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05000275/FIN-2018404-0330 September 2018 23:59:59Diablo CanyonLicensee-identifiedLicensee-Identified Violation
05000275/FIN-2017002-0330 June 2017 23:59:59Diablo CanyonNRC identifiedFailure to Report a Permanent Medical Condition Within 30 DaysSL -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-2017002-0230 June 2017 23:59:59Diablo CanyonNRC identifiedFailure to Conduct Required Biennial Medical Examinations Within Two YearsSL -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-2015003-0330 September 2015 23:59:59Diablo CanyonLicensee-identifiedLicensee-Identified ViolationPG&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-2012005-0531 December 2012 23:59:59Diablo CanyonLicensee-identifiedLicensee-Identified ViolationThe licensee identified on May 18, 2012, that Diablo Canyon Power Plant implemented changes to the site emergency plan on September 26, 2001, that reduced the plans effectiveness and had the potential to impact the licensees ability to implement protective measures. Title 10 of the Code of Federal Regulations, Part 50.54(q)(4) states in part, The changes to a licensees emergency plan that reduce the effectiveness of the plan may not be implemented without prior approval of the NRC. Contrary to the above, between September 26, 2001, and June 20, 2012, the licensee implemented changes to the site emergency plan that reduced the effectiveness of the plan without prior approval of the NRC. Specifically, the licensee excluded as many as 900 site workers from assembly and accountability, making it more difficult to implement protective measures, either for those individuals or by utilizing those individuals. This finding is more than minor because it affected the NRCs ability to perform its regulatory function. The violation was evaluated using the NRCs Enforcement Policy, Section 6.6.d, and determined to be a Severity Level IV violation because it degraded the licensees ability to meet or implement a regulatory requirement not related to assessment or notification. The licensee documented this issue in their corrective action program as Notification 50483005. This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the NRC Enforcement Policy.
05000275/FIN-2012005-0631 December 2012 23:59:59Diablo CanyonLicensee-identifiedLicensee-Identified ViolationThe licensee identified on August 15, 2011, that Diablo Canyon Power Plant had provided information to the NRC that was not complete and accurate in all material respects. Title 10 of the Code of Federal Regulations, Part 50.9(a), requires, in part, that information provided to the Commission by a licensee shall be complete and accurate in all material respects. Contrary to the above, between August 17, 2005, and June 21, 2011, information provided to the Commission by Pacific Gas and Electric Company was not complete and accurate in all material respects. Specifically, licensee response DCL-05-094, Thirty Day Response to NRC Bulletin 2005-02, stated that site procedures had been modified to ensure that plant page announcements accomplish the described onsite protective measures; however, the plant page system was not adequate for this purpose in that not all personnel required to be covered by protective measures worked in buildings covered by the plant paging system. This violation is more than minor because it affected the NRCs ability to perform its regulatory functions. The finding was evaluated using Section 6.9 of the NRC Enforcement Policy and determined to be a Severity Level IV violation because accurate information would likely not have caused the NRC to reconsider a regulatory position or undertake a substantial further inquiry. The licensee documented this issue in their corrective action program as Notifications 50390230 and 50441808. This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the NRC Enforcement Policy.
05000275/FIN-2010006-0330 September 2010 23:59:59Diablo CanyonNRC identifiedFailure to Submit Complete and Accurate Information for a Requested License AmendmentThe team identified a noncited violation of 10 CFR 50.9(a), Completeness and Accuracy of Information with multiple examples. Specifically, information supplied to the NRC in License Amendment Request 01- 10, dated February 24, 2010, related to the revision of Technical Specification 3.8.1, AC Sources - Operating, were not complete and accurate in all material respects. Following NRC questioning of the discrepancies the licensee withdrew the amendment request. The finding is more than minor because the inaccurate information was material to the NRC. Specifically, this information was under review by the NRC to evaluate specific changes to the surveillance requirements associated with the emergency diesel generators. Following management review, this violation was determined to be of very low safety-significance because the amendment request was withdrawn before the NRC amended the facility technical specifications. Because this issue affected the NRC\'s ability to perform its regulatory function, it was evaluated with the traditional enforcement process. Consistent with the guidance in Section IV.A.3 and Supplement VII, paragraph D.1, of the NRC Enforcement Policy, this finding was determined to be a Severity Level IV noncited violation. The finding had a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program because the licensee did not adequately evaluate the extent of condition and take appropriate corrective actions after the NRC identified a similar violation
05000275/FIN-2010007-1030 June 2010 23:59:59Diablo CanyonNRC identifiedFailure to Update Text to Reflect Credited Design Class I Makeup Flowpath to Component Cooling Water Expansion Tank in the Final Safety Analysis Report UpdateThe team identified a Severity Level IV noncited violation of 10 CFR 50.71, \"Maintenance of records, asking of reports.\" Title 10 CFR 50.71, paragraph (e) states, \"Each person licensed to operate a nuclear power reactor shall update periodically the final safety analysis report originally submitted as part of the application for the license, to assure that the information included in the report contains the latest information developed.\" In the Final Safety Analysis Report Update Table 9.2-7, Component 5, it states, \"This 250 gpm, Design Class I, makeup water flowpath, described under Makeup Provisions in Subsection 2.3.3 (Section 9.2.2.3.3), can be started within 10 minutes.\" Final Safety Analysis Report Update, Section 9.2.2.3.3 states, \"All piping and valves in the makeup path from the condensate storage tank (including their crossconnections) and the firewater tank, through the makeup water transfer pumps up to and including the makeup valves on the component cooling water system lines, are Design Class I.\" Text later in the section implied that the flow path from the firewater tank was not Design Class I. Review by the licensee staff revealed that the only Design Class I flow path to provide makeup to the component cooling \'vvater expansion tank is via the condensate storage tank. This revealed that the text provided in Final Safety Analysis Report Update, Section 9.2.2.3.3 stating that both the condensate storage tank and firewater tank makeup paths are credited is incorrect. Contrary to above, since 1984 (Final Safety Analysis Report Update, Revision 0), the licensee did not update Final Safety Analysis Report Update, Section 9.2.2.3.3 to correct the error of including firewater as a possible makeup path to the component cooling water expansion tank. The licensee has entered this issue into their corrective action process as Notification 50301884. Failure to periodically update the Final Safety Analysis Report Update with a known error is a performance deficiency. Using Inspection Manual Chapter 0612, Appendix B, the team determined that this performance deficiency was to be evaluated using the traditional enforcement process because the performance deficiency had the potential for impacting the NRC\'s ability to perform its regulatory function. Using General Statement of Policy and Procedure for NRC Enforcement Actions, Supplement I, Reactor Operations, dated January 14, 2005, to evaluate the significance of this violation, the team concluded that the violation is more than minor because the incorrect Final Safety Analysis Report Update information had a potential impact on safety and licensed activities. Using Supplement I, Section D, Item 6, of the NRC Enforcement Policy, this performance deficiency will be treated as a Severity Level IV violation. Because this violation is of very low safety significance and entered into the licensee\'s corrective action program, this violation is being treated as a noncited violation, consistent with Section VI.A.1 of the NRC Enforcement Policy. The team reviewed the finding for crosscutting aspects and none were identified.
05000275/FIN-2010007-0130 June 2010 23:59:59Diablo CanyonNRC identifiedLess than Adequate Change Evaluation to the Facility as Described in the Final Safety Analysis Report UpdateThe inspectors identified a Severity Level IV noncited violation of 10 CFR 50.59 for the licensee\'s failure to demonstrate that prior NRC approval was not required prior to making changes to the facility degraded voltage protection scheme as described in the Final Safety Analysis Report Update. In response to this violation, the licensee re-performed the corresponding safety analysis to demonstrate that the subject change to the facility degraded voltage protection scheme was consistent with General Design Criteria 17, \"Electric Power Systems.\" The violation is in the licensee\'s corrective action program as Notification 50306053. The failure of Pacific Gas and Electric to perform a 10 CFR 50.59 evaluation of modifications to the offsite power protection scheme, in accordance with NEI 96 07, was a performance deficiency. The violation was more than minor because of a reasonable likelihood the change to the facility would require Commission review and approval prior to implementation. The violation screened as very low safety significance (Green) because the finding was not a design or qualification deficiency confirmed not to result in loss of operability or functionality, did not represent a loss of system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The finding has a crosscutting aspect in the area of human performance associated with the decision-making component because the licensee did not adopt the requirement to demonstrate that the proposed action was safe in order to proceed rather than a requirement to demonstrate that the proposed action was unsafe in order to disapprove the action, in that the Plant Safety Review Committee did not require that a 50.59 evaluation be performed to demonstrate that the proposed action was safe in order to proceed.
05000275/FIN-2010007-0230 June 2010 23:59:59Diablo CanyonNRC identifiedFailure to Adequately Evaluate Changes to the Diesel Testing as Described in the Final Safety Analysis Report UpdateThe team identified two examples of a Severity Level IV noncited violation of 10 CFR 50.59 after the licensee failed to perform an adequate evaluation to demonstrate that prior NRC approval was not required before making changes to the frequency and voltage recovery criteria and to the diesel testing commitments as described in the Final Safety Analysis Report Update. Specifically, the 1998 Final Safety Analysis Report Update identified a change from Safety Guide 9 to Regulatory Guide 1.9, Revision 2. The scope involved the removal of the KWS relay and included new requirements for voltage and frequency response. This resulted in a reduction in acceptance criteria. The team also identified a second example where the licensee failed to evaluate the 2005 Final Safety Analysis Report Update change from Regulatory Guide 1.9, Revision 2 to Revision 3 for diesel testing and interval frequency. Using NEI 96-07, \"Guidelines for 10 CFR 50.59 Evaluations,\" Revision 1, the team concluded that these changes resulted in a departure from a method of evaluation described in the Final Safety Analysis Report Update establishing the facility design bases. In addition, the licensee\'s 50.59 evaluation, for DCP E-049424, Revision 0, \"EDG Starting, and Loading Capability\" was less than adequate to conclude that prior NRC approval was not required for the changes. The licensee has entered these issues into their corrective action program as Notifications 50302467 and 50302481. The failure of Pacific Gas and Electric to perform an adequate 10 CFR 50.59 evaluation prior to changing the facility as described in the Final Safety Analysis Report Update is a performance deficiency. The violation was more than minor because of a reasonable likelihood the change to the facility would require Commission review and approval prior to implementation. The violation screened as very low safety significance (Green) because the finding was not a design or qualification deficiency confirmed not to result in loss of operability or functionality, did not represent a loss of system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. Because this violation is of very low safety significance and was entered into the licensee\'s corrective action program, this violation is being treated as a noncited violation, consistent with Section VLA.1 of the NRC Enforcement Policy. The finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component. In License Amendment Request 10-01, dated February 24, 2010, the licensee did not thoroughly evaluate the original problem of using the 10 CRF 50.59 evaluation process to justify using Regulatory Guide 1.9, Revision 2, Section C, Position 4, as an exception to meeting the frequency and voltage criteria identified in Safety Guide 9.
05000275/FIN-2010007-0430 June 2010 23:59:59Diablo CanyonNRC identifiedFailure to Update the Final Safety Analysis Report Update with the Current Plant Design Bases SectionThe inspectors identified a Severity Level IV noncited violation of 10 CFR 50.71 after Pacific Gas and Electric failed to include the current plant design basis for the 230kV degraded voltage protection scheme in the Final Safety Analysis Report Update. Title 10 CFR 50.71.(e) states in part, \"Each person licensed to operate a nuclear power reactor shall update periodically, as provided in paragraphs (e) (3) and (4) of this section, the Fina! Safety ,ll.na!ysis Report Update originally submitted as part of the application for the operating license, to assure that the information included in the report contains the latest information developed. Contrary to the above, on March 14, 2010, the inspectors identified that Pacific Gas and Electric failed to update the Final Safety Analysis Report Update to include complete design basis information for the offsite degraded voltage protection scheme. The inspectors identified that Final Safety Analysis Report Update did not include the design basis for the allowable time delay or the limiting voltage setpoints. The licensee has entered this issue into their corrective action process as Notification 50313763. Failure to include the current plant design basis for the 230kV degraded voltage protection scheme in the Final Safety Analysis Report Update is a performance deficiency. Using inspection Manual Chapter 0612, Appendix B, the team determined that this issue was to be evaluated using the traditional enforcement process because the performance deficiency was a failure to meet a requirement or standard, had the potential for impacting the NRC\'s ability to perform its regulatory function, and the concern was within the licensee\'s ability to foresee and correct and should have been prevented. The team used the General Statement of Policy and Procedure for NRC Enforcement Actions, Supplement I \"Reactor Operations,\" dated January 14,2005, to evaluate the significance of this violation. The team concluded that the violation is more than minor because the incorrect Final Safety Analysis Report Update information had a potential impact on safety and licensed activities. Using Supplement I, Section D, Item 6, of the NRC Enforcement Policy, this performance deficiency will be treated as a Severity Level IV violation because the erroneous information was not used to make any unacceptable change to the facility or procedures. Using Inspection Manual Chapter 0609.04, \"Phase 1 Initial Screening and Characterization of Findings,\" the team concluded that the issue screened as having low safety significance (Green) under the significance determination process. Because this violation is of very low safety significance and was entered into the licensee\'s corrective action program, this violation is being treated as a noncited violation, consistent with Section VI.A.1 of the NRC Enforcement Policy. Because the violation included a performance deficiency, the inspectors also concluded the issue was a finding under the Reactor Oversight Process and had a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee did not adequately evaluate the extent of the condition and take appropriate corrective actions after the NRC identified a similar violation.
05000275/FIN-2010003-0330 June 2010 23:59:59Diablo CanyonNRC identifiedFailure to Report a Condition that Could Have Prevented the Fulfillment of a Safety FunctionThe inspectors identified a noncited violation of 10 CFR 50.73(a)(2)(i)(B) and 10 CFR 50.73(a)(2)(v)(B) and after Pacific Gas and Electric failed to submit a required licensee event report within 60 days following discovery of a condition prohibited by the plant technical specifications and a condition that could have prevented the fulfillment of a safety function. On March 9, 2010, Pacific Gas and Electric identified that the degraded voltage protection scheme, required by Technical Specification 3.3.5, Loss of Power Diesel Generator Start Instrumentation, was inadequate to protect operating engineering safety feature pump motors. The licensee concluded that sustained degraded voltage could result in an overcurrent condition affecting equipment powered from the preferred offsite power supply. This condition was required to be reported to the NRC because the degraded voltage protection scheme rendered engineered safety feature pumps inoperable for a period in excess of the allowable technical specification out of service time and the condition resulted in the loss of the degraded voltage protection scheme safety function on all three vital 4 kV power buses. The inspectors evaluated this finding using the traditional enforcement process because the failure to submit a required event report affected the NRCs ability to perform its regulatory function. The inspectors concluded the violation was a Severity Level IV because the licensee failed to submit an adequate licensee event report. The inspectors determined that the violation was also a finding under the reactor oversight process because licensee personnel failed to adequately evaluate a condition adverse to quality for operability and reportability, as required by station procedures. The inspectors concluded that the finding is more than minor because the failure to properly evaluate degraded plant equipment for past operability and reportability could reasonably be seen to lead to a more significant condition. The inspectors concluded that the finding had very low safety significance because the failure to adequately evaluate the condition did not result in an actual loss of a system safety function or equipment required by technical specifications, or involve the loss or degradation of equipment specifically designed to mitigate a seismic, flooding, or sever weather initiating event, and did not involve the total loss of any safety function that contributes to an external event initiated core damage accident sequence This finding has a crosscutting aspect in the area of problem identification and resolution, associated with the corrective action program component because the licensee failed to perform an adequate evaluation of the degraded voltage protection scheme such that the resolutions address causes and extent of conditions, as necessary.
05000275/FIN-2010007-0930 June 2010 23:59:59Diablo CanyonNRC identifiedFailure to Update Feedwater Rupture Accident Analysis in the Final Safety Analysis Report UpdateThe team identified a Severity Level IV noncited violation of 10 CFR 50.71, \"Maintenance of records, making of reports.\" Paragraph (e) states, \"Each person licensed to operate a nuclear power reactor shall update periodically the final safety analysis report originally submitted as part of the application for the license, to assure that the information included in the report contains the latest information developed.\" in the Diablo Canyon Final Safety Analysis Report Update section addressing the feedwater line break accident, it states that operator actions are credited with precluding the operation of pressurizer safety valves based on determinations in Westinghouse study WCAP-11667 (1998) (Final Safety Analysis Report Update, Section 15.4.2.2.2). Review of this study, and associated correspondence on the topic during 2006 indicated that the Westinghouse study did not state that operator actions could be credited for this event, but analysis of the worst case pressurizer overfill accidents by the licensee may show that this is the bounding case for such accidents, and that it did not need to be addressed in the feedwater line break analysis. In 2006, the licensee indicated that they would revise the Final Safety Analysis Report Update text to remove this reference to the Westinghouse study, which had been in the Final Safety Analysis Report Update since Revision 16. Contrary to the above, since 2006 (Final Safety Analysis Report Update Revision 16), the licensee failed to update Final Safety Analysis Report Update, Section 15.4.2.2.2. The licensee has entered this issue into their corrective action process as Notification 50301747. Failure to periodically update the Final Safety Analysis Report Update with a known error is a performance deficiency. Using Inspection Manual Chapter 0612, Appendix B, the team determined that this issue was to be evaluated using the traditional enforcement process because the performance deficiency was a failure to meet a requirement or standard, had the potential for impacting the NRC\'s ability to perform its regulatory function, and the concern was within the licensee\'s ability to foresee and correct and should have been prevented. The team used the General Statement of Policy and Procedure for NRC Enforcement Actions, Supplement I, \"Reactor Operations,\" dated January 14, 2005, to evaluate the significance of this violation. The team concluded that the violation is more than minor because the incorrect Final Safety Analysis Report Update information had a potential impact on safety and licensed activities. Using Supplement I, Section 0, Item 6, of the NRC Enforcement Policy, this performance deficiency will be treated as a Severity Level IV violation. Because this violation is of very low safety significance and entered into the licensee\'s corrective action program, this violation is being treated as a noncited violation, consistent with Section VI.A.1 of the NRC Enforcement Policy. The inspectors reviewed the finding for crosscutting aspects and none were identified.
05000275/FIN-2010002-0231 March 2010 23:59:59Diablo CanyonNRC identifiedFailure to Update the Final Safety Analysis Report with the Current Plant Design BasesThe inspectors identified a noncited violation of 10 CFR 50.71 after Pacific Gas and Electric failed to update the Final Safety Analysis Report Update with the current design basis. The inspectors identified that the current Final Safety Analysis Report Update, Revision 18, Sections 3.1, 6.4, 6.5, and 9.4 did not capture the current design basis for the control room, component cooling water, and auxiliary feedwater systems. The failure of the licensee to provide current design basis information in the Final Safety Analysis Report Update had an adverse impact on the plant modification process, the licensees ability to assess operability for degraded plant systems, and the NRCs ability to ensure that regulatory requirements were met. The licensee entered this violation into the corrective action program as Notifications 50308588, 50306131, 5030799, and 50307476. The inspectors evaluated this violation using the traditional enforcement process because the issue affected the NRCs ability to perform its regulatory function. The inspectors concluded that the violation is more than minor because the incorrect Final Safety Analysis Report Update information had a potential impact on safety and licensed activities. The inspectors concluded the violation is Severity Level IV because the erroneous information was not used to make an unacceptable change to the facility or procedures that would have resulted in greater than very low safety significance under the Significance Determination Process. Because the violation included a performance deficiency, the inspectors also concluded the issue was a finding under the Reactor Oversight Process. The finding had a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee did not adequately evaluate the extent of condition of previous similar violation and take appropriate corrective actions (P.1(c))
05000275/FIN-2010002-0331 March 2010 23:59:59Diablo CanyonNRC identifiedFailure to Report a Condition That Could Have Prevented the Fulfillment of a Safety FunctionThe inspectors identified a noncited violation of 10 CFR 50.73(a)(1) after Pacific Gas and Electric failed to submit a required licensee event report within 60 days after discovering a condition that could have prevented the fulfillment of a safety function. On November 22, 2005, the licensee determined that plant operators may not have had the capability to align either residual heat removal train to the cold leg recirculation mode of emergency core cooling following certain small break loss of coolant accidents. Plant engineers determined that the residual heat removal containment sump suction valve operators were inadequately sized to open against the differential pressure generated by the pumps operating in recirculation for an extended period. Plant engineers identified this condition during a follow up of industry operating experience. The licensee initially concluded that the condition was not reportable because the operating experience was not applicable to Diablo Canyon. The licensee failed to re-screen the issue for reportability after determining that the plant was susceptible to the condition. The licensee entered this issue into the corrective action program as Notifications 50301839 and 50295784. The inspectors evaluated this finding using the traditional enforcement process because the failure to submit a required event report affected the NRCs ability to perform its regulatory function. Consistent with the guidance in Section IV.A.3 and Supplement I, Paragraph D.4, of the NRC Enforcement Policy, the inspectors concluded the violation was a Severity Level IV because the licensee failed to submit a required licensee event report. The inspectors did not assign a crosscutting aspect because the performance deficiency represented a latent issue
05000275/FIN-2010002-0531 March 2010 23:59:59Diablo CanyonNRC identifiedFailure to Submit a LERE for a Common-Cause Inoperability of Independent Trains or ChannelsThe inspectors identified a noncited violation of 10 CFR 50.73(a)(1) after Pacific Gas and Electric failed to submit a required licensee event report within 60 days after discovery of a common-cause failure of three control room radiation monitors. The inspectors concluded that monitors failed on October 13, 2009 as a result of water intrusion due to heavy rains. The inspectors concluded that common cause failure of the radiation monitors was reportable under 10 CFR 50.73(a)(2)(vii). Pacific Gas and Electric subsequently reported the event on February 17, 2010, as Licensee Event Report 2010-001-00, Control Room Ventilation Pressurization Due to Radiation Detector Failures. The licensee entered this issue into the corrective action program as Notification 50301839. The inspectors evaluated this finding using the traditional enforcement process because the failure to submit a required event report affected the NRCs ability to perform its regulatory function. Consistent with the guidance in Section IV.A.3 and Supplement I, Paragraph D.4, of the NRC Enforcement Policy, the inspectors concluded that this was a Severity Level IV noncited violation because the licensee failed to submit a required licensee event report. Because the violation included a performance deficiency, the inspectors also concluded the issue was a finding under the Reactor Oversight Process. This finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee failed to thoroughly evaluate the failure of the radiation monitor failures to ensure NRC reportability requirements were met (P.1(c))
05000323/FIN-2009009-0331 March 2010 23:59:59Diablo CanyonNRC identifiedFailure to Evaluate a Change to the Facility as Described in the Final Safety Analysis Report Update Associated with the Addition of Manual Actions in the Safety AnalysisThe inspection team identified a noncited violation of 10 CFR 50.59, which states that a licensee may make changes to the facility as described in the final safety analysis report without obtaining a license amendment if the change does not result in a departure from a method of evaluation described in the final safety analysis report used in establishing the design bases or in the safety analyses. This regulation further requires the licensee to include a written evaluation providing the basis for concluding that a license amendment is not required. On November 21,2005, the licensee failed to provide a written evaluation concluding that a license amendment was not required for a change to the facility as described in the final safety analysis report. Specifically, the licensee identified a condition where large differential pressure across the residual heat removal suction containment sump valves could cause them to fail to open during certain small break loss of coolant accidents. On October 5, 2005, the licensee revised Emergency Operating Procedure E-1, Loss of Reactor or Secondary Coolant, to add an operator action to align component cooling water to the residual heat removal heat exchanger. On June 16, 2009, the licensee again revised Emergency Operating Procedure E-1 to specify that operator action to align component cooling water within 30 minutes was a time critical operator action. The licensee did not evaluate either change to determine if prior NRC approval was required for the new manual actions. The licensee entered this issue into their corrective action program as Notification 50276288. The failure of the licensee to perform a 10 CFR 50.59 evaluation of a new manual action supporting the plants design basis was a performance deficiency within the licensees ability to foresee and correct. The inspectors evaluated this issue using the traditional enforcement process because the performance deficiency had the potential for impacting the NRCs ability to perform its regulatory function. The inspectors concluded that the issue was more than minor because of a reasonable likelihood that the change to the facility would require Commission review and approval prior to implementation. The inspectors also evaluated the significance of this issue under the Significance Determination Process using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The inspectors concluded that the issue affected the Mitigating Systems Cornerstone and screened Green because the finding was a design or qualification deficiency confirmed not to result in loss of operability. The issue was classified as Severity Level IV because the violation of 10 CFR 50.59 involved conditions resulting in very low safety significance by the significance determination process. This finding had a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee did not thoroughly evaluate the change to the facility as described in the Final Safety Analysis Report Update to determine if prior NRC approval was required (P.1 (c)
05000275/FIN-2009005-0331 December 2009 23:59:59Diablo CanyonNRC identifiedInadequate 50.59 Evaluation for Steam Generator Tube Rupture AnalysisThe inspectors identified a Severity Level IV noncited violation of 10 CFR 50.59 after Pacific Gas and Electric failed to perform an adequate evaluation of a change to the facility as described in the Final Safety Analysis Report Update. In 1992, the licensee identified that auxiliary feedwater and steam generator power-operated relief valve flow rates assumed in the steam generator tube rupture accident analysis were non-conservative. To address the non-conforming condition, Pacific Gas and Electric changed the accident analysis to include a new time critical operator action to terminate turbine-driven auxiliary feedwater flow 5.54 minutes after the reactor trip and credit motor driven auxiliary feedwater automatic level control to the ruptured steam generator. The licensee did not perform a 10 CFR 50.59 safety evaluation of these changes. The NRC basis of approval of the accident analysis include four time critical operator actions, each assumed to occur after the first 10 minutes following the accident. The inspectors concluded that NRC approval was required before the licensee added the new time critical manual action under the 10 CFR 50.59 Rule in effect at the time because the change reduced the margin to safety to the basis of Technical Specification 3.7.4, 10% Atmospheric Dump Valves. The inspectors also concluded that prior NRC approval was required under the current 50.59 Rule because the change result in a departure from a method of evaluation described in the Final Safety Analysis Report Update. The performance deficiency, a less than adequate 50.59 evaluation, was the result of a latent issue. However, the inspectors concluded that the licensee had reasonable recent opportunities to identify the problem. The inspectors also concluded that plant programs, processes or organizations have not changed such that the problem would not reasonably occur today and that the most significant contributor to the performance deficiency was reflective of current plant performance. The licensee entered this issue into their corrective action program as Notification 50270786. The failure of Pacific Gas and Electric to perform a 10 CFR 50.59 evaluation of the changes to the steam generator tube rupture accident analysis was a performance deficiency. The inspectors evaluated this issue using traditional enforcement because the performance deficiency had the potential for impacting the NRCs ability to perform its regulatory function. The issue was more than minor because of reasonable likelihood the change to the facility would require Commission review and approval prior to implementation. The inspectors also evaluated the significance of this issue under the Significance Determination Process using Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings. The finding affected the Mitigating Systems Cornerstone because the change described the operator actions required to mitigate steam generator tube rupture accident. The inspectors concluded the finding screened Green because the finding was a design deficiency that did not result in the loss of operability or functionality. The inspectors concluded that the violation was a Severity Level IV because the issue screened Green under the Significance Determination Process. The inspectors concluded that this finding had a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee did not thoroughly evaluate the steam generator tube rupture analysis such that the resolutions addressed causes and extent of condition (P.1(c
05000275/FIN-2009005-0531 December 2009 23:59:59Diablo CanyonNRC identifiedLess than Adequate Change Evaluation to the Facility as Described in the Final Safety Analysis Report UpdateThe inspectors identified a Severity Level IV noncited violation of 10 CFR 50.59 after the licensee failed to perform an adequate evaluation to demonstrate that prior NRC approval was not required before making changes to the facility as described in the Final Safety Analysis Report Update. In October 2009, the inspectors identified that the replacement reactor head contractor used incorrect damping values in the replacement head design. The contractor was unable to demonstrate that the design met ASME Code using the damping values specified in the plant design basis. On November 5, 2009, the licensee incorporated the new damping values and revised the method for determining the seismic response spectra. Using NEI 96-07, Guidelines for 10 CFR 50.59 Evaluations, Revision 1, the inspectors concluded that these changes resulted in a departure from a method of evaluation described in the Final Safety Analysis Report Update establishing the facility design bases. The licensees 50.59 evaluation, Licensing Basis Impact Evaluation LBIE 2009-021, Integrated Head Assembly, was less than adequate to conclude that prior NRC approval was not required for the changes. The licensee entered this issue into their corrective action program as 50276288. The failure of Pacific Gas and Electric to perform an adequate 10 CFR 50.59 evaluation prior to changing the facility as described in the Final Safety Analysis Report Update is a performance deficiency. The inspectors evaluated this issue using the traditional enforcement process because the performance deficiency had the potential for impacting the NRCs ability to perform its regulatory function. The inspectors concluded that the issue was more than minor because of a reasonable likelihood the change to the facility would require Commission review and approval prior to implementation. The inspectors also evaluated this issue using the Significance Determination Process. The inspectors concluded that the violation affected the Initiating Events Cornerstone because the change potentially decreased the structural integrity of the control rod drive mechanism reactor coolant pressure barrier and screened Green because assuming worst case degradation, the finding would not result in exceeding the technical specification limit for reactor coolant system leakage nor have a likely effect on other mitigation systems resulting in a total loss of their safety function. The inspectors concluded that the violation was a Severity Level IV because the issue screened Green under the Significance Determination Process. The finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee did not thoroughly evaluate the original problem associated with the replacement reactor head design such that the resolutions address causes and extent of conditions, as necessary (P.1(c)
05000275/FIN-2009004-0430 September 2009 23:59:59Diablo CanyonNRC identifiedFailure to Update the Final Safety Analysis Report Update with Current Accident AnalysisThe inspectors identified a noncited violation of 10 CFR 50.71 after Pacific Gas and Electric failed to update the Final Safety Analysis Report Update with a critical operator action assumed in the plant steam generator tube rupture accident analysis. The steam generator tube rupture accident analysis assumed that the ruptured steam generator will not overfill with water during the accident. To ensure a margin to overfill, the accident analysis included a critical assumption that plant operators would manually trip the turbine-driven auxiliary feedwater pump within 5.54 minutes following the reactor trip. Final Safety Analysis Report Update Section 15.4.3.1, Identification of Causes and Accident Description, and Final Safety Analysis Report Update Table 15.4-12, Operator Action Times for Design Basis SGTR Analysis, provided a detailed description of the time dependant operator actions assumed in the accident analysis. The inspectors identified that neither section included the critical assumed operator action to trip the turbine-driven auxiliary feedwater pump. The inspectors concluded that the licensee had a reasonable opportunity to identify and correct the problem when the results of the revised steam generator tube rupture accident, supporting steam generator replacement, was updated in the Final Safety Analysis Report Update in October 2008. The licensee entered this violation into the corrective action program as Notification 50269753. The inspectors evaluated this finding with the traditional enforcement process because the issue affected the NRCs ability to perform its regulatory function. The inspectors concluded that the finding is greater than minor because the failure to update the required critical operator action assumed in the accident analysis could have a material impact on safety or licensed activities. The inspectors concluded that the violation is Severity Level IV because the erroneous information was not used to make an unacceptable change to the facility or procedures. The inspectors concluded that this finding had a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee failed to implement a corrective action program with a low threshold for identifying issues and failed to identify the inaccuracies in the accident analysis as described in the Final Safety Analysis Report Update (P.1(a)
05000275/FIN-2009003-0630 June 2009 23:59:59Diablo CanyonNRC identifiedFailure to Evaluate a Change to the Facility as Described in the Final Safety Analysis Report Update Associated with 500 kV Offsite Power SourceThe inspectors identified a noncited violation of 10 CFR 50.59 after Pacific Gas and Electric failed to perform an adequate evaluation of a thermal hydraulic analysis to determine if prior NRC approval was required for a 30-minute delay time to align offsite power. This analysis, Calculation STA-274, RETRAN Evaluation of GDC-17 Loss of AC Scenario, Revision 0, demonstrated that the 30-minute delayed offsite power source was acceptable. On December 31, 2008, a Pacific Gas and Electric 10 CFR 50.59 screen concluded that Calculation STA-274 was not required to be evaluated to determine if prior NRC approval was required for the delay time. On March 31, 2009, the inspectors concluded that the licensee was required to evaluate Calculation STA-274 to determine if prior NRC approval was needed. On May 27, 2009, Pacific Gas and Electric completed the 50.59 evaluation and concluded that prior NRC approval was required for the 30-minute delay time to align offsite power. The inspectors concluded that the finding is more than minor because the changes made to the facility required prior NRC review and approval. The finding affected the Mitigating Systems Cornerstone because the change described how the delayed offsite power source met the design basis. The inspectors concluded the finding is of very low safety significance because the finding was a design deficiency that did not result in the loss of operability or functionality. Because the issue affected the NRCs ability to perform its regulatory function, the inspectors evaluated this finding using the traditional enforcement process. This issue was classified as Severity Level IV because the violation of 10 CFR 50.59 involved conditions resulting in very low safety significance by the significance determination process. This finding had a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because Pacific Gas and Electric did not thoroughly evaluate the change to the facility as described in the Final Safety Analysis Report Update to determine if prior NRC approval was required (P.1(c))(Section 40A5)
05000275/FIN-2009003-0230 June 2009 23:59:59Diablo CanyonNRC identifiedFailure to Submit a Licensee Event Report for a Condition Prohibited by the Plants Technical SpecificationsThe inspectors identified a noncited violation of 10 CFR 50.73(a)(1) after Pacific Gas and Electric failed to submit a required licensee event report within 60 days after discovery of a condition prohibited by technical specifications. The licensee failed to correctly evaluate the March 18, 2009, failure of the Unit 2 control rod demand position indicators for reportability. The inspectors concluded that the failure of control rod position indicators was a condition prohibited by Technical Specification 3.17, Rod Position Indication. This finding is greater than minor because the NRC relies on licensees to identify and report conditions or events meeting the criteria specified in the regulations in order to perform its regulatory function. This finding affected the mitigating systems cornerstone. Because this issue affected the NRC\\\'s ability to perform its regulatory function, it was evaluated with the traditional enforcement process. Consistent with the guidance in Section IV.A.3 and Supplement I, Paragraph D.4, of the NRC Enforcement Policy, this finding was determined to be a Severity Level IV, noncited violation. The licensee entered this issue into the corrective action program as Notification 50242153. This finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee failed to thoroughly evaluate the failure of the Unit 2 control rod demand position indicators for reportability (P.1(c))(Section 40A2
05000275/FIN-2009003-0330 June 2009 23:59:59Diablo CanyonNRC identifiedFailure to Update the Final Safety Analysis Report Update with Current Plant Design CriteriaThe inspectors identified a noncited violation of 10 CFR Part 50.71 after Pacific Gas and Electric failed to update the Final Safety Analysis Report Update with current plant design criteria. The Final Safety Analysis Report Update stated that Diablo Canyon was designed to comply with the Atomic Energy Commission General Design Criteria for Nuclear Power Plant Construction Permits, published in July 1967. The inspectors identified that the Diablo Canyon Safety Evaluation Report stated that the NRC used General Design Criteria published in July 1971 to review the plant design. In addition, during the initial licensing process, the licensee stated that the plant was evaluated against the 1971 design criteria during the licensing process. The inspectors evaluated this finding using the traditional enforcement process because the failure to update the Final Safety Analysis Report affected the NRCs ability to perform its regulatory function. The inspectors concluded that the failure to update the Final Safety Analysis Report was a Severity Level IV violation based on the General Statement of Policy and Procedure for NRC Enforcement Actions, Supplement I Reactor Operations, dated January 14, 2005, because the erroneous information was not used to make an unacceptable change to the facility or procedures. The inspectors concluded that this finding had a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee did not take appropriate corrective actions to address safety issues and adverse trends in a timely manner (P.1(d))(Section 40A2
05000275/FIN-2008009-0131 December 2008 23:59:59Diablo CanyonNRC identifiedFailure to update the Final Safety Analysis ReportThe team identified a non-cited violation of 10 CFR 50.71(e) for the failure of the licensee to periodically (every 24 months) update its Final Safety Analysis Report Update with all changes made in the facility or procedures. Specifically, in July 2005, the licensee stopped using the boric acid evaporator system as described in the Final Safety Analysis Report Update, Section 11.2.6, and did not submit an update to the NRC regarding this operational change. This issue was entered into the licensees corrective action program as Notification 50116337. The team determined that the failure to update the Final Safety Analysis Report Update to reflect changes made to the facility was a performance deficiency. This issue is subject to traditional enforcement because it had the potential for impacting the NRCs ability to perform its regulatory function. The finding is characterized as a Severity Level IV, non-cited violation in accordance with NRC Enforcement Policy, Supplement I, Example D.6, in that, the erroneous information in the Final Safety Analysis Report Update was not used to make an unacceptable change to the facility or procedures. (Section 2PS1
05000275/FIN-2005005-0631 December 2005 23:59:59Diablo CanyonNRC identifiedFailure to Accurately Assess and Report Performance Indicator DataThe inspector identified an noncited violation of 10 CFR 50.9 because Pacific Gas and Electric Company failed to provide complete and accurate information in a submittal of data for the emergency preparedness drill and exercise performance indicator. Specifically, Pacific Gas and Electric Company staff failed to identify three missed opportunities for emergency notification accuracy during the second calendar quarter of 2005. Pacific Gas and Electric Company took prompt action to correct the second quarter data, which resulted in the drill and exercise performance indicator color to cross from GREEN to WHITE. Pacific Gas and Electric Company also initiated a 100 percent review of the second and third quarter drill and exercise performance indicator data and discovered one additional administrative error in the third quarter performance indicator data, which had been previously evaluated, but not yet reported to the NRC. Pacific Gas and Electric Company had previously initiated a root cause evaluation in its corrective action program to determine the reason for the declining indicator and, subsequently, initiated another root cause evaluation to determine the reason for the failure to adequately evaluate and report the performance indicator data. The finding also had human performance crosscutting aspects in that the reviews that were performed were not adequate to identify the actual failures that had occurred Because this issue affected the NRC's ability to perform its regulatory function, it was evaluated using the traditional enforcement process. Supplement 7, Section D.3, of the NRC Enforcement Policy describes this finding as a Severity Level IV violation. The issue is significant because it indicates a declining trend in the attention to detail shown by senior licensed operators in performing emergency notifications to the state and local authorities. This issue is documented in Pacific Gas and Electric Company's corrective action program as Nonconformance Report NO002200.