately caused the limit switch to fail. Following identification of this issue, the licensee replaced the limit switch for valve RHR-2-8700B and implemented actions to modify maintenance procedures for installing, calibrating, and testing motor-operated valve external limit switches. The licensee entered this issue into their corrective action program as Notification 50852345. The performance deficiency is more than minor, and therefore a finding, because it is associated with the procedure quality attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, maintenance procedure MP E-53.10R, Augmented Stem Lubrication for Limitorque Operated Valves, used to perform limit switch adjustments on the Unit 2 valve RHR-2-8700B, did not provide adequate acceptance criteria to prevent overtravel of the limit switch actuating lever. This resulted in a subsequent failure of the limit switch, preventing the open permissive signal for valve SI-2-8982B, residual heat removal pump 2-2 suction from the containment recirculation sump, used during the emergency core cooling system (ECCS) recirculation mode. The inspectors evaluated the finding using the Attachment 0609.04, "Initial Characterization of Findings," worksheet to Inspection Manual Chapter (IMC) 0609, Significance Determination Process, issued June 19, 2012. The attachment instructs the inspectors to utilize IMC 0609, Appendix A, Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that the finding required a detailed risk evaluation because it represented an actual loss of function of the train B ECCS for greater than its technical specification allowed outage time. A senior reactor analyst performed a detailed risk evaluation in accordance with IMC 0609, Appendix A, Section 6.0, Detailed Risk Evaluation. The calculated increase in core damage frequency was dominated by small and medium loss of coolant accident initiators with failures of the opposite train of ECCS or related support systems. The analyst did not evaluate the large early release frequency because this performance deficiency would not have challenged the containment. The NRC preliminarily determined that the increase in core damage frequency for internal and external initiators was 7.6E-06/year, a finding of low to moderate risk significance (White). The inspector did not identify a cross-cutting aspect with this finding because it was not reflective of current performance. The inadequate procedure was developed in 2011 and did not reflect the licensees current performance related to procedure development.
05000323/FIN-2016002-01	Misplaced Spent Fuel Assembly in the Spent Fuel Pool	2016Q2	The inspectors reviewed a self-revealed, non-cited violation of Technical Specification (TS) 5.4.1.a, Procedures, for the licensees failure to place a spent fuel assembly in its correct location in the spent fuel pool (SFP) in accordance with Procedure OP B-8H, Spent Fuel Pool Work Instructions. Specifically, the fuel handling crew moved spent fuel assembly TT69 to location E-37 rather than its intended location E-27. In response to this error, reactor engineering performed a technical specification verification in order to ensure that fuel assembly TT69 could remain in Cell E-37. The licensee suspended further fuel movements pending corrective action and remediation of the operators. The licensee entered this into the corrective action program as Notifications 50846834 and 50847067. The licensees failure to place a spent fuel assembly in its correct location in the SFP was a performance deficiency. The 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 (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Using Inspection Manual Chapter 0609, Attachment 04, Initial Characterization of Findings, and Appendix A, Exhibit 3, Barrier Integrity Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because: (1) the finding did not adversely affect decay heat removal capabilities from the spent fuel pool causing the pool temperature to exceed the maximum analyzed temperature limit specified in the site-specific licensing basis, (2) the finding did not result from fuel handling errors, dropped fuel assembly, dropped storage cask, or crane operations over the SFP that caused mechanical damage to fuel clad and a detectible release of radionuclides, (3) the finding did not result in a loss of spent fuel pool water inventory decreasing below the minimum analyzed level limit specified in the site-specific licensing basis, and (4) the finding did not affect the SFP neutron absorber, fuel bundle misplacement (i.e., fuel loading pattern error) or soluble Boron concentration. This finding had a cross-cutting aspect in the area of human performance associated with avoiding complacency. Specifically, individuals failed to recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes and individuals failed to implement appropriate error reduction tools (Section 4OA2). (H.12)
05000275/FIN-2016301-01	Insufficient Procedural Direction Contained Within Procedure EOP E-2, Faulted Steam Generator Isolation	2016Q2	The examiners identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings. Specifically, Procedure EOP E-2, Faulted Steam Generator Isolation, does not contain sufficient procedural direction for isolating auxiliary feedwater flow to a faulted steam generator in the event that auxiliary feedwater control valves cannot be closed from the control room. Procedure EOP E-2, Appendix HH, Isolated Faulted Steam Generator, Step 1.d, and its associated column, Response Not Obtained, does not ensure that a faulted steam generator would remain isolated under all conditions. The Response Not Obtained column permits operators to either locally close auxiliary feedwater control valves OR secure the auxiliary feedwater pump feeding the faulted steam generator. However, due to the absence of pull-to-lock or hard stop switches for the auxiliary feedwater pumps, the possibility exists for an automatic restart of an auxiliary feedwater pump and a re-initiation of feedwater to a faulted steam generator. The failure to ensure that Procedure EOP E-2 contained sufficient direction to isolate a faulted steam generator when auxiliary feedwater flow control valves cannot be closed from the control room was a performance deficiency. This performance deficiency was of more than minor safety significance because it was associated with the procedure quality attribute of the Barrier Integrity cornerstone (reactor coolant system and containment) and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the re-initiation of feedwater to an isolated, faulted steam generator has the potential to adversely affect the reactor coolant system barrier by causing an additional unintended cooldown of the reactor coolant system, increased potential for pressurized thermal shock, and thermal stress to the steam generator u-tubes. Additionally, the containment barrier would be affected by the reinitiation of feedwater to a steam line break within containment. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, the team determined that the finding required a detailed risk evaluation due to the potential to affect the reactor coolant system boundary. A senior reactor analyst performed a bounding detailed risk evaluation and estimated the maximum increase in core damage frequency to be 5.9E-8/year, and therefore the finding was determined to be of very low safety significance (Green). This increase in core damage frequency was mitigated by the low probability of multiple equipment failures in the auxiliary feedwater system when combined with the low initiating event frequency of a faulted steam generator. Because the violation was of very low safety significance (Green) and the issue was entered into the licensees corrective action program as Notification 50847218, this violation is being treated as a non-cited violation consistent with Section 2.3.2.a of the Enforcement Policy: NCV 05000275/2016301; 05000323/2016301-01, Insufficient Procedural Direction Contained Within E-2, Faulted Steam Generator Isolation. This finding has a crosscutting aspect in the area of human performance associated with resources because the organization did not ensure procedures are available and adequate to support nuclear safety (H.1).
05000275/FIN-2016007-03	Failure to Ensure Safety-Related Alternating Current and Direct Current Equipment Functionality at Maximum Allowable Voltages	2016Q1	The team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, The design control measures shall provide for verifying or checking the adequacy of design, such as by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Specifically, prior to February 10, 2016, the licensee failed to verify the design of (1) equipment on the nominally 125 Vdc system at the maximum voltage specified in Procedure OP J-9:IV, Performing a Battery Equalizing Charge, and (2) equipment on 480 Vac and 120 Vac vital buses at maximum voltages specified in Procedure OP J-2:VIII, Guidelines for Reliable Transmission Service for DCPP, by the use of alternate or simplified calculational methods, to ensure equipment functionality. In response to this finding, the licensee conducted a preliminary evaluation of the affected equipment and concluded that any past exposure to voltages above their maximum rating would not have caused a loss of functionality. This finding was entered into the licensee's corrective action program as Notifications 50834558, 50835906, 50835394, 50835945, 50835949, 50836376, 50836439, 50836638, 50836872, and 50836995. The team determined the failure to evaluate operation of 125 Vdc and 480 and 120 Vac equipment at maximum allowable voltages was a performance deficiency. The performance deficiency was more-than-minor, and therefore a finding, because it related to the equipment performance attribute of the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, operation of equipment outside of its rated or analyzed maximum allowable voltages adversely affects the reliability and capability of that equipment required to perform safety-related functions. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated July 19, 2012, the finding 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 nontechnical 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 human performance associated with design margins because the licensee failed to ensure that the organization operated and maintained equipment within design margins and that margins were carefully guarded and changed only through a systematic and rigorous process (H.6).
05000275/FIN-2016007-04	Failure to Evaluate the Extent of Condition for a Degraded Condition on a Nonsafety-Related 4160 Vac Breaker	2016Q1	The team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, which states, in part, Activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Specifically, in October of 2015, the licensee failed to evaluate the extent of condition of a cracked holding pawl on a nonsafety-related 4160 Vac SF6 breaker, which was procured as safety-related, in accordance with Procedure OM7.ID1, Problem Identification and Resolution, when the failure of the component could adversely impact safety-related breakers of the same make and model. In response to this finding, the licensee is performing a procedure review to include steps to perform an extent of condition analysis for unplanned nonsafety-related equipment issues that may also affect similar safety-related equipment. This finding was entered into the licensee's corrective action program as Notifications 50836859 and 50836689. The team determined the failure to evaluate the impact of a cracked holding pawl identified on a nonsafety-related 4160 Vac SF6 breaker on additional safety-related 4160 Vac SF6 breakers was a performance deficiency. The performance deficiency was more-thanminor, and therefore a finding, because it related to the equipment performance attribute of the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the 4160 Vac breaker with the cracked holding pawl was procured as safetyrelated; therefore, the condition extends to safety-related 4160 Vac breakers of the same make and model and potentially adversely affects the ability to perform their safety function. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated July 19, 2012, the finding 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 human performance associated with conservative bias because the licensee failed to ensure that individuals used decision-making practices that emphasized prudent choices (H.14).
05000275/FIN-2016001-01	Failure to Verify Adequate Design Airflow for 480 volt AC Switchgear and 125 volt DC Inverter Rooms	2016Q1	The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to verify the design adequacy of the safety-related ventilation system for the 480-volt AC switchgear and 125-volt DC inverter rooms. Specifically, the licensee failed to verify sufficient ventilation system airflow to ensure the temperature in rooms housing safety-related electrical equipment remained below 104 degrees Fahrenheit. The licensees corrective actions were documented in Notification 50840266. The failure to provide design control measures to verify the adequacy of the 480-volt AC switchgear and 125-volt DC inverter rooms ventilation system design was a performance deficiency. The performance deficiency was more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the reduction in airflow to the rooms impacts the reliability of the safety-related equipment ventilation system to maintain the temperatures in these rooms below design limits for the duration of all accident scenarios. Using NRC Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined the finding was of very low safety significance because (1) the finding was not a deficiency affecting the design or qualification of a mitigating system; (2) the finding did not represent a loss of system and/or function; (3) the finding did not represent an actual loss of function of a single train for greater than its technical specification allowed outage time; and (4) the finding does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. The inspectors determined that this finding did not have a cross-cutting aspect because the most significant contributor of this finding occurred more than three years ago, and is therefore, not representative of current licensee performance.
05000275/FIN-2016007-02	Failure to Promptly Correct the Lack of Design Verification of 460 Vac Motors at Maximum Allowable Frequency	2016Q1	The team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, which states, in part, Measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. Specifically, prior to March 16, 2016, the licensee failed to assure that the lack of design verification of 460 Vac motors, which could be overloaded at the maximum allowable diesel generator frequency, was promptly corrected after having been identified in a 2013 apparent cause evaluation and again in a 2015 self-assessment as documented in Notifications 50572850 and 50826105, respectively. In response to this finding, the licensee performed a preliminary evaluation of the affected 460 Vac motors and concluded that operation at maximum emergency diesel generator frequency would not cause them to overheat or trip on overcurrent. This finding was entered into the licensee's corrective action program as Notifications 50835699 and 50838988. The team determined the failure to correct the lack of design verification of 460 Vac motors at maximum allowable frequency when powered from the emergency diesel generators was a performance deficiency. The performance deficiency was more-than-minor, and therefore a finding, because it related to the design control attribute of the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, operation of 460 Vac motors above their rated or analyzed maximum allowable frequencies could result in motor overheating or a trip of the thermal overload relays. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated July 19, 2012, the finding 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 nontechnical 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 evaluation because the licensee failed to ensure that the organization thoroughly evaluated issues to ensure that resolutions address causes and extent of conditions (P.2).
05000275/FIN-2016007-05	Failure to Evaluate the Voltage Effects of Limiting Design Basis Events on the 230 kV Offsite Power Circuit	2016Q1	The team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, The 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 January 30, 2014, the licensee failed to verify the design of the 230 kV preferred offsite power source, such as by the performance of design reviews or use of alternate or simplified calculational methods, by assuming in calculation 359-DC, Determination of 230 kV Grid Capability Limits as DCPP Offsite Power Source, that the reactor trip and engineered safety features actuation system signals are coincident in time for all postulated design basis events. However, the plant is designed such that, during some events, the signals are separate in time and would result in a greater vital bus voltage depression than analyzed. In response to this finding, the licensee conducted a preliminary evaluation and concluded that the current transmission grid conditions were such that the calculation criteria would be met in the event of a design basis event involving non-coincident reactor trip and engineered safety features actuation system signals. This finding was entered into the licensee's corrective action program as Notification 50839137. The team determined the failure to evaluate the voltage effects of a limiting design basis event with non-coincident reactor trip and engineered safety features actuation system signals on the 230 kV offsite power circuit was a performance deficiency. The performance deficiency was more-than-minor, and therefore a finding, because it related to the design control attribute of the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to ensure adequate bus voltages as a result of a design basis event with non-coincident reactor trip and engineered safety features actuation system signals would result in a trip of the undervoltage relays and the loss of the preferred offsite power circuit. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated July 19, 2012, the finding 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 human performance associated with design margins because the licensee failed to ensure that the organization operated and maintained equipment within design margins and that margins were carefully guarded and changed only through a systematic and rigorous process (H.6).
05000275/FIN-2016007-01	Failure to Evaluate 480 Vac Motor Starters with Circuit Breaker Trip Settings Higher than Manufacturers Specifications	2016Q1	The team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, The design control measures shall provide for verifying or checking the adequacy of design, such as by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Specifically, prior to September 10, 2013, the licensee failed to verify the design of 480 Vac combination motor starter instantaneous magnetic circuit breakers settings, by the use of alternate or simplified calculational methods, for those breakers whose settings are higher than their manufacturers' specifications, as documented in calculation 195B-DC, MCCB Settings for 460VAC Class 1E Motors, to provide the required level of protection and ensure that certain failures that could be caused by sustained fault currents below the circuit breaker trip setting would not occur. In response to this finding, the licensee conducted a preliminary evaluation of some of the affected equipment and concluded that sustained fault currents below the trip settings are unlikely. This finding was entered into the licensee's corrective action program as Notification 50838071. The team determined the failure to evaluate 480 Vac combination motor starters with instantaneous magnetic circuit breaker trip current settings higher than their manufacturers' specifications was a performance deficiency. The performance deficiency was more-thanminor, and therefore a finding, because it related to the design control attribute of the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, improper motor starter breaker trip settings could result in a fire in the motor control center cubicle, damage to motor starter components, spurious tripping of the entire motor control center, or lack of protection for downstream components during fault conditions. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated July 19, 2012, the finding 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 most significant causal factor of the performance deficiency did not reflect current licensee performance.
05000275/FIN-2016007-06	Failure to Translate Appropriate Load Tap Changer Timing Acceptance Criteria into Periodic Tests	2016Q1	The team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, which states, in part, Instructions, procedures, or drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished. Specifically, prior to November 25, 2015, the licensee failed to include appropriate quantitative acceptance criteria in Procedure MP E-62.3, Tap Changer Functional Test for Standby-Startup Transformer 11, to ensure that the load tap changer speed for standbystartup transformer 11 was adequate to restore vital bus voltages to the required level during design basis events. In response to this finding, the licensee performed a preliminary evaluation of the condition and concluded that the most recently measured speed of the load tap changer was adequate to ensure that it would restore vital bus voltage within the required time. This finding was entered into the licensee's corrective action program as Notification 50839333. The team determined the failure to translate appropriate load tap changer timing acceptance criteria into functional tests to ensure that design assumptions were being maintained was a performance deficiency. The performance deficiency was more-than-minor, and therefore a finding, because it related to the design control attribute of the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the load tap changer could meet its functional test acceptance criterion, but not operate fast enough to restore vital bus voltages within the required time during design basis events, which would result in an undervoltage trip and loss of the preferred offsite power circuit. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated July 19, 2012, the finding 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 nontechnical 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 human performance associated with design margins because the licensee failed to ensure that the organization operated and maintained equipment within design margins and that margins were carefully guarded and changed only through a systematic and rigorous process (H.6).
05000275/FIN-2015004-01	Failure to Properly Evaluate for Aggregate Impact of Fire Impairments	2015Q4	The inspectors identified a non-cited violation of Technical Specification 5.4.1.d, Procedures, for the failure to follow approved fire protection program procedures to review the fire impairments list to assess the aggregate impact on the fire protection design and safe shutdown analysis. Specifically, from August 31 to September 2, 2015, the licensee failed to evaluate the aggregate impact of having three fire doors simultaneously blocked open in adjacent Unit 1 vital battery charger rooms. The licensee implemented immediate corrective actions by assigning a continuous fire watch to the area and documented the issue in the corrective action program as Notification 50826793. The failure to follow approved fire protection program procedures to review the fire impairments list to assess the aggregate impact on the fire protection design and safe shutdown analysis was a performance deficiency. This performance deficiency was more than minor, and therefore a finding, because it was associated with the Initiating Events cornerstone attribute of Protection against External Factors (Fire) and adversely affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during plant operations. Specifically, the failure to evaluate the aggregate impact of multiple fire system impairments affected the licensee ability to limit the impact of a potential fire. The inspectors evaluated the finding using IMC 0609, Attachment 4, Phase 1Initial Screening and Characterization of Findings. Because the finding involved fire protection, the inspectors transitioned to IMC 0609, Appendix F Fire Protection Significance Determination Process. The inspectors characterized the finding using IMC 0609, Appendix F, Attachment 1, "Fire Protection SDP Phase 1 Worksheet," dated September 20, 2013. The finding screened as very low safety significance (Green), per Attachment 1, Question 1.4.3-A since the fire finding category was determined to be fire confinement, due to the fire doors being propped open, and the combustion loading on both sides of the door was determined to be a duration of 30 minutes as documented in licensee calculation M-824, Controlled Combustion Loading Tracking. In addition, the inspectors determined this finding had a cross-cutting aspect in human performance associated with the teamwork component because the licensees work groups did not properly communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety was maintained. Specifically, the work planners did not properly communicate to the fire protection department that all three fire doors would be open at the same time during battery charger load testing. (H.4)
05000275/FIN-2015004-02	Failure to Identify a Cause and Implement Actions to Prevent Recurrence of a Significant Condition Adverse to Quality	2015Q4	The inspectors reviewed a self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI Corrective Action, for the failure to identify the cause and take corrective action to prevent recurrence of a significant condition adverse to quality impacting both trains of the Unit 1 safety-related residual heat removal (RHR) system. Specifically, the licensee failed to identify a definitive cause and implement corrective actions to prevent recurrent failures of the socket weld for relief valve RHR-1-RV-8708 for both trains of the RHR system. As immediate corrective actions, the licensee installed additional piping supports to mitigate the vibrations at the socket weld and documented this issue in the corrective action program as Notification 50680750. The failure to identify the cause of the RHR vibration-induced problems and to take adequate corrective actions to prevent recurrence of the weld failures was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because if left uncorrected, it could lead to a more significant safety concern. Specifically, no additional supports were installed and no actions were taken to reduce or eliminate the vibrations to prevent recurring weld failures, which could affect the availability of the RHR system. The lack of corrective actions to prevent recurrence could leave RHR components and other components physically connected to the system susceptible to future failures. Using Inspection Manual Chapter 0609, Appendix A, the inspectors determined the issue to have very low safety significance (Green) because the performance deficiency, which affected the mitigating systems cornerstone, did not result in a loss of safety function and did not result in an actual loss of function for greater than the technical specification allowed outage time. The licensee entered this into their corrective action program as Notification 50680750. In addition, this finding has a cross-cutting aspect in the human performance area associated with conservative bias decision making component because individuals failed to use decision making practices that emphasize prudent choices over those that are simply allowable. Specifically, the licensee chose to only install a fatigue resistance weld rather than install additional pipe supports as were in the Unit 2 system (H.14).
05000275/FIN-2015004-03	Failure to Design the Emergency Diesel Generators to operate under Worst Case Environmental Conditions	2015Q4	The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III Design Control, for the failure to implement design control measures to verify the adequacy of the Unit 1 emergency diesel generators (EDGs) cooling system design to ensure operation of the EDGs under worst-case environmental conditions. Specifically, since initial licensed operations began in 1984, the licensee failed to ensure the Unit 1 EDGs were designed and built to operate under worst-case high wind and temperature conditions. As a result, sustained high winds from specific directions could have impacted EDG radiator performance resulting in the unavailability of the Unit 1 EDGs. Immediate corrective actions included issuing shift orders to the reactor operators to monitor for specific weather conditions (high air temperature, high wind speed and direction) and provide additional room cooling using established procedures, as necessary. The licensee documented the issue in the corrective action program as Notification 50599190. The failure to implement design control measures to ensure the emergency diesel generators could perform their design basis function was a performance deficiency. The performance deficiency was more than minor, and is therefore a finding, because it was associated with the design control attribute of the mitigating system 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 performance deficiency resulted in a condition where sustained high winds from specific directions could have impacted EDG radiator performance resulting in the unavailability of the Unit 1 EDGs. The inspectors evaluated the finding using Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process for Findings At Power, dated June 19, 2012. The inspectors determined that a detailed risk evaluation by an NRC senior reactor analyst was required since the finding was associated with a loss of EDG function. The regional senior reactor analyst performed a Phase 3 SDP analysis for the finding. The results of analysis established the incremental conditional core damage probability (ICCDP) was 2.74E-07, less than 1 x 10-6, and therefore the analyst determined that the subject finding was of very low safety significance (Green). A cross-cutting aspect was not assigned to the finding since the finding did not represent current licensee performance. The condition existed since original construction of the plant.
05000275/FIN-2015003-01	Failure to Document an Adequate Evaluation for a Change in Seismic Load Combination Methodology	2015Q3	The inspectors identified a non-cited violation of 10 CFR 50.59(d)(1) which requires, in part, that the licensee shall maintain records of changes in the facility, of changes in procedures, and of tests and experiments made pursuant to paragraph (c) of this section. These records must include a written evaluation which provides the bases for the determination that the change, test, or experiment does not require a license amendment pursuant to paragraph (c)(2). Specifically, the licensee changed the method for combining earthquake loads and loss of coolant accident loads from the absolute summation method to square root sum of the squares (SRSS) method without sufficient justification to demonstrate the change did not require prior NRC approval. 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. The licensee entered the issue into the corrective action program as Notification 50811191. In accordance with the licensees corrective action program, this issue will be addressed by the licensee through a re-evaluation of the methodology change and the required actions that need to be taken by the licensee will be implemented. Additionally, the licensee performed an operability determination for the affected structures, systems, and components that established a reasonable expectation for operability pending final resolution of the issue. This performance deficiency was more than minor, and therefore a finding, 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 determine that use of SRSS in the Watts Bar safety evaluation report cited in the PG&E evaluation represented a change in a method of evaluation, in that the Watts Bar safety evaluation report was very narrow in scope and not appropriate for the intended application at Diablo Canyon. 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 result in the inoperability of the system. Because this performance deficiency had the potential to impact the NRCs ability to perform its regulatory function, the inspectors also evaluated the performance deficiency using traditional enforcement. Since the violation is associated with a Green finding having very low safety 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. This finding had a cross cutting aspect in the area of human performance associated with design margins because individuals failed to ensure margins were carefully guarded and changed only through a systematic and rigorous process (H.6).
05000275/FIN-2015003-02	Failure to Secure a Locked High Radiation Area	2015Q3	The inspectors reviewed a self-revealing, non-cited violation of Technical Specification 5.4.1(a), Procedures, for failure to secure a locked high radiation area. Specifically, the padlock on the Letdown Filter 1-1 locking bar was found unlocked. Upon discovery, the licensee guarded the area until properly secured. This issue was entered into the licensees corrective action program as Notification 50710852. The failure to secure a locked high radiation area was a performance deficiency. The performance deficiency was more than minor because, if left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, failure to adequately secure the locked high radiation area could result in unintended exposure to high levels of radiation. Using Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008, the inspectors determined the violation was of very low safety significance (Green) because: (1) it was not an as low as 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. The finding had an avoid complacency cross-cutting aspect, in the area of human performance, because individuals failed to recognize and plan for the possibility of mistakes, even while expecting positive outcomes. Specifically, licensee personnel failed to ensure that the padlock was secured after completing the task (H.12).
05000275/FIN-2015404-01	Licensee-Identified Violation	2015Q3
05000275/FIN-2015003-03	Licensee-Identified Violation	2015Q3	PG&E Part 72 license SNM-2511, Condition #11 requires, in part, that The licensee shall operate the installation in accordance with the Technical Specifications in the Appendix. Appendix Technical Specification 2.1.2 requires in part that Preferential fuel loading shall be used during uniform loading. Contrary to the above, from July 18, 2009 through June 6, 2015, PG&E failed to load 19 casks in accordance with Appendix Technical Specification 2.1.2 for preferential fuel loading. Specifically, the licensee failed to load fuel assemblies with longest cooling times in the periphery of the basket. This violation was identified by PG&E and placed in their corrective action program. The licensee submitted Event Notification 51134 to the NRC on June 6, 2015 and later updated the Event Notification on June 9, 2015. Following the event notification, PG&E submitted a 30-day report to the NRC on July 6, 2015 (ML15187A239). This violation did not have any safety impact, in that all fuel assemblies met the requirements for burn-up, decay heat, and cooling time. All fuel and casks remain in a safe and analyzed condition. However, in order to re-establish compliance with PG&Es Part 72 license, the licensee must submit a license amendment request to the NRC. In accordance with the NRC Enforcement Policy Section 2.2 and IMC 0612 Section 03.23, Part 72, ISFSI inspection findings follow the traditional enforcement process and are not dispositioned through the Reactor Oversight Process or the Significance Determination Process. The violation screened as having very low safety significance, Severity Level IV, and is being treated as an NCV, consistent with Section 2.3.2 a. of the Enforcement Policy. The violation was determined to be more than minor since the violation requires DCPP to request a License Amendment from the NRC for their Part 72 license in order to restore compliance for the 19 affected casks. The violation was entered into the licensees corrective action program as Notifications 50706314 and 50706501. Following identification of the issue, the licensee performed an assessment that showed the casks would continue to perform their design function. Corrective actions for this issue included issuing the revised procedure, performing an extent of condition review, providing just-in-time training to Reactor Engineering staff involved, and added an independent third party review requirement for fuel contents loaded into the canister.
05000275/FIN-2015002-06	High Voltage Insulator Flashover Resulted in Loss of 230 kV Offsite Power and Start of Emergency Diesel Generators	2015Q2	The inspectors reviewed a self-revealing, Green finding for the licensees failure to adequately implement procedure OM7.ID1, Problem Identification and Resolution, to prevent a high voltage insulator flashover event in the 230 kV switchyard that occurred on October 31, 2014. Specifically, corrective actions from three previous root cause evaluations were not effective to prevent a loss of the 230 kV start-up power and subsequent auto start of all of the safety standby emergency diesel generators (EDGs). This issue was entered into the licensees corrective action program as Notification 50699230. The licensees failure to adequately implement procedure OM7.ID1, Problem Identification and Resolution was a performance deficiency. The performance deficiency was more than minor because it was associated with the human performance attribute of the Initiating Events cornerstone and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions. Specifically, this failure resulted in another high-voltage insulator flashover, which resulted in loss of 230 kV offsite startup power and activation of all safety-related EDGs, on October 31, 2014. In accordance with IMC 0609.04, Initial Characterization of Findings, the inspectors determined that the impact of the finding on Unit 1 should be evaluated using Exhibit 1 of IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, and further determined that this finding required a detailed risk evaluation by the regional senior risk analyst because the finding involved a partial loss of offsite power, a support system that contributes to the likelihood of an initiating event and affected mitigation equipment. The risk analyst determined that, with the 230 kV system de-energized, any plant transient would result in a plant-centered loss of offsite power. Therefore, the risk analyst calculated the incremental conditional core damage probability for an exposure period of 9 hours to be 2.09 x 10-7, which is lower than the 1 x 10-6 threshold in the significance determination process; this finding is of very low safety significance (Green) for Unit 1. In accordance with IMC 0609.04, Initial Characterization of Findings, the inspectors determined that the impact of the finding on Unit 2 should be evaluated using IMC 0609, Appendix G, Shutdown Operations Significance Determination Process, because the finding pertained to operations, an event, or a degraded condition while the plant was shut down. Unit 2 was shutdown in a refueling outage when the event occurred on October 31, 2014. Because of the shutdown configuration of Unit 2, the loss of 230 kV support system did not impact the ability to continue to provide decay heat removal for the unit. Therefore, the analyst determined qualitatively that this finding is also of very low safety significance (Green) for Unit 2. This finding has a cross-cutting aspect of work management, in the area of human performance, for failing to implement a process of planning, controlling, and executing work activities such that nuclear safety is an overriding priority. Specifically the licensee failed to effectively plan and coordinate preventative maintenance strategies associated with root causes from previous high-voltage insulators flashover or failures since 2008 to prevent the loss of offsite 230 kV and the transient on October 31, 2014 (H.5).
05000275/FIN-2015002-01	Failure to Appropriately Pre-plan and Perform Maintenance on Hydrogen Guard Piping	2015Q2	The inspectors identified a Green, non-cited violation of Technical Specification 5.4.1 involving the failure to appropriately pre-plan and implement written procedures associated with configuration control of the hazard barrier hydrogen guard piping in the proximity and impacting safety-related equipment. This issue was entered into the licensee corrective action program as Notification 50778755. The inspectors determined that the failure to consider the impact to the fire hazard analysis and the seismic configuration of the hydrogen guard pipe was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external events attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems (i.e. hazard barriers) that respond to initiating events, such as fires, to prevent undesirable consequence. Though there were no actual consequences, the breaching of the seismically qualified hydrogen guard piping removed a designed hazard barrier and has the potential to vent hydrogen into rooms containing safety related equipment. Using IMC 0609, Appendix F, Fire Protection Significance Determination Process, Phase 1 Worksheet, the finding was determined to be of very low safety significance (Green) because it represented a low degradation of fire prevention and administrative controls element of the plant combustible material controls program, and the breaching of the hydrogen guard piping would not have prevented the safe shutdown of the plant. This finding has a cross-cutting aspect of design margins associated with the human performance area. Specifically, the most significant contributor for the performance deficiency was the licensee did not have an adequate work process that focused on maintaining defense in depth related to a fire hazard barrier, such as a hydrogen guard piping, during maintenance activities. Breaching hydrogen guard piping impacts defense in depth and design margins used to protect safety-related equipment, and special attention is required to carefully guard and change the configuration with great thought and care (H.6).
05000275/FIN-2015403-01	Licensee-Identified Violation	2015Q2
05000275/FIN-2015002-08	Licensee-Identified Violation	2015Q2	Title 10 of the Code of Federal Regulations (10 CFR) 50.74(c) requires, in part, that licensees shall notify the appropriate Regional Administrator within 30 days of a permanent disability of a licensed operator as described in 10 CFR 55.25. Contrary to the above, from 2009 to March 4, 2013, the licensee failed to notify the appropriate Regional Administrator when a licensed operator was diagnosed with a permanent disability. The licensee documented this issue in DA 50540600. This violation was determined to impact the regulatory process and was evaluated using Section 2.2.2 of the NRC Enforcement Policy. In accordance with Section 6.4.d of the NRC Enforcement Policy, this violation was determined to be a Severity Level IV violation because of the failure to report a medical condition that would have required a license restriction to maintain medical qualifications.
05000275/FIN-2015002-04	Technical Specification 3.3.4 Not Met Due to Inoperable Remote Shutdown System Function	2015Q2	The inspectors reviewed a self-revealing Green, non-cited violation of Technical Specification 3.3.4 Remote Shutdown System, for the licensees failure to maintain adequate configuration control of fuses associated with an emergency diesel generator (EDG). The licensees failure to maintain adequate configuration control by not verifying that fuses were properly installed, and adequate post maintenance testing was performed, following maintenance activities was a performance deficiency. Specifically, following the 1R17 refueling outage, from approximately June 13, 2013 until November 22, 2013, EDG 1-3 would not have been able to perform its remote shutdown function due to not being able to be adequately operated at the local EDG control cubicle. The licensee entered this issue into the corrective action program as Notification 50595473, and took prompt actions to restore the fuses to the correct position and verify the positions of the fuses in the other EDG output breaker cubicles. The failure to properly install fuses in the local manual operation circuitry of EDG 1-3 was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external events (fire) attribute of the Mitigating Systems Cornerstone, and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, it affected the ability to reach and maintain safe shutdown conditions in case of a fire causing a control room abandonment. The inspectors evaluated this finding using Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process," dated September 20, 2013. Because it affected the ability to reach and maintain safe shutdown conditions in case of a fire that led to control room evacuation, the Phase 2 methodology of Inspection Manual Chapter 0609, Appendix F, was not appropriate for this finding. Therefore, the senior reactor analyst performed a Phase 3 evaluation to determine the risk significance. The analyst determined that the performance deficiency only increased the risk of the plant as it related to the need to locally control EDG 1-3 following a postulated control room evacuation. The Senior Risk Analyst determined that the change in core damage frequency was less than 1 x 10-6, and the finding was not significant with respect to large, early release frequency. The analyst determined that this finding was of very low risk significance (Green). This finding had a cross-cutting aspect in the area of human performance associated with the work practices component, because the licensee did not ensure supervisory and management oversight of work activities, such that nuclear safety was supported (H.5).
05000275/FIN-2015002-02	Failure to Maintain Operator Licensing Examination Integrity	2015Q2	The inspectors reviewed a self-revealing, Severity Level IV non-cited violation of 10 CFR 55.49, Integrity of Examinations and Tests, and an associated Green finding for the licensees failure to provide adequate examination security measures during administration of the 2015 biennial requalification examination. On May 26, 2015, a licensed operator was able to obtain plant computer information that led to the discovery of specific plant events contained on the NRC-required annual operating test. The licensee entered this issue into the corrective action program as Notification 50704195 and retested the crew with a new scenario. The failure of the licensee to provide adequate measures for examination security for the biennial requalification examinations was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it adversely affected the human performance attribute of the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using NRC Inspection Manual Chapter 0609, Significance Determination Process, Attachment 4, Tables 1 and 2 worksheets (issue date June 19, 2012); and the corresponding Appendix I, Licensed Operator Requalification Significance Determination Process (SDP), Flowchart Block #10 (issue date December 6, 2011), the finding was determined to have very low safety significance (Green). Although the 2015 finding resulted in a compromise of the integrity of biennial dynamic simulator examinations had no compensatory actions been taken, the equitable and consistent administration of the biennial dynamic simulator examination was not actually affected by this compromise. The traditional enforcement violation was determined to be a Severity Level IV violation consistent with Section 6.4.d of the Enforcement Policy. This finding has a cross-cutting aspect in the resources component of the human performance cross-cutting area because the licensee failed to ensure the procedures are adequate to ensure nuclear safety (H.1).
05000275/FIN-2015002-03	Inadequate Design Control for High-Energy Line Break Vent Flow Path	2015Q2	The inspectors identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, involving the licensees failure to ensure credited design features, such as flow vent paths, protect safety-related systems, from temperature and pressure effects of a high-energy line break (HELB) in the auxiliary building. Specifically, the licensee allowed obstruction of a credited flow path with acrylic glass plates not qualified in the original design and not verified to function under a HELB scenario. The licensee entered this issue into the corrective action program as Notifications 50697910 and 50698102, and took immediate actions to remove the acrylic glass plates from the vent path doors in the auxiliary building. The performance deficiency was determined to be more than minor because it affected the Mitigating Systems Cornerstone attribute of Design Control and adversely affected the cornerstone objective of ensuring the reliability, availability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee did not have adequate measures in place to ensure that qualified components were available to mitigate the consequences of a HELB in the auxiliary building. The finding screened as of very low safety significance (Green) because the finding did not affect the design or qualification of mitigating structures, systems, and components; the finding did not represent a loss of system and/or function; the finding did not represent an actual loss of a function of a single train for greater than the technical specification (TS) allowed outage time; the finding did not represent an actual loss of a function of one or more non-TS trains of equipment; and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The finding was not assigned a cross-cutting aspect since the performance deficiency is not indicative of current plant performance.
05000275/FIN-2015002-05	Failure to Appropriately Scope 230 KV Switchyard into the Maintenance Rule Monitoring Program	2015Q2	The inspectors identified a Green, non-cited violation of 10 CFR 50.65(b)(2) for the licensees failure to appropriately scope the 230 kV switchyard in the Maintenance Rule monitoring program. Specifically, from the inception of the facilities monitoring program through May 18, 2015, the licensee failed to properly scope or evaluate the 230 kV switchyard to include the entire switchyard up through the first inter-tie circuit breakers CB262 and CB282 into the Maintenance Rule program. Electrical faults within the 230 kV switchyard can cause loss of offsite power which is relied upon to mitigate accidents and cause an actuation of a safety-related systems, such as, emergency diesel generators, and should have been included into its Maintenance Rule program. This issue was entered into the licensees corrective action program as Notifications 50702970 and 50703118. The inspectors determined that the licensees failure to scope the 230 kV offsite power source including the switchyard up through the first breakers from the transmission system into the Maintenance Rule program was contrary to the requirements of 10 CFR 50.65 and therefore a performance deficiency. The performance deficiency was determined to be more than minor because it is associated with the initiating events attribute of protections against external factors and adversely affected the cornerstone objective, in that, a 230 kV switchyard failure can upset plant stability and challenge critical safety functions during shutdown as well as power operations. Failure to monitor the performance or condition of 230 kV offsite power source (including the switchyard up through the first breakers from the transmission system) in a manner sufficient to provide reasonable assurance the offsite power was capable of fulfilling the intended functions affected the reliability of the plant equipment to perform their safety function. The inspectors determined if the 230 kV switchyard was properly scoped into the Maintenance Rule program the loss of offsite power due to the flash over event may have been prevented. However the direct cause of the event has been identified as untimely corrective actions associated with an ineffective corrective action program. As such, improper Maintenance Rule scoping was not the direct cause. Therefore, the inspectors determined the finding could be evaluated using the significant determination process in accordance using IMC 0609, Appendix A, Significance Determination Process (SDP) for Findings At-Power, Exhibit 1, Initiating Events Screening Questions. The inspectors determined that the finding was of very low safety significance (Green) because the finding was determined not to be the cause of the actual 230 kV failure such that all of the screening questions in Exhibit 1 could be answered no. The inspectors determined that since the scoping of the switchyard systems had occurred more than 3 years ago, and the opportunity to reevaluate system scoping had not recently occurred, the finding did not represent current licensee performance and therefore a cross-cutting aspect was not assigned.
05000275/FIN-2015403-02	Licensee-Identified Violation	2015Q2

Semantic search

Query log

Navigation menu