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| Start date | Site | Identified by | Title | Description | |
|---|---|---|---|---|---|
| 05000361/FIN-2012009-02 | 30 June 2013 23:59:59 | San Onofre | NRC identified | Failure to Verify Adequacy of Thermal-Hydraulic and Flow-Induced Vibration Design for the Unit 3 Replacement Steam Generators | The inspectors identified an apparent violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to verify the adequacy of the thermal-hydraulic and flow-induced vibration design of the Unit 3 replacement steam generators, which resulted in significant and unexpected steam generator tube wear after 11 months of operation and an associated apparent violation of Technical Specification 5.5.2.11, Steam Generator Program, loss of tube integrity on Unit 3 Steam Generator 3E0- 88. The licensee initiated Nuclear Notification NN 202447265 to address this issue in the corrective action program. Southern California Edison revised the thermal-hydraulic code of record and ensured that the code was in accordance with ASME guidance. Subsequently, on June 7, 2013, Southern California Edison announced that Units 2 and 3 would be permanently shut down. This finding is more than minor because it is associated with the equipment performance attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the failure to verify the adequacy of the thermal-hydraulic and flow-induced vibration design resulted in excessive and rapid tube wear due to fluid elastic instability, which challenged the structural integrity of the tubes to perform their pressure boundary function. The inspectors used NRC Inspection Manual Chapter 0609, Attachment 4 and Appendix A, to evaluate the significance of this finding. In accordance with Exhibit 1 of Inspection Manual Chapter 0609, Appendix A, the inspectors determined that this finding required evaluation in accordance with Inspection Manual Chapter 0609, Appendix J, because the finding involved a degraded steam generator tube condition where one tube could not sustain three times the differential pressure across a tube during normal full power, steady-state operation. In accordance with Inspection Manual Chapter 0609, Appendix J, this finding required a detailed risk analysis, since it involved two or more tubes that could not sustain three times the normal differential pressure and one or more steam generators that violated accident-induced leakage performance criterion. A Phase 3 analysis was completed using the San Onofre SPAR model, Revision 8.22, assuming average test and maintenance, and a truncation limit of 1.0E-11. Based on the best available information, the performance deficiency was preliminarily characterized as a finding of low to moderate safety significance (White). The final significance of this finding is to be determined. No cross-cutting aspect was assigned because this performance deficiency occurred in the 2005 to 2008 timeframe. Substantial management and personnel changes have occurred, including taking actions to address a chilled work environment and other safety culture issues. The NRC determined that the performance behavior that existed at that time is not indicative of current performance |
| 05000361/FIN-2013003-02 | 30 June 2013 23:59:59 | San Onofre | NRC identified | Procedure Adequacy for Responding to External Flooding Events | The inspectors identified an unresolved item for a potentially inadequate procedure for response to external flooding events and potential failure to follow procedures to complete an appropriate evaluation for inaccessible flood protection features. The inspectors reviewed alarm response Procedure SO23-15-57.C, Alarm Response Instruction, Revision 28, which included a guide for identifying which flood sensor area was transmitting the alarm to the control room. The procedure included step 1.0, Required Actions, step 2.0, Corrective Actions, step 3.0, Associated Responses, and step 4.0, Compensatory Actions. Step 1.0 included immediately dispatching an operator to ensure doors and manholes in the area were closed. Step 2.0 listed specific causes that may be the source of internal flooding from specific components (tanks, pumps, valves, etc.). Step 3.0 included steps to isolate or secure the component identified as the cause of the internal flooding, as well as initiation of Procedure SO23-2-16, Operating Instruction, Revision 36, for removing water by use of temporary sump pumps for some of the spaces, where other spaces relied on internal sumps. The procedure did not provide details on manpower, equipment locations, equipment maintenance, disposal of flood waters, nor analysis supporting how quickly the water would need to be removed to prevent affecting safety-related equipment or spaces for conditions associated with an external flooding event. As part of this inspection, the inspectors reviewed Calculation M-0120-015, Plant Flood Analysis, which included an assessment of the worst case internal flooding event in spaces, and compared it to external flooding scenarios. Additionally, the inspectors observed that Procedure SO-23-XV-94, step 6.6.2.1.1, required that, if more than one inaccessible flood protection feature with potential loss of function is reported, then an evaluation of the aggregate effect on flood protection features must be provided. The licensee did not evaluate a possible common cause failure of these features and its cumulative effect on flooding protection features on site, as required in NEI 12-07, Section 5.1, which stated in part that, if more than one inaccessible flood protection feature with potential loss of function is reported, then an evaluation of the aggregate effect flood protection features must be provided. The licensee documented these issues in the corrective action program as Nuclear Notifications NN202370058 and NN202157052. The inspectors had a number of questions and concerns regarding assumptions and conclusions from the flooding analysis that were not resolved at the end of the inspection period. Additional inspection is necessary to complete the review and determine whether procedures were inadequate for response to external flooding events and whether an evaluation for certain inaccessible flood protection features was appropriate and necessary. As a result, the inspectors identified Unresolved Item URI 05000361;362/2013003-02, Procedure Adequacy for Responding to External Flooding Events. |
| 05000361/FIN-2012009-01 | 30 June 2013 23:59:59 | San Onofre | NRC identified | Failure to Verify Adequacy of Thermal-Hydraulic and Flow-Induced Vibration Design for the Unit 2 Replacement Steam Generators | The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to verify the adequacy of the thermal-hydraulic and flow-induced vibration design of the Unit 2 replacement steam generators, resulting in excessive and unexpected steam generator tube wear after one cycle of operation. The licensee initiated Nuclear Notification NN 202447268 to address this issue in the corrective action program. Southern California Edison revised the thermal-hydraulic code of record and ensured that the code was in accordance with ASME guidance. Subsequently, on June 7, 2013, Southern California Edison announced that Units 2 and 3 would be permanently shut down. The finding is more than minor because it is associated with the equipment performance attribute of the Initiating Event Cornerstone and adversely affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors used NRC Inspection Manual Chapter 0609, Attachment 4 and Appendix A, to evaluate the significance of this finding. In accordance with Exhibit 1 of Inspection Manual Chapter 0609, Appendix A, the inspectors determined that the finding was of very low safety significance because the finding did not involve a degraded steam generator tube that could not sustain three times the normal operating differential pressure and did not violate the accident leakage performance criterion. No cross-cutting aspect was assigned because this performance deficiency occurred in the 2005 to 2008 timeframe. Substantial management and personnel changes have occurred, including taking actions to address a chilled work environment and other safety culture issues. The NRC determine d that the performance behavior that existed at that time is not indicative of current performance. |
| 05000361/FIN-2013003-01 | 30 June 2013 23:59:59 | San Onofre | NRC identified | Failure to Properly Scope All the Pertinent External Flood Protection Features into the Walkdown List in Accordance with Industry Guidance NEI 12-07 | The inspectors identified one finding of very low safety significance for the licensees failure to follow procedures regarding the Fukushima event response for flood protection to comply with NRC endorsed NEI 12-07, Guidelines for Performing Walkdowns of Plant Flood Protection Features. Specifically, the licensee failed to evaluate the conduits beneath the grating of the diesel generator building for inclusion in the walkdown scope and failed to establish adequate procedures that included accurate assessment of the Available Physical Margin of flooding protection features included in the flooding walkdown scope. This finding was entered into the licensees corrective action program as Nuclear Notifications NN 202369978 and NN 202375161. The performance deficiency is greater than minor, and therefore a finding, because it is associated with the Mitigating Systems Cornerstone attribute of Protection Against External Factors (Flood Hazard) and it adversely affects the associated cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined that the finding could be evaluated using the significance determination process in accordance with IMC 0609, Significance Determination Process, and conducted a Phase 1 characterization and initial screening. Phase 1 initial screening determined that IMC 0609 Appendix A, Exhibit 2, Mitigating Systems Screening Questions, should be used. Because the finding did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding, or severe weather initiating event (e.g., seismic snubbers, flooding barriers, tornado doors), the finding screened as Green. The finding has a cross-cutting aspect in the area of human performance, associated with the decision-making component, because the licensee did not verify the validity of the underlying assumptions and identify possible unintended consequences. |
| 05000361/FIN-2013403-02 | 31 March 2013 23:59:59 | San Onofre | NRC identified | Security | |
| 05000361/FIN-2013002-01 | 31 March 2013 23:59:59 | San Onofre | NRC identified | Failure to Completely Inspect and Maintain PMF Berm | The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to accomplish activities in accordance with procedures. Specifically, prior to March 4, 2013, the licensee failed to accomplish inspections and maintenance of the downstream face of the probable maximum flood berm in accordance with Attachments 1 and 3 of Procedure SO123-XVIII-35, Inspection and Maintenance of Seawall, Offsite Probable Maximum Flood Berm and Channel, and Related Drainage Facilities. These issues have been entered into the licensees corrective action program as Nuclear Notifications NN 202346674, NN 202354058, and NN 202359197. The inspectors determined that the licensees failure to accomplish inspections and maintenance in accordance with Procedure SO123-XVIII-35 was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because, if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Specifically, the licensee routinely failed to maintain and inspect the downstream face of the berm for vegetation overgrowth, structural integrity, and animal burrows, resulting in identified degradation conditions during subsequent inspections. Using NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, the finding screened as potentially risk important, affecting the Mitigating Systems Cornerstone attribute for external events mitigating systems, because the finding resulted in the degradation of equipment specifically designed to mitigate a flooding initiating event. Therefore, a Region IV senior reactor analyst performed a detailed risk evaluation using NRC Inspection Manual Chapter 0609, Appendix M, Significance Determination Process Using Qualitative Criteria. Based on the inspectors observation of the condition of the berm, the senior reactor analyst determined that, even though the berm was degraded, it remained functional. Since the probable maximum flood berm remained functional, there was no quantifiable change to the core damage frequency or the large early release frequency. Therefore, the finding was of very low safety significance (Green). This finding had a crosscutting aspect in the area of human performance, resources component, because the licensee did not ensure personnel were available and adequate to assure nuclear safety |
| 05000361/FIN-2013002-02 | 31 March 2013 23:59:59 | San Onofre | NRC identified | Failure to Properly Screen Nuclear Notifications Results in Missed Operability Determinations and Functionality Assessments | The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure of operations personnel to implement procedures associated with evaluating the impact of degraded or nonconforming conditions on the operability of equipment required by technical specifications. Specifically, between December 2010 and February 2013, the inspectors identified 15 examples of operations personnel failing to follow Procedure SO123-XV-50.CAP-2, SONGS Nuclear Notification Screening, Attachment 3, step 6.2.9, resulting in the failure to complete the immediate operability determination or the immediate functionality assessment as required. This issue has been entered into licensees corrective action program as Nuclear Notification NN 202337603. The inspectors determined that the failure of operations personnel to follow Procedure SO123-XV-50.CAP-2 for screening nuclear notifications was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it is associated with the Mitigating Systems Cornerstone attribute for equipment performance and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined that Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, was appropriate based on the plant conditions present when most of the examples of this performance deficiency occurred. The finding did not require a quantitative assessment because adequate mitigating equipment remained available and the finding did not constitute a loss of control, as defined in Appendix G. Therefore, the finding screened as having very low safety significance (Green). This finding had a crosscutting aspect in the area of the human performance decision-making component, because operations personnel used nonconservative assumptions about depth of corrosion and corrosion rates to screen multiple degraded or nonconforming conditions out of the operability determination process |
| 05000361/FIN-2013403-04 | 31 March 2013 23:59:59 | San Onofre | Licensee-identified | Security | |
| 05000361/FIN-2013405-01 | 31 March 2013 23:59:59 | San Onofre | Licensee-identified | Licensee-Identified Violation | |
| 05000361/FIN-2013002-04 | 31 March 2013 23:59:59 | San Onofre | NRC identified | Untimely Corrective Actions for Nitrogen Gas Accumulation in the Auxiliary Feedwater System | The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions, associated with the licensees failure to take appropriate and prompt corrective actions regarding nitrogen gas accumulation in the safety-related auxiliary feedwater system. Specifically, from March 2012 until January 2013, a condition adverse to quality related to the - 6 - accumulation of gas, from steam generator nitrogen purge into piping and safety-related pumps in the Unit 2 auxiliary feedwater system, was not promptly identified and corrected until a gas binding event occurred during a start of an auxiliary feedwater pump in Unit 3 on January 2, 2013. This issue has been entered into the licensees corrective action program as Nuclear Notifications NN 202268941 and NN 202382092. The inspectors determined that the failure to take prompt corrective actions for nitrogen gas accumulation in the safety-related auxiliary feedwater system as required by 10 CFR Part 50, Appendix B, Criterion XVI was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the Mitigating Systems Cornerstone attribute for equipment performance and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined that Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, was appropriate based on the plant conditions present when this performance deficiency occurred. The finding did not require a quantitative assessment because adequate mitigating equipment remained available and the finding did not constitute a loss of control, as defined in Appendix G. Therefore, the finding screened as having very low safety significance (Green). The inspectors determined that the finding had a crosscutting aspect in the area of the human performance decision-making component, because the licensee did not make safety-significant or risk-significant decisions using a systematic process when they identified a degraded condition of gas accumulation in the auxiliary feedwater system |
| 05000361/FIN-2013403-03 | 31 March 2013 23:59:59 | San Onofre | Licensee-identified | Security | |
| 05000361/FIN-2013002-06 | 31 March 2013 23:59:59 | San Onofre | NRC identified | Failure to Follow Procedure for Plant Preservation Rust Grading | The inspectors identified a Green finding for failure to follow the requirements of the Plant Preservation Rust Grading and Budget Preparation Guide. Specifically, prior to February 28, 2013, licensee personnel failed to initiate nuclear notifications for plant areas that received a rust grade of 4 or higher. This issue has been entered into the licensees corrective action program as Nuclear Notification NN 202341172. The inspectors determined that the failure to initiate nuclear notifications for the areas assigned a rust grade of 4 as required by the Plant Preservation Rust Grading and Budget Preparation Guide was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the Initiating Events Cornerstone attribute of equipment performance and adversely affected the associated cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors determined that Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, was appropriate based on the plant conditions present when most of the examples of this performance deficiency occurred. The finding did not require a quantitative assessment because adequate mitigating equipment remained available and the finding did not constitute a loss of control, as defined in Appendix G. Therefore, the finding screened as having very low safety significance (Green). This finding has a crosscutting aspect in the area of problem identification and resolution, corrective action program component, because the licensee failed to implement a corrective action program with a low threshold for identifying issues |
| 05000361/FIN-2013002-03 | 31 March 2013 23:59:59 | San Onofre | NRC identified | Failure to Write Nuclear Notifications for Degraded or Nonconforming Conditions | The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure of the licensee to implement procedures associated with entry of degraded or nonconforming issues into the corrective action program. Specifically, the NRC staff identified seven examples of problems that were not documented in a nuclear notification until prompted by NRC many days or years after they were known to the licensee between June 2009 and January 2013. This issue has been entered into licensees corrective action program as Nuclear Notification NN 202364842. The inspectors determined that the failure of licensee personnel to write nuclear notifications in accordance with Procedure SO123-XV-50.CAP-1 was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the Mitigating Systems Cornerstone attribute for equipment performance and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined that Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, was appropriate based on the plant conditions present when most of the examples of this performance deficiency occurred. The finding did not require a quantitative assessment because adequate mitigating equipment remained available and the finding did not constitute a loss of control, as defined in Appendix G. Therefore, the finding screened as having very low safety significance (Green). This finding had a crosscutting aspect in the area of the problem identification and resolution corrective action program component, because the licensee failed to implement a corrective action program with a low threshold for identifying issues |
| 05000361/FIN-2013002-05 | 31 March 2013 23:59:59 | San Onofre | NRC identified | Two Examples of Failure to Follow Procedures for Control of Maintenance | The inspectors identified a Green noncited violation of Technical Specification 5.5.1.1 for the failure by licensee personnel to follow Procedure SO23-XX-30, Nuclear Maintenance Order (NMO) Generation, Screening and Classification, Revision 9 EC1, and Procedure SO23-XX-36, Toolpouch Maintenance Program, Revision 1 EC1. Specifically, prior to March 5, 2013, the licensees Nuclear Maintenance Order Screening Committee failed to assign the appropriate job type and priority to seven corrosion-related nuclear maintenance orders in accordance with Procedure SO23-XX-30. Additionally, between February 9, 2012, and February 19, 2013, the Nuclear Maintenance Order Screening Committee failed to ensure the required conditions were met prior to assignment of toolpouch maintenance tasks for four nuclear notifications in accordance with Procedure SO23-XX-36. This issue has been entered into licensees corrective action program as Nuclear Notifications 202346546 and 202351959. The inspectors determined that the failure by the licensees personnel to follow Procedure SO23-XX-30 to assign the appropriate job types and priority for corrosion-related nuclear maintenance orders, and the failure to follow Procedure SO23-XX-36 for the conduct of toolpouch maintenance were performance deficiencies. These performance deficiencies were more than minor, and therefore a finding, because they were associated with the Initiating Events Cornerstone attribute of equipment performance and adversely affected the associated cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors determined that Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, was appropriate based on the plant conditions present when most of the examples of this performance deficiency occurred. The finding did not require a quantitative assessment because adequate mitigating equipment remained available and the finding did not constitute a loss of control, as defined in Appendix G. Therefore, the finding screened as having very low safety significance (Green). This finding had a crosscutting aspect in the area of human performance, decision-making component, because the Nuclear Maintenance Order Screening Committee failed to use conservative assumptions in decision making when assigning job types and tool pouch maintenance tasks for nuclear notifications |
| 05000361/FIN-2013403-01 | 31 March 2013 23:59:59 | San Onofre | NRC identified | Security | |
| 05000361/FIN-2012005-01 | 31 December 2012 23:59:59 | San Onofre | NRC identified | Failure to Provide Complete and Accurate Information Regarding Auxiliary Feedwater System Operability | The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.9, Completeness and Accuracy of Information, for the failure of the licensee to provide complete and accurate information in all material respects in operability and reportability review supporting documents. Specifically, on September 29, 2011, the licensee did not provide information that was complete and accurate in all material respects, in that Evaluation Report FAI/11-0655, Evaluation of Potential Cooling of the SONGS Steam Line for the AFW Turbine, used inaccurate information to inappropriately determine that the turbine-driven auxiliary feedwater pump was operable, the condition was not reportable per the requirements of 10 CFR 50.73, and the compensatory measures implemented on May 5, 2011, could be removed. The compensatory measures were improperly removed on October 27, 2011. This violation has been entered into the licensees corrective action program as Nuclear Notification NN 202280026. The failure of the licensee to provide complete and accurate information related to the operability of the AFW system was a performance deficiency. The significance determination process is not suited to assess the significance of a violation of 10 CFR 50.9 because it affected the ability of the NRC to perform its regulatory oversight function and, as such, it was assessed using traditional enforcement. This violation was determined to be a Severity Level IV violation based on NRC Enforcement Policy examples provided in Section 6.9. No crosscutting aspect was assigned because the performance deficiency was assessed using traditional enforcement |
| 05000361/FIN-2012404-01 | 30 September 2012 23:59:59 | San Onofre | Self-revealing | Security | |
| 05000361/FIN-2012004-02 | 30 September 2012 23:59:59 | San Onofre | NRC identified | Failure to Update the Final Safety Analysis Report for Solid Radioactive Waste | The inspectors identified a Severity Level IV, non-cited violation of 10 CFR 50.71, Maintenance of Records, Making of Reports, paragraph (e) which states, in part, 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. Contrary to the above, from 1985 to June 2012, the licensee failed to update the Final Safety Analysis Report to assure that the information included in the report contains the latest information developed. Specifically, since its construction in 1985, the licensee stored a significant source of radioactivity in the Multi-Purpose Handling Facility (South Yard Storage Facility), but failed to describe the source, volume, and storage of radioactive equipment in the Final Safety Analysis Report. The licensee has entered this violation into their corrective action program as Nuclear Notification NN 202076593. The inspectors determined that the failure to update the Final Safety Analysis Report as required by 10 CFR 50.71(e), Maintenance of Records, Making of Reports is a performance deficiency. This performance deficiency was dispositioned using traditional enforcement because failing to update a Final Safety Analysis Report had the potential to adversely impact the NRCs ability to perform its regulatory function. The performance deficiency is characterized as a Severity Level IV violation in accordance with the NRC Enforcement Policy, Section 6.1.d.3. Since this issue was dispositioned using traditional enforcement, there is no cross-cutting aspect |
| 05000361/FIN-2012004-03 | 30 September 2012 23:59:59 | San Onofre | Licensee-identified | Licensee-Identified Violation | The licensee identified a Green non-cited violation of Technical Specification 5.5.2.3(c), in part, for failure to monitor and sample radioactive liquid effluents with the methodology in accordance with the offsite dose calculation manual. Specifically, the offsite dose calculation manual, Section 4.1.1 and Table 4-1, address the requirements for compensatory sampling when the turbine plant sump sample compositor is inoperable. Contrary to these requirements, on February 1 and 2, 2012, the licensee failed to restart the compositor and obtain samples for the weekly composite sample as required. The finding was more than minor because it adversely affected the Public Radiation Safety Cornerstone with the potential for an unmonitored release of radioactive materials via liquid effluents. The licensee entered this issue into their corrective action program as Nuclear Notification NN 201841920. |
| 05000362/FIN-2012010-01 | 30 September 2012 23:59:59 | San Onofre | NRC identified | Inadequate Trip/Transient and Event Review Procedure | The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, because Operations Procedure SO123-0-A8, "Trip/Transient and Event Review," Revision 5, was not adequate in that it did not define what unplanned reactor trip meant, and the operators did not complete the procedure as required. In response, the licensee revised the procedure to describe unplanned reactor trips as explained in industry guidance. This issue was entered into the licensees corrective action program as Nuclear Notifications NN 201915602 and NN 202161665. This finding is more than minor because if left uncorrected the performance deficiency could be viewed as a precursor to a significant event. Using Inspection Manual Chapter 0609.04, Phase 1 Initial Screening and Characterization of Findings, and Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, the inspectors determined the finding to be of very low safety significance because a reactor trip was initiated with no loss of mitigating equipment or other associated initiators. This finding does not have a cross-cutting aspect because the associated procedure change occurred in 2003 and was not representative of current performance. |
| 05000362/FIN-2012010-02 | 30 September 2012 23:59:59 | San Onofre | NRC identified | Failure to Comply with Requirements for Handling, Storage, and Shipping | The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion XIII, "Handling, Storage, and Shipping," involving the licensees failure to take appropriate measures to control preservation of safety-related equipment during shipping, specifically the protective environment provided for the Unit 3 steam replacement generators was not appropriately specified or monitored. The licensee conducted an analysis of the shipping environment and determined no detrimental impact occurred. This issue was entered into the licensees corrective action program as Nuclear Notifications NN 202160749, NN 201960027, and NN 20191118. The finding is more than minor because it is associated with initiating events cornerstone attribute of equipment performance and affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Using Inspection Manual Chapter 0609.04, Phase 1 Initial Screening and Characterization of Findings, and Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, the inspectors determined the finding to be of very low safety significance because it did not result in exceeding the reactor coolant system leak rate for a small loss of coolant accident or result in a total loss of systems used to mitigate a loss of coolant accident. This finding does not have a cross-cutting aspect because the associated performance deficiency was not representative of current performance. |
| 05000361/FIN-2012004-01 | 30 September 2012 23:59:59 | San Onofre | NRC identified | Failure to Correct Drill Performance Weaknesses | The inspectors identified a non-cited violation of 10 CFR 50.47(b)(14) for failure to correct weaknesses or deficiencies that are identified in formal critiques of drills or exercises. The licensee did not take corrective actions for fourteen weaknesses in site assembly and evacuation, tracking of non-licensed operators, and provision of radiation protection to non-licensed operators, identified in critiques between September 2010 and June 2012. The failure to correct weaknesses identified in drills and exercises was a performance deficiency within the licensees control. This failure has been entered into the licensees corrective action program as Nuclear Notifications NNs 201974817, 201811829, and 201645589 This finding is more than minor because it affected the emergency response organization cornerstone attribute. The finding was evaluated using the Emergency Preparedness Significance Determination Process and determined to be of very low safety significance because it was a failure to comply and was not a loss of planning standard function. The finding was not a loss of planning standard function because the weaknesess that were not corrected were not associated with risk significant planning standards. This finding was assigned a corrective action cross-cutting aspect because San Onofre did not take corrective actions for numerous drill weaknesses in a timely manner commensurate with their safety significance |
| 05000361/FIN-2012010-03 | 30 September 2012 23:59:59 | San Onofre | Licensee-identified | Licensee-Identified Violation | Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that design control measures shall be established to 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. Contrary to the above, on December 11, 2007, and July 3, 2008, the licensee failed to establish design control measures during the review of Mitsubishis design Calculations SO23-617-1-C749 and SO23-617-1-C157, respectively, to verify or check the adequacy of the retainer bars design with respect to the susceptibility of the smaller diameter retainer bars to flow-induced vibration. Specifically, the licensee failed to ensure that there was sufficient analytical effort in the design methodology of the anti-vibration bar assembly to support the conclusion that tube wear would not occur as a result of contact with the retainer bars due to flow-induced vibration. Consequently, the smaller diameter retainer bars vibrated during normal operation causing wear on the adjacent tubes, that challenged the integrity of the reactor coolant system boundary. The inspectors determined that the licensees failure to verify the adequacy of the retainer bar design as required by Procedure SO123-XXIV-37.8.26 was of very low safety significance (Green) based on Inspection Manual Chapter 0609.04, Phase 1 Initial Screening and Characterization of Findings, and Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, because the finding did not involve a degraded steam generator tube condition where one tube could not sustain 3 times the differential pressure across a tube during normal full power, steady state operation and none of the replacement steam generators violated the accident leakage performance criterion in plant Technical Specifications as a result of the retainer bar vibration. The licensee also implemented actions to inspect all affected tubes in Unit 2 and 3 and remove from service all those tubes surrounding the smaller retainer bars that could wear due to vibration of the retainer bar. Because this violation has been determined to be of very low safety significance (Green) and has been entered in the licensees corrective action program as Nuclear Notification NN 201843216, it will be dispositioned as a non-cited violation in accordance with Section 2.3.2 of the NRCs Enforcement Policy. |
| 05000361/FIN-2012003-04 | 30 June 2012 23:59:59 | San Onofre | NRC identified | Failure to Follow Design Control Procedures | The inspectors identified a non-cited violation 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure of engineering personnel to follow Procedure SO123-XXIV-10.1, Engineering Design Change Process NECPs, Revision 28, to change the design, through physical plant modifications, of a facility used to handle radioactive material. Specifically, on February 2, 2012, engineering personnel issued as-built engineering change package NECP 800841701 which physically modified the design of the fuel reconstitution gantry crane with no turnover when an issued for construction engineering change package with turnover was required. This issue was entered into the licensees corrective action program as Nuclear Notification NN 202026584. The performance deficiency is more than minor, and therefore a finding, because it would become a more significant safety concern if left uncorrected since handling fuel with improperly modified equipment could result in fuel barrier damage. This finding cannot be evaluated by the significance determination process because Manual Chapter 0609, Significance Determination Process, Appendix A, Significance Determination of Reactor Inspection Findings for At- Power Situations, and Appendix G, Shut-down Operations Significance Determination Process, do not apply to the spent fuel pool. This finding affects the Barrier Integrity Cornerstone and is determined to be of very low safety significance by NRC management review because it was a deficiency that did not result in the actual degradation of spent fuel. This finding has a cross-cutting aspect in the area of human performance associated with the work practices component because the expectations regarding procedural compliance, and that personnel follow procedures were not effectively communicated to Design Engineering, Nuclear Fuels Management, and Project Management Organization personnel |
| 05000361/FIN-2012007-05 | 30 June 2012 23:59:59 | San Onofre | NRC identified | Shipping Requirements Not in Accordance with Industry Standards | The team identified an unresolved item associated with Unit 3 steam generators not shipped in accordance with specification SO23-617-01, Design and Fabrication of Replacement Steam Generators for Unit 2 and Unit 3, Revision 4, and requirements for handling, storage, and shipping. Specifically, ANSI N45.2-1977, Quality Assurance Program Requirements for Nuclear Facilities, required a special protective environment for handling, storage, and shipping of the replacement steam generators. However, because of schedule changes, the Unit 3 protective environment which included maintaining a nitrogen pressure and a monitoring plan was altered significantly. The team evaluated specifications associated with the shipping and handling of the Unit 2 and 3 replacement steam generators. Based on the information evaluated by the team, the steam generators procurement and shipping specifications required monitoring and maintenance of nitrogen atmosphere inside the replacement steam generators during shipment. Supplier Deviation Request SDR 10041870-09091 dated December 1, 2009, documents a request not to control the positive pressure, the dew point of nitrogen, and the oxygen content on the primary and secondary side of the Unit 3 replacement steam generators to accelerate delivery schedule. Specification SO23-617-01, Section 3.16.3, specifies the supplier shall be responsible for monitoring and maintaining nitrogen atmosphere inside the steam generators during their shipping from Mitsubishi to the California port discharge point. The team noted that Unit 3 steam generators did not require, monitoring or control of dew point, oxygen concentration, inside nitrogen pressure. The team could not identify if this was properly evaluated (Reference Section 5 of shipping and handling procedure SO23-617-1-M1350). Additional review and follow up will be required to review the evaluations and corrective actions associated with the maintaining the Unit 3 replacement steam generators protective environment during shipping and then determine whether this issue represents a performance deficiency or constitutes a violation of NRC requirements. |
| 05000361/FIN-2012007-04 | 30 June 2012 23:59:59 | San Onofre | NRC identified | Evaluation of Changes in Dimensional Controls During the Fabrication of Unit 2 and Unit 3 Replacement Steam Generators | The team identified an unresolved item associated with the dimensional controls of critical dimensions throughout the fabrication of Unit 2 and Unit 3 replacement steam generators. Based on the information gathered by the team on the differences in dimensional controls of critical parameters in Unit 2 and Unit 3 replacement steam generators, the team determined that Mitsubishi did not consider the potential impact of improving dimensional controls for tube roundness and anti-vibration bars on the final tube bundle clearances at normal operating conditions. Additional review by the NRC is required following completion of Mitsubishis cause evaluation to fully assess how the dimensional controls contributed to the tube-to-tube wear in Unit 3 and then determine whether this issue represents a performance deficiency or constitutes a violation of NRC requirements. |
| 05000361/FIN-2012007-03 | 30 June 2012 23:59:59 | San Onofre | NRC identified | Evaluation of Retainer Bars Vibration During the Original Design of the Replacement Steam Generators | The team identified an unresolved item associated with the design of the retainer bars in Unit 2 and Unit 3 replacement steam generators. In February 2012, the licensee identified wear indications in Unit 2 replacement steam generators at the tube locations in contact with the retainer bars. Some of the indications showed excessive wear with a maximum degradation of 90 percent through wall. The team identified that the design of the replacement steam generators did not expect any potential vibration concerns in the area of the tube bundle where the retainer bars were located. The basis for Mitsubishis design philosophy relied on the following factors: Based on the calculated natural frequency of the retainer bar, Mitsubishi considered that there would not be a resonant vibration condition relative to the flow conditions in the location of retainer bars. The vibration analysis of the tube bundle only considered out-of-plane vibration because in-plane vibration was not expected to be an operational concern for the retainer bars. The outermost tubes were considered the least susceptible to flow-elastic instability; therefore retainer bar locations were not included in the vibration analysis. Fluid-elastic instability was found not applicable to the retainer bar because this mechanism did not apply to a single tube in cross flow. Vortex-induced vibration was found not applicable to the retainer bar because it was considered a vibration mode applicable to a single cylinder in uniform cross flow in a large area and the flow condition around the retainer bars was considered slug-froth two phase flow. However, upon identification of retainer bar-to-tube wear in Unit 2 replacement steam generators, Mitsubishi performed an evaluation to identify the cause of excessive wear. The analysis considered three vibration mechanisms: fluid-elastic instability, vortex-induced vibration, and turbulence-induced vibration (random vibration). The analysis for turbulence-induced vibration determined that random vibration was the possible cause of the retainer vibration, based on the peculiar flow around the retainer bar, combined with the rather low natural frequency of the retainer bar. The analysis used the two phase flow conditions around the retainer bars and identified various modes of vibration at those flow conditions that could lead to retainer bar vibration and consequently to tube wear. Additional review by the NRC is required following completion of the Mitsubishis cause evaluation to determine whether this issue represents a performance deficiency or constitutes a violation of NRC requirements. |
| 05000361/FIN-2012003-06 | 30 June 2012 23:59:59 | San Onofre | Licensee-identified | Licensee-Identified Violation | The following finding of very low safety significance (Green) was identified by the licensee and is a violation of NRC requirements which meets the criteria of Section 2.3.2 of the NRC Enforcement Policy, NUREG-1600, for being dispositioned as a non-cited violation. License Condition 2.C (14), Fire Protection, requires the licensee to implement and maintain in effect all provisions of the approved fire protection program. The approved fire protection program requires the licensee to meet the requirements of 10 CFR Part 50, Appendix R, Section III.G.3, which requires that alternative or dedicated shutdown capability and its associated circuits, independent of cables, systems or components in the area, room, or zone under consideration, should be provided where the protection of systems whose function is required for hot shut-down does not satisfy the requirement of 10 CFR Part 50, Appendix R, Section III.G.2. Contrary to the above, from original installation in 1981 to June 8, 2011, the licensee failed to provide an alternative shutdown capability that was independent of cables, systems, or components in alternative shutdown areas. Specifically, the licensee mis-wired a switch needed for fire isolation for the Unit 3 train A charging pump motor control circuitry. In this condition, a hot short on the indication portion of the Unit 3 Train A charging pump motor control circuitry would energize the trip function causing a loss of control to the credited charging pump. This issue was determined to have very low safety significance by a senior reactor analyst because of the low probability of an Anticipated Transient Without Scram (ATWS) and fire-induced Loss of Coolant Accident (LOCA) event, the limited accident mitigating function of credited charging pumps, and the availability of High Pressure Safety Injection (HPSI) to mitigate a small loss of coolant accident. This issue was identified in the licensees corrective action program as Nuclear Notification NN 201456724. |
| 05000361/FIN-2012007-10 | 30 June 2012 23:59:59 | San Onofre | NRC identified | Evaluation of Departure of Method of Evaluation for 10 CFR 50.59 Processes | The NRR technical specialist reviewed SCEs 10 CFR 50.59 evaluation contained in Engineering Change Packages 800071702 and 800071703 for the Unit 2 and Unit 3 replacement steam generators, respectively, in which SCE determined that the impact of the replacement steam generators on the current licensing basis and any need for NRC approval as required by 10 CFR 50.59. The NRR technical specialist reviewed the SCEs 10 CFR 50.59 evaluation against 10 CFR 50.59(c)(2)(viii) which requires that licensees obtain a license amendment pursuant to 10 CFR 50.90 prior to implementing a proposed change if the change would result in a departure from a method of evaluation described in the final safety analysis report (as updated) used in establishing the design bases or in the safety analyses. Industry guidance NEI 96-07, Revision 1, Section 3.10, Methods of Evaluation, states, Definition: Methods of evaluation means the calculational framework used for evaluating behavior or response of the facility or structures, systems, and components. Regulation 10 CFR 50.59 a(2), states, Departure from a method of evaluation described in the FSAR (as updated) used in establishing the design bases or in the safety analyses means (i) changing any of the elements of the method described in the FSAR (as updated) unless the results of the analysis are conservative or essentially the same; or (ii) changing from a method described in the FSAR to another method unless that method has been approved by NRC for the intended application. Regulation 10 CFR 50.59(d)(1) requires that the licensee maintain records of changes in the facility that includes a written evaluation which provides the bases for the determination that the change, test, or experiment does not require a license amendment.... The technical specialist evaluated SCEs bases for determining that the changes would not result in the departure from the method of evaluation used in establishing the design bases or in the safety analyses. Specifically, the technical specialist evaluated whether the changes involved: (a) changing of any of the elements of the method described in the updated final safety analysis report, which consistent with 10 CFR 50.59 a(2)(i) would be justified by demonstrating that the results of the analysis are conservative or essentially the same; or (b) changing from a method described in the updated final safety analysis report to another method, which consistent with 10 CFR 50.59 a(2)(ii) would be justified by demonstrating that method has been approved by NRC for the intended application. The NRR technical specialist reviewed SCEs 10 CFR 50.59 evaluation and found two instances that failed to adequately address whether the change involved a departure of the method of evaluation described in the updated final safety analysis report. (a) Use of ABAQUS instead of ANSYS: Updated Final Safety Analysis Report Sections 3.9.1.2.2.1.11 and 3.9.1.2.2.2.3 were revised to reflect that the SONGS Unit 2 and 3 original steam generators stress analyses for reactor coolant system structural integrity utilized the ANSYS computer program, whereas the replacement steam generators analyses utilized the ABAQUS computer program. The SCEs 50.59 evaluation incorrectly determined that using the ABAQUS instead of ANSYS was a change to an element of the method described in the updated final safety analysis report did not constitute changing from a method described in the updated final safety analysis report to another method, and as such, did not mention whether ABAQUS has been approved by the NRC for this application. (b) Use of ANSYS instead of STRUDL and ANSYS: Updated Final Safety Analysis Report Section 5.4.2.3.1.3 was revised to reflect that the SONGS Unit 2 and 3 evaluation of tube stress under loss of coolant accident conditions for the original steam generators consisted of a two-step process utilizing the STRUDL and ANSYS computer programs to calculate displacement histories and tube stresses, respectively, while the corresponding replacement steam generators analysis determined tube stresses from blow-down forces using only the ANSYS computer program. While SCEs 50.59 evaluation correctly considered this a change from a method described in the FSAR to another method, the 50.59 evaluation did not mention whether the method has been approved by NRC for this application. |
| 05000361/FIN-2012003-03 | 30 June 2012 23:59:59 | San Onofre | NRC identified | Failure to Maintain Foreign Material Exclusion Controls During Maintenance | The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure of maintenance personnel to implement procedures associated with foreign material exclusion controls while performing maintenance activities on safety-related 120Vac inverter equipment. Specifically, on June 8, 2012, maintenance personnel failed to follow Procedure SO123-FO-1, Site Foreign Material - 4 - Exclusion Control Program, Revision 6, and Procedure SO123-I-1.18, Foreign Material Exclusion (FME) Control, Revision 18, when maintenance personnel failed to implement adequate foreign material exclusions controls during inverter 2Y004 troubleshooting activities. This issue was entered into the licensees corrective action program as Nuclear Notification NN 202016714. The performance deficiency is more than minor, and therefore a finding, because it is associated with the Mitigating Systems Cornerstone attribute for human performance and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, maintenance personnel failed to implement adequate controls, as required, to prevent the introduction of foreign materials during troubleshooting and repair activities associated with electrical cabinet of inverter 2Y004. Using Checklist 4 from the Manual Chapter 0609, Appendix G, Shut-down Operations Significance Determination Process, Phase 1 guidance, the finding is determined to have very low safety significance because all safety function guidelines were met, and thus, the finding did not require a quantitative assessment. This finding has a cross-cutting aspect in the area of human performance associated with the work practices component because the expectations regarding procedural compliance, and that personnel follow procedures were not effectively communicated to maintenance personnel regarding foreign material exclusion controls for unattended and opened electrical components |
| 05000361/FIN-2012403-01 | 30 June 2012 23:59:59 | San Onofre | NRC identified | Security | |
| 05000361/FIN-2012003-02 | 30 June 2012 23:59:59 | San Onofre | Self-revealing | Failure to Maintain Foreign Material Exclusion Controls in SAFETY-RELATED Components | The inspectors reviewed a self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure of maintenance personnel to implement procedures associated with foreign material exclusion controls while performing maintenance activities on safety-related 120Vac inverter equipment. Specifically, between October 2009 and April 2012, maintenance personnel failed to follow Procedure SO123-FO-1, Site Foreign Material Exclusion Control Program, Revision 6, and Procedure SO123-I-1.18, Foreign Material Exclusion (FME) Control, Revision 18, to prevent the introduction of a metal air filter frame that was left inside an energized electrical cabinet. This issue was entered into the licensees corrective action program as Nuclear Notification NN 201958287. The performance deficiency is more than minor, and therefore a finding, because it is associated with the Mitigating Systems Cornerstone attribute for human performance and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, maintenance personnel failed to prevent the introduction of foreign material into the energized electrical cabinet of inverter 2Y004. The resident inspectors performed the initial significance determination for the inverter finding. The inspectors used the NRC Inspection Manual 0609, Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings. The finding screened to a Phase 2 significance determination because it involved a potential loss of safety function. A Region IV senior reactor analyst performed a Phase 2 significance determination and attempted to use the pre-solved worksheet from the Risk Informed Inspection Notebook for the San Onofre Nuclear Generating Station, Revision 2.01a. However, the pre-solved worksheet did not include this inverter. Therefore, the analyst performed a bounding Phase 3 significance determination. The bounding change to the core damage frequency was 1.1E-8/yr, and therefore, determined to be of very low safety significance. The small population of affected equipment (included in the probabilistic risk assessment model) helped to minimize the safety significance. The contributing core damage sequences involved a seismic event and a consequential failure of auxiliary feedwater flow control and bypass valves to one steam generator. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because maintenance personnel failed to have an appropriate threshold for identifying issues associated with a degraded air filter and its impact to foreign material exclusion controls to ensure there would be no adverse impact to system operability |
| 05000361/FIN-2012007-01 | 30 June 2012 23:59:59 | San Onofre | NRC identified | Adequacy of the Trip/Transient and Event Review Procedure | The team identified an unresolved item associated with Operations Procedure SO123-0-A8, Trip/Transient and Event Review, that required a formal review of operator actions and safety systems to determine if important systems responded as design. The formal review was not completed. Description: On March 19, 2012, the team requested to review the results of operations post trip/transient evaluation of the January 31, Unit 3 tube leak event. Operations Procedure SO123-0-A8, Trip/Transient and Event Review, Revision 8, required a detailed post trip review following unplanned reactor trips. However, a formal trip/transient and event review was not available because operations personnel determined the Unit 3 event was planned and therefore a formal review was not required. On March 27, 2012, Team met with operations personnel to discuss items on a draft Unit 3 trip/transient evaluation provided to the team. The team also discussed with operations personnel the requirements in Operations Procedure SO123-0-A8, and concluded a basis for what was planned and unplanned was not defined. Operations personnel determined the Unit 3 reactor trip was a planned reactor trip because Abnormal Operating Instruction SO23-13-14, Reactor Coolant Leak, Revision 16, Section 4, had described actions for primary-to-secondary leakage. Specifically, this section stated, in part, under plant conditions of increasing steam generator tube leakage, operations personnel were required to perform a rapid power reduction to less than 35 percent power, then trip the reactor. The team discussed with operations personnel that definitions for unplanned events have been established through industry standards to report on plant performance. These standards include NEI 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 6. Industry Guidance NEI 99-02 indicated that Unit 3 reactor trip should be considered unplanned since the reactor trip was required by an abnormal operating instruction and would count against the performance indicator for unplanned reactor trips. NRC Regulatory Issue Summary (RIS) 2000-08, Voluntary Submission of Performance Indicator Data, Revision 1, allows industry to use NEI 99-02 to report performance indicator data. Additional review and follow up will be required to review the corrective actions associated with the procedural guidance for an event review and then determine whether this issue represents a performance deficiency or constitutes a violation of NRC requirements. |
| 05000361/FIN-2012007-09 | 30 June 2012 23:59:59 | San Onofre | NRC identified | Evaluation of the Effects of Divider Plate Weld Repairs in Unit 3 Replacement Steam Generators | The team identified an Unresolved Item associated with the adequacy of evaluation and controls for the divider plate weld repairs. The cracking of the divider plate weld in both Unit 3 replacement steam generators required extensive repairs affecting the channel head, divider plate, and tube-sheet. Based on interviews with licensee personnel and the review of documentation for the repairs, the team determined that Mitsubishi did not perform a comprehensive evaluation to assess the impact of the divider plate repairs on the integrity of the tube bundle. The team determined that the areas listed below were not considered or evaluated in sufficient depth to identify the potential adverse effects of the planned weld repairs. Additional Rotations The repair activities for the Unit 3 steam generators required additional rotations of the steam generator assembly while these were oriented in the horizontal position. The repairs resulted in approximately 300 additional rotations in each steam generator, which could have affected the configuration of the tube bundle in terms of anti-vibration bar gaps or distortion. The team identified that Mitsubishi did not fully evaluate the impact of additional rotations on the configuration of the steam generators since rotation was considered a normal evolution in the fabrication process. Heat Input The repair process included extensive heat-adding activities such as grinding, flame cutting, and post-weld heat treatment. While these activities were performed in accordance with the construction code of record and an approved repair plan, they could have resulted in thermal expansion and unintended distortion of steam generator components. For example, the channel heads were removed using flame cutting and Mitsubishis evaluation for the impact of this activity was limited to the base material area in the vicinity of the cut, i.e. the heat affected zone. Mitsubishi did not fully assess the impact this activity could have on the overall configuration of the steam generator in terms of thermal expansion of the tubesheet or distortion. Dimensional Checks after Repair The team identified that Mitsubishi did not perform dimensional verifications (e.g. clearances) of the tube bundle or other secondary side dimensions after the repairs of the Unit 3 steam generators to confirm that critical dimensions were not affected by the repairs. Additional review by the NRC is required following completion of the licensees cause evaluation to fully assess how the repair activities affected the Unit 3 replacement steam generators and then determine whether this issue represents a performance deficiency or constitutes a violation of NRC requirements. |
| 05000361/FIN-2012007-02 | 30 June 2012 23:59:59 | San Onofre | NRC identified | Evaluation of Unit 3 Vibration and Loose Parts Monitoring System Alarms | The team identified an unresolved item associated with the number of valid vibration and loose parts alarms observed in Unit 3 steam generators compared to Unit 2 steam generators, during steady state conditions. During the review of operational differences between Unit 2 and 3 steam generators the team identified a significant difference in number of valid vibration and loose parts monitoring system alarms. The vibration and loose parts monitoring system was designed to provide continuous monitoring and conditioning of loose parts accelerometer signals. Two separate accelerometers were installed on each of the steam generators. The location of these instruments are on the steam generators lower supporting structures and provide acoustic information about loose parts impacts specifically on the reactor coolant or primary side of the steam generators. The vibration and loose parts monitoring system real time functions consist mainly of impact alarm validation of suspected loose part events and recording acoustic data. Long term vibration monitoring and loose part event trending were done by engineering personnel using recorded data. Unit 3 returned to service in February 2011, and the resident inspectors noted a number of nuclear notifications associated with Unit 3 steam generators vibration and loose parts monitoring alarms. On January 20, 2012, prior to the Unit 3 tube leak, engineering personnel also identified this trend and documented in Nuclear Notification NN 201818719 this problem and assigned an action to do further evaluation. On February 3, 2012, engineering personnel sent two sets of alarm signatures to Westinghouse, which contained impact data on alarms for time periods of steady state operation (i.e., no major temperature changes). Westinghouse engineering personnel concluded that the acoustic signals picked up by the accelerometers were valid and similar in nature to acoustic signatures caused by thermal movement of a steam generator expected during changes in thermal conditions, such as plant startup or shutdown. However the data obtained and analyzed had been taken during steady state operations. The team noted that Unit 2 steam generators did not receive the same number and type of alarms during a similar period of steady state operations. Engineering personnel also compared hot leg temperature changes linked to Unit 3 operations from February 18, 2011, to January 31, 2012, and confirmed about 30 valid alarms during this period were not associated with thermal transients. Additional review and follow up will be required of the vibration and loose parts monitoring system alarms, including evaluation and disposition of Unit 3 alarms and then determine whether this issue represents a performance deficiency or constitutes a violation of NRC requirements. |
| 05000361/FIN-2012003-05 | 30 June 2012 23:59:59 | San Onofre | NRC identified | Failure to Follow Battery Testing Procedure for Battery Exceeding 85% Service Life | The inspectors identified a finding for the failure to follow the battery testing procedure for non-class 1E batteries. Specifically, licensee personnel failed to implement the non-class 1E battery testing procedure, SO23-I-9.96, Non-1E Battery Bank Performance Test, Revision 5, for Unit 2 battery 2B011 during refuel R2C17 when the battery is greater than 85% of its expected service life. The licensee submitted and approved an outage scope change request to test the battery during the current outage, and generated Nuclear Notification NN 201997619, to determine when battery 2B011 testing can be performed during the outage. The issue was entered into the licensees corrective action program as Nuclear Notification NN 201994131. The performance deficiency is more than minor, and therefore a finding, because it was associated with the human performance attribute of the Mitigating Systems Cornerstone, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was determined to have very low safety significance (Green) since it did not meet any of the greater than green criteria in Table 4A of Manual Chapter 0609, Attachment 04. The finding had a crosscutting aspect in the human performance area, associated with the decision-making component because the licensee did not use conservative assumptions to demonstrate that the battery would maintain minimum capacity until the next refueling outage |
| 05000361/FIN-2012403-02 | 30 June 2012 23:59:59 | San Onofre | Self-revealing | Security | |
| 05000361/FIN-2012007-08 | 30 June 2012 23:59:59 | San Onofre | NRC identified | NON-CONSERVATIVE THERMAL-HYDRAULIC Model Results | The team identified an unresolved item associated with the adequacy of Mitsubishis FIT-III thermal-hydraulic code. The FIT-III code predicted non-conservative low velocity and low void fraction results which were used as inputs to the vibration code FIVATS. These non-conservative thermal-hydraulic results lead Mitsubishi to conclude that margins to instability were significantly larger than they actually were. Replacement steam generators were designed and manufactured in accordance with SONGS Design Specification SO23-617-1and ASME Section III, Rules for Construction of Nuclear Facility Components. The replacement steam generators had enhanced materials and maintenance. The tube bundle, comprised of 9727 u-tubes, is supported by a set of seven tube support plates which are maintained and spaced by a network of tie-rods. The ends of the u-tubes were welded onto the tube sheet lower face cladding and were full depth expanded in the tube sheet holes. The u-bends are supported by a set of 6 anti-vibration bars, having a maximum of 12 contact points, in the center of the bundle. For shorter tubes near the periphery, a fewer number of anti-vibration bars are present. One of the major enhancements of the replacement steam generators was the use of Alloy-690 tubing versus Alloy-600 for corrosion resistance. Alloy-690 has lower heat conductivity so, to achieve the same power, the heat transfer surface area must be increased by at least 10 percent. This required more tubes to be used in the replacement steam generators. The increased number of tubes resulted in a more tightly compacted tube bundle and elimination of the stay cylinder. The increase in the number of tubes could lead to increases in primary reactor coolant flow through the steam generators. Orifices were machined as part of the steam generator inlet nozzles to ensure maximum allowed primary system flow-rates were not exceeded. The tube layout indexing or incrementation used in these generators was smaller than other replacement steam generator designs. The tighter indexing results in smaller pitch/diameter ratio in critical regions of the tube bundle u-bends. In addition, it was noted that the anti-vibration bar support structure is not connected to the wrapper for lateral or vertical support; instead the anti-vibration bar system structure is only supported vertically by resting on the tubes. Other operational and physical comparisons of the replacement steam generators and original steam generators were reviewed by the team and no significant differences were noted. Additional review by the NRC is required to fully assess the adequacy of Mitsubishis FIT-III thermal-hydraulic code and then determine whether this issue represents a performance deficiency or constitutes a violation of NRC requirements. |
| 05000361/FIN-2012007-06 | 30 June 2012 23:59:59 | San Onofre | NRC identified | Shipping Requirements Not in Accordance with Design and Fabrication Specifications | The team identified an unresolved item associated with the shipping and handling specifications requiring methods of tube bundle support. The team could not determine if this requirement to provide a tube bundle support method was adequately evaluated by SCE. Based on the information gathered by the team on shipping and handling specifications associated with the Unit 2 and 3 replacement steam generators, the team could not determine that Mitsubishi or SCE adequately considered the potential impact of not providing methods of tube bundle supports as required in Specification SO23-617-01. In response to the team questions regarding tube bundle support methods, the team was provided with results from Procedure L5-04GA069, Sagging Measurement Procedure, Revision 7. However the team noted the procedure is considered a non-quality affecting procedure and used for reference only. Additional review and follow up will be required to review the evaluations associated with the requirements to provide tube bundle support during shipping for the Unit 2 and 3 steam generators and then determine whether this issue represents a performance deficiency or constitutes a violation of NRC requirements. |
| 05000361/FIN-2012007-07 | 30 June 2012 23:59:59 | San Onofre | NRC identified | Unit 3 Steam Generator 3E0-88 Stresses Related to Handling | The team identified an unresolved item associated with evaluation of excessive shipping induced forces of Unit 3 replacement steam generator 3E-088. The team reviewed the SG shipping accelerometer data reports for both Unit 2 and Unit 3. In addition, the team also reviewed shipping and handling records and identified the following: Different transoceanic shipping companies and ships were used (U2: Happy Ranger, U3: Enchanter) During the discharge from the ship Unit 3 replacement steam generator 3E0-88 (3B) recorded simultaneous signals on the three attached accelerometers. Unit 3 steam generator 3E0-88 was the only steam generator to record simultaneous signals on the three attached accelerometers. Unit 3 steam generators received significantly more accelerometers hits compared to Unit 2.Unit 3 replacement steam generator 3E0-88 accelerometers indicated up to a 1.23 g spike with a simultaneous recording on all three of the attached accelerometers. Mitsubishi provided an evaluation of the forces which showed loads were within allowable stress limits but exceeded stress for an operating basis earthquake. The team was not able to determine if this was properly considered. Additional review by the NRC is required to fully assess if the shipping forces contributed to the tube-to-tube wear in Unit 3 and then determine whether this issue represents a performance deficiency or constitutes a violation of NRC requirements. |
| 05000361/FIN-2012003-01 | 30 June 2012 23:59:59 | San Onofre | NRC identified | Degraded Fire Barrier Separating Unit 2 Shutdown Cooling Heat Exchanger Rooms | The inspectors identified a non-cited violation of License Condition 2.C.(14) and the Updated Fire Hazards Analysis for the failure of the licensee to maintain the 3-hour penetration fire seal that separated redundant post-fire safe shutdown equipment. Specifically, prior to May 25, 2012, the licensee failed to maintain the 3 hour fire barrier between fire areas 2-SE-(-15)-138 and 2-SE-(- 15)-139. The issue was entered into the licensees corrective action program as Nuclear Notification NN 202003184. The performance deficiency is more than minor, and therefore a finding, because it was associated with the external factors attribute (i.e. fire) of the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the availability and reliability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, Phase 1 Initial Screening and Characterization of Findings, the finding was determined to require additional evaluation under Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process. Using Inspection Manual Chapter 0609, Appendix F, Attachment 2, Table A2.2, the inspectors concluded the penetration fire seal represented a moderate A degradation of the fire confinement element of the fire protection program. Using the supplemental screening for fire confinement findings, the inspectors concluded that the finding was of very low safety significance (Green) because the degraded penetration fire seal provided a minimum of 20 minutes of fire protection and no fire ignition sources or combustible materials would have caused direct flame impingement on the fire barrier. This finding has a cross-cutting aspect in the area of human performance associated with the resources component because the licensee failed to ensure that personnel were adequately trained to inspect this type of penetration |
| 05000361/FIN-2012002-01 | 31 March 2012 23:59:59 | San Onofre | NRC identified | Failure to Maintain Seismic Controls in Safety-Related Areas | The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure of operations personnel to follow Procedure SO123-XV-1.20, Seismic Controls, Revision 4. Specifically, between March 2 and March 6, 2012, operations personnel failed to follow Procedure SO123-XV-1.20, and allowed tools and equipment in the vicinity of safety-related shutdown cooling components in the room for shutdown cooling heat exchanger train B that could have become an operability hazard during a seismic event. The issue was entered into licensees corrective action program as Nuclear Notifications NNs 201884141 and 201910392. The performance deficiency is more than minor, and therefore a finding, because it was associated with the Mitigating Systems Cornerstone attribute for protection against external events and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Manual Chapter 0609, Appendix M, Significance Determination Process Using Qualitative Criteria, was used since Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, does not specifically address the particular condition in cold shutdown, in which time to boil is greater than 2 hours. The management review was performed using the Manual Chapter 0609, Appendix G, Attachment 1, Phase 1 guidance, to establish a bounding analysis. Using the bounding analysis, the finding is determined to have very low safety significance because the finding did not represent a potential loss of both trains of the shutdown cooling system. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because operations and Project Management Organization personnel failed to have an appropriate threshold to identify that tools and equipment in the vicinity of safety-related shutdown cooling components needed to be addressed to ensure there would be no adverse impact to system operability |
| 05000361/FIN-2012405-01 | 31 March 2012 23:59:59 | San Onofre | Self-revealing | Security | |
| 05000361/FIN-2012002-02 | 31 March 2012 23:59:59 | San Onofre | NRC identified | Failure to Implement Timely Corrective Actions on Safety-Related Pumps | The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the failure of maintenance and engineering personnel to promptly correct a degraded condition associated with safety-related equipment. Specifically, since December 1988, the licensee failed to address long-term pump bearing oil leaks on safety-related component cooling water pumps, and deferred effective corrective actions with temporary gasket sealant. The licensee has issued work orders to install larger size O-rings and remove the sealant material from the outside of the bearing housing. The issue was entered into licensees corrective action program as Nuclear Notification NN 201840078. The performance deficiency is more than minor, and therefore a finding, because it is associated with the design control attribute of the Mitigating Systems Cornerstone and affects the associated cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, engineering personnel determined it was acceptable to use applications of gasket sealant to temporarily repair oil leaks, and delay permanent repairs on CCW pump bearing housings. Using the Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheets, the finding is determined to have very low safety significance because it was not a design or qualification deficiency confirmed not to result in loss of operability or functionality; did not result in a loss of system safety function; did not represent an actual loss of safety function of a single train for greater than its technical specification allowed outage time; was not an actual loss of safety function of one or more non-technical specification trains of equipment designated as risk significant per 10 CFR 50.65 for greater than 24 hours; and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding was determined not to have a cross-cutting aspect because it is not reflective of current performance. |
| 05000361/FIN-2012002-06 | 31 March 2012 23:59:59 | San Onofre | Licensee-identified | Licensee-Identified Violation | Technical Specification 3.8.9, Distributions Systems Operating, limiting condition for operation requires that AC and DC electrical power distribution systems shall be operable in Modes 1 through 4. If one AC electrical power distribution system is inoperable for greater than 8 hours, the required action is to place the unit in Mode 3 within 6 hours and Mode 5 within 36 hours. Contrary to the above, on October 4, 2010, operations personnel failed to comply with required action of Technical Specification 3.8.9 limiting condition for operation to restore train A Class 1E 4kV bus 2A04 (AC electrical power distribution system) to operable status or place the unit in Mode 3 within 6 hours and Mode 5 within 36 hours. A Phase 3 evaluation was performed by a senior reactor analyst since the finding screened as potentially risk-significant due to a seismic initiating event. Using the San Onofre SPAR model, the delta-CDF for Bus 3A04 being non-functional was 5.2E-7/yr. The exposure period for this finding was 10.7 days over the TS 3.8.1 AOT and 13.37 days over the TS 3.8.9 allowed outage time. Therefore, this finding was determined to have very low significance (Green). This issue was entered into licensees corrective action program as Nuclear Notification NN 201113611. |
| 05000361/FIN-2012002-07 | 31 March 2012 23:59:59 | San Onofre | Licensee-identified | Licensee-Identified Violation | The inspectors reviewed a Severity Level IV problem consisting of two non-cited violations committed by an instrumentation and control technician for attempting to readjust a potentiometer to its original position without proper documentation and failing to notify the control room of a plant status control error. Technical Specification 5.5.1.1.a requires procedures to be established, implemented, and maintained covering the applicable procedures recommended by Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Regulatory Guide 1.33, Quality Assurance Program Requirements (Operation), Appendix A, Typical Procedures for Pressurized Water Reactors and Boiling Water Reactors, Section 1.c, includes typical safety-related activities that should be accomplished in accordance with written procedures, such as equipment control. Procedure SO123-XV-15 , Maintaining Plant Status Control, Section 6.4.1 requires that plant manipulations are only made via an approved tracking document and Section 6.6.1 requires that the Shift Manager must be informed of any actual or suspected plant status control error. Contrary to the above, on March 28, 2011, plant manipulations were made without an approved tracking document and the Shift Manager was not informed of an actual plant status control error. Specifically, an Instrumentation and Control technician manipulated the Channel A potentiometer without an approved tracking document and failed to notify the Shift Manager of the plant status error. This was entered into the licensees corrective action program as NN 201393301. The licensee subsequently verified the operability of both Channels A and B and took actions to prevent potential cross-train errors for future instrumentation and control work. This violation is being treated as a non-cited violation in accordance with Section 2.3.2 of the NRC Enforcement Policy because the licensee identified the violation and promptly reported it to the NRC; it was an isolated action of an employee in a low level position without management involvement; it was not caused by a lack of management oversight; and, the licensee took appropriate remedial action commensurate with the circumstances. |
| 05000361/FIN-2012002-04 | 31 March 2012 23:59:59 | San Onofre | Licensee-identified | Licensee-Identified Violation | The inspectors reviewed a Severity Level IV non-cited violation committed by a radiography boundary guard for leaving his boundary post without approval. San Onofre Nuclear Generating Station Technical Specification 5.5.1.1.a requires procedures to be established, implemented, and maintained covering the applicable procedures recommended by Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Regulatory Guide 1.33, Quality Assurance Program Requirements (Operation), Appendix A, Typical Procedures for Pressurized Water Reactors and Boiling Water Reactors, Section 7.e, includes radiation protection procedures for access control to radiation areas. Procedure SO123-VII-20.10.7, Radiography Health Physics Controls, section 6.1.3.2 states that Radiography boundary guard duties are to guard the boundary and prevent personnel from crossing the posted Radiation Area boundary for radiography. Contrary to the above, on November 29, 2010, a radiography boundary guard did not guard the radiography boundary. Specifically, the radiography boundary guard left the radiography boundary post between radiographic shots without being properly relieved. This issue was documented in the licensees corrective action program as Nuclear Notification NN 201219666. This violation is being treated as a non-cited violation in accordance with Section 2.3.2 of the NRC Enforcement Policy because the licensee identified the violation and promptly reported it to the NRC, it was an isolated action of an employee in a low-level position without management involvement, it was not caused by a lack of management oversight, and the licensee took appropriate remedial action commensurate with the circumstances. |
| 05000361/FIN-2012002-05 | 31 March 2012 23:59:59 | San Onofre | Licensee-identified | Licensee-Identified Violation | Technical Specifications 3.8.1, AC Sources - Operating, limiting condition for operation requires that AC electrical sources shall be operable in Modes 1 through 4. Technical Specification 3.8.9, Distributions Systems Operating, limiting condition for operation requires that AC and DC electrical power distribution systems shall be operable in Modes 1 through 4. If one required offsite circuit is inoperable for greater than 72 hours, or one AC electrical power distribution system is inoperable for greater than 8 hours, the required action is to place the unit in Mode 3 within 6 hours and Mode 5 within 36 hours. Contrary to the above, on September 2, 2010, operations personnel failed to comply with required action of Technical Specification 3.8.9 limiting condition for operation to restore train A Class 1E 4kV bus 3A04 (AC electrical power distribution system) to operable status or place the unit in Mode 3 within 6 hours and Mode 5 within 36 hours; and on September 5, 2010, operations personnel failed to comply with required action of Technical Specification 3.8.1 limiting condition for operation to restore bus 3A04 feeder breaker from the reserve auxiliary transformer (AC electrical source) to operable status or place the unit in Mode 3 within 6 hours and Mode 5 within 36 hours. A Phase 3 evaluation was performed by a senior reactor analyst since the finding screened as potentially risk-significant due to a seismic initiating event. Using the San Onofre SPAR model, the delta-CDF for Bus 3A04 being non-functional was 5.2E-7/yr. The exposure period for this finding was 10.7 days over the TS 3.8.1 AOT and 13.37 days over the TS 3.8.9 allowed outage time. Therefore, this finding was determined to have very low significance (Green). This issue was entered into licensees corrective action program as Nuclear Notification NN 201113611. |
| 05000361/FIN-2012002-03 | 31 March 2012 23:59:59 | San Onofre | Self-revealing | Failure to Control Work Activities and Prevent RCS Perturbations | The inspectors reviewed a self-revealing non-cited violation of Technical Specification 5.5.1.1 for the failure of operations personnel to follow Procedure SO23-3-1.8, Draining the Reactor Coolant System to a Reduced Inventory Condition, Revision 32, Attachment 13, Reduced Inventory Condition RCS Perturbation Control. Specifically, on February 8, 2012, operations personnel failed to document potential reactor coolant system perturbations and the measures, controls, and enhanced monitoring used to prevent perturbations. Consequently, work activities performed by heath physics personnel were not appropriately documented and controlled which resulted in a reactor coolant system perturbation while in reduced inventory conditions. The issue was entered into licensees corrective action program as Nuclear Notification NN 201848706. The performance deficiency is more than minor, and therefore a finding, because it was associated with the Initiating Events Cornerstone attribute of configuration control and affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Additionally, the failure to appropriately control work activities that could impact reactor coolant system inventory while in reduced inventory conditions, if left uncorrected, would have the potential to lead to a more significant safety concern. Using the Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, Phase 1 guidance, a Phase 2 analysis is required because the finding increased the likelihood of a loss of reactor coolant system inventory during reduced inventory conditions as a result of inadequate controls implemented to avoid operations that could lead to perturbations in reactor coolant system level control. The finding was evaluated using the Phase 2 guidance in IMC 0609, |
| 05000361/FIN-2011005-01 | 31 December 2011 23:59:59 | San Onofre | Self-revealing | Failure to Control Work in a High Radiation Area | The inspectors reviewed a self-revealing non-cited violation of Technical Specification 5.8.1 for the failure to control work in a high radiation area. On August 25, 2011, diving was performed in a high radiation area using stay time calculations instead of the radiation protection coverage described in the Technical Specifications. The licensee suspended further diving operations until interim corrective actions were put in place. The licensee placed this issue into their corrective action program as Nuclear Notification NN 201620253. The failure to adequately control work in a high radiation area was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it negatively impacted the Occupational Radiation Safety cornerstone attribute of program and process and affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation, in that a worker received unplanned, unintended radiation dose. Using NRC Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, the finding was determined to be of very low safety significance because: (1) it was not associated with ALARA planning or work controls, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. This finding has a cross-cutting aspect in the area of human performance related to resources. Specifically, the licensee did not have a diving procedure to control this evolution (H.2.(c)) |