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05000275/FIN-2010002-06Licensee-Identified Violation2010Q1The following violation of very low safety significance (Green) was identified by the licensee and is a violation of NRC requirements which meets the criteria of Section VI of the NRC Enforcement Policy, NUREG-1600, for being dispositioned as a noncited violation. Title 10 CFR Part 50.73(a)(2)(v), required that the licensee submit a Licensee Event Report for safety system functional failures within 60 days. Contrary to this, the licensee failed to report the July 21, 2008 failure of both auxiliary building exhaust fans as a safety system functional failure within 60 days. On January 28, 2010, the licensee identified the failure to make the report during a review of past events. The licensee subsequently reported the condition as Licensee Event Report 05000323/2010-001-00, Diablo Canyon Power Plant Unit 2 Loss of Auxiliary Building Ventilation System Exhaust Fans. The inspectors concluded that the failure to make a required Licensee Event Report was a Severity Level IV violation using the General Statement of Policy and Procedure for NRC Enforcement Actions, Supplement I - Reactor Operations, dated January 14, 2005. Pacific Gas and Electric entered the issue into their corrective action program as Notifications 50265954 and 50301839
05000275/FIN-2010003-01Failure to Perform Unit 1 Containment Concrete Inspections2010Q2The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criteria XVI, Corrective Action, after Pacific Gas and Electric failed to implement prompt corrective actions after identifying a nonconservative technical specification. In December 2008, the inspectors identified that the diesel generator loading calculations were inadequate to demonstrate that the design basis were met. On January 9, 2009, the licensee entered this condition into the corrective action program. On April 9, 2009, Pacific Gas and Electric concluded that Technical Specification Surveillance Requirement 3.8.1, AC Sources Operating, was not adequate to preserve plant safety and applied the provisions of Technical Specification Surveillance Requirement 3.0.3, and Administrative Letter 98-10, Dispositioning of Technical Specifications that are Insufficient to Assure Plant Safety. The licensee did not complete the necessary actions to correct the deficient technical specification by submitting an adequate license amendment request. The inspectors concluded the most significant contributor to the finding was a less than adequate engineering evaluation to support the new emergency diesel generator loading profiles following the previous violation. The licensee entered the performance deficiency into the corrective action program as Notification 50232181. The inspectors determined that the performance deficiency is more than minor because if left uncorrected, the failure to implement prompt corrective actions has the potential to lead to a more significant safety concern. The inspectors concluded the finding was of very low safety significance because the finding was a design deficiency confirmed not to result in the loss of operability or functionality. The finding is associated with the Mitigating Systems Cornerstone. This finding had a crosscutting aspect in the area of problem identification and resolution, associated with the corrective action program component because the licensee failed to perform an adequate evaluation of the nonconservative technical specification such that the resolutions address causes and extent of conditions, as necessary.
05000275/FIN-2010003-02Inadequate Corrective Actions Following Identification of a Nonconservative Technical Specification2010Q2The inspectors identified a noncited violation of Technical Specification 5.5.16.a.1, Containment Leakage Rate Testing Program, after Pacific Gas and Electric failed to perform containment concrete inspections in accordance with the requirements of and frequency specified by ASME Section XI code, Subsection IWL. The licensee entered this into their corrective action program as Notification 50310054. The inspectors concluded that the failure of Pacific Gas and Electric to perform the technical specification required inspections is a performance deficiency. The finding is more than minor because the performance deficiency is associated with the Barrier Integrity Cornerstone human performance attribute and adversely affected the cornerstone objective to provide reasonable assurance that containment physical design barrier protects the public from radionuclide releases caused by accidents or events. The inspectors concluded that the finding is of very low safety significance because the performance deficiency did not represent a degradation of the radiological barrier function provided for the control room, auxiliary building, or spent fuel pool, did not represent a degradation of the barrier function of the control room against smoke or toxic atmosphere, did not represent an actual open pathway in the physical integrity of reactor containment, and did not involve an actual reduction in function of hydrogen igniters in the reactor containment. The inspectors did not assign a crosscutting aspect to this finding because the performance deficiency did not occur within the past three years and is not reflective of present performance.
05000275/FIN-2010003-03Failure to Report a Condition that Could Have Prevented the Fulfillment of a Safety Function2010Q2The inspectors identified a noncited violation of 10 CFR 50.73(a)(2)(i)(B) and 10 CFR 50.73(a)(2)(v)(B) and after Pacific Gas and Electric failed to submit a required licensee event report within 60 days following discovery of a condition prohibited by the plant technical specifications and a condition that could have prevented the fulfillment of a safety function. On March 9, 2010, Pacific Gas and Electric identified that the degraded voltage protection scheme, required by Technical Specification 3.3.5, Loss of Power Diesel Generator Start Instrumentation, was inadequate to protect operating engineering safety feature pump motors. The licensee concluded that sustained degraded voltage could result in an overcurrent condition affecting equipment powered from the preferred offsite power supply. This condition was required to be reported to the NRC because the degraded voltage protection scheme rendered engineered safety feature pumps inoperable for a period in excess of the allowable technical specification out of service time and the condition resulted in the loss of the degraded voltage protection scheme safety function on all three vital 4 kV power buses. The inspectors evaluated this finding using the traditional enforcement process because the failure to submit a required event report affected the NRCs ability to perform its regulatory function. The inspectors concluded the violation was a Severity Level IV because the licensee failed to submit an adequate licensee event report. The inspectors determined that the violation was also a finding under the reactor oversight process because licensee personnel failed to adequately evaluate a condition adverse to quality for operability and reportability, as required by station procedures. The inspectors concluded that the finding is more than minor because the failure to properly evaluate degraded plant equipment for past operability and reportability could reasonably be seen to lead to a more significant condition. The inspectors concluded that the finding had very low safety significance because the failure to adequately evaluate the condition did not result in an actual loss of a system safety function or equipment required by technical specifications, or involve the loss or degradation of equipment specifically designed to mitigate a seismic, flooding, or sever weather initiating event, and did not involve the total loss of any safety function that contributes to an external event initiated core damage accident sequence This finding has a crosscutting aspect in the area of problem identification and resolution, associated with the corrective action program component because the licensee failed to perform an adequate evaluation of the degraded voltage protection scheme such that the resolutions address causes and extent of conditions, as necessary.
05000275/FIN-2010003-04Licensee-Identified Violation2010Q2The 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 the NRC Enforcement Policy, NUREG-1600, for being dispositioned as a noncited violation. Title 10 of the Code of Federal Regulations, Part 50, Appendix B, Criteria V, Instructions, Procedures, and Drawings, required that activities affecting quality shall be prescribed by procedures, and shall be accomplished in accordance with these procedures. Procedure TS6.ID2, Control and Accountability of Special Nuclear Material, Revision 21, Attachment 7.15, Spent Fuel Pool Decay Heat Loading Requirements, specified restrictions for placement of recently discharged spent fuel assemblies in the spent fuel pool. Contrary to this, on April 27, 2010, Pacific Gas and Electric identified five recently discharged fuel assemblies improperly located in the spent fuel pool. The licensee took corrective action to restore the spent fuel pool configuration consistent with procedural requirements. The inspectors used Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, to analyze the finding. The inspectors concluded that the failure to accomplish an activity affecting quality in accordance with Procedure TS6.ID2, Attachment 7.15, was a Green finding of very low safety significance because the finding did not result in loss of cooling to the spent fuel pool, whereby operator or equipment failures could preclude restoration of cooling prior to boiling, did not result from fuel handling errors that caused damage to fuel clad integrity or a dropped assembly, and did not result in a loss of spent fuel pool inventory greater than 10 percent of the spent fuel pool volume. Pacific Gas and Electric entered the issue into the corrective action program as Notification 50314008
05000275/FIN-2010004-01Failure to Identify a Degraded Fire Barrier2010Q3The inspectors identified a noncited violation of the Diablo Canyon Facility Operating License Condition (5), Fire Protection, after Pacific Gas and Electric failed to maintain the integrity of a fire door in the rated configuration. On August 19, 2010, the inspectors identified that Fire Door 223 was inoperable. Fire Door 223 was required to provide a 3-hour rated barrier between Fire Areas 5-A-4 and 5-B-4. A fire in either of these areas could have prevented operation of the auxiliary feedwater, auxiliary saltwater, or component cooling water pumps or steam generator level control from the remote shutdown panel. Equipment Control Guideline 18.7, Fire Rated Assemblies, required the licensee to either maintain Fire Door 223 operable or implement compensatory actions within one hour. The inspectors concluded the most significant contributor to the finding was that licensee personnel did not identify and enter the degraded fire door into1 the Corrective Action Program. The licensee entered the performance deficiency associated with this finding into the corrective action program as Notification 50336901 and completed repairs to the door on August 23,2010. The inspectors concluded that the performance deficiency was more than minor because the degraded fire barrier affected the mitigating systems cornerstone external factors attribute objective to prevent undesirable consequences due to fire. The inspectors determined that the inoperable door was a fire confinement category finding and that the fire barrier was moderately degraded because the door would not perform the rated fire barrier function. The inspectors concluded the finding was of very low safety significance because the degraded barrier would have provided a minimum of 20 minutes fire endurance protection and ignition sources and combustible materials were positioned that had a fire spread to secondary combustibles, the degraded barrier would not have been subject to direct flame impingement. This finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee did not implement a low threshold for identifying and entering issues into the Corrective Action Program (P.1 (a)).
05000275/FIN-2010004-02Inadequate Risk Management during a Planned Auxiliary Saltwater Pump Outage2010Q3The inspectors identified a finding after Pacific Gas and Electric failed to adequately manage risk during planned maintenance activity as required by Procedure AD7.DC6, On-line Maintenance Risk Management. On April 5, 201 0, work control personnel requested that plant operators simultaneously remove Auxiliary Saltwater Pump 2-2 and Component Cooling Water Heat Exchanger 2-2 from service for two scheduled maintenance activities. Plant operators identified that the combination of the auxiliary saltwater pump and component cooling water heat exchanger out of service at the same time would result in an elevated maintenance risk (Yellow). Procedure AD7.DC6, On-line Maintenance Risk Management, Section 2.1, required that the licensee manage plant risk during on-line maintenance by minimizing the number of risk significant equipment simultaneously removed from service. The inspectors concluded that these two maintenance activities could have been performed in series rather than in parallel without affecting the duration either component was unavailable for maintenance. The licensee entered the performance deficiency into the corrective action program as Notification 50309451. The inspectors determined that the performance deficiency is more than minor because the performance deficiency affected the Mitigating Systems Cornerstone attribute of human performance and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Also, the finding is similar to Example 7.e in Inspection Manual Chapter 0612, Appendix E, Examples of Minor Issues, because the work scope unnecessarily placed the plant into a higher licensee-established risk category and required additional risk management actions. The inspectors concluded that the finding is of very low safety significance (Green) based on an actual incremental core damage probability deficit of less than 1X10-6 and an evaluation using Flowchart 1 of Appendix K of Inspection Manual Chapter 0609, Maintenance Risk Assessment and Risk Management Significance Determination Process. This finding had a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee failed to implement adequate corrective actions to prevent unnecessarily entering elevated plant risk for the planned maintenance (P.1 (d)).
05000275/FIN-2010004-03Inadequate Risk Assessment during Planned Maintenance Activities2010Q3The inspectors identified a noncited violation of 10 CFR 50.65 after Pacific Gas and Electric failed to perform a risk assessment after plant conditions had changed. On July 13, 201 0, Pacific Gas and Electric identified that station personnel failed to complete Technical Specification Surveillance Requirement 3.3.4.2, Remote Shutdown System, within the specified frequency for both Units. As provided by Surveillance Requirement 3.0.3, the licensee performed a risk evaluation to extend the required surveillance completion time beyond twenty-four hours. The licensee initiated the missed surveillance tests and identified results were outside acceptance criteria. On July 26, 2010, Operations personnel declared several remote shutdown system functions inoperable because reasonable expectation no longer existed that remote shutdown system could perform its safety function. Pacific Gas and Electric failed to reassess the effect on plant risk resulting from inoperable remote shutdown system functions before continuing with scheduled maintenance. A subsequent risk assessment concluded that plant risk was in a higher risk category due to planned maintenance activities conducted during this time frame. The licensee entered the performance deficiency into the corrective action program as Notification 501331 841. The inspectors determined that the performance deficiency is more than minor because the performance deficiency affected the Mitigating Systems Cornerstone attribute of human performance and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Also, the finding is similar to Example 7.e in Inspection Manual Chapter 0612, Appendix E, Examples of Minor Issues, because the overall elevated plant risk would put the plant into a higher licensee-established risk category. The inspectors concluded that the finding is of very low safety significance (Green) based on an actual incremental core damage probability deficit of less than 1x10-6 and an evaluation using Flowchart 1 of Appendix K of Inspection Manual Chapter 0609, Maintenance Risk Assessment and Risk Management Significance Determination Process. This finding had a crosscutting aspect in the area of human performance associated with the work practices component because the licensee failed to follow its maintenance risk procedure and reassess plant risk due to changing plant conditions (H.4(b)).
05000275/FIN-2010004-04Inadequate Operability Determination2010Q3The inspectors identified a noncited violation of 10 CFR 50, Appendix B, Criteria V, instructions, Procedures, and Drawings, after Pacific Gas and Electric failed to promptly evaluate two nonconforming conditions for operability as required by Procedure 8M7.ID12, Operability Determination. The first example involved the failure of engineering personnel to promptly notify plant operations of the failure of the emergency diesel generators to meet licensing and design frequency and voltage recovery requirements. This issue was identified by the NFIC on May 11 , 2010 but not evaluated for the effect on diesel operability until September 9, 201 0. The second example also involved the failure of engineering personnel to promptly notify plant operations to evaluate a nonconforming condition associated with a common cross-tie line that connected both auxiliary saltwater trains. This issue was identified by the NRC on July 22, 2010 but niot evaluated for the effect on auxiliary saltwater operability until August 4, 201 0. In both examples, engineering personnel failed to follow Procedure OM7.ID12, Operability Determination, Section 5.1 , which required any individual identifying a degraded or nonconforming condition that potentially impacts operability of a system, structure or component to ensure that operations shift management is informed. The licensee entered the performance deficiency associated with this finding into the corrective action program as Notifications 50340417 and 50335847. The inspectors concluded that the performance deficiency is more than minor because the Mitigating Systems Cornerstone initial design control attribute and objective to ensure the availability, reliability, and capability of systems that respond to initiating1 events; to prevent undesirable consequences were affected. The finding was of very low safety significance (Green) because neither of the two examples was subsequently determined to result in the loss of operability or functionality. This finding had a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because Pacific Gas and Electric did not thoroughly evaluate the nonconforming conditions for operability (P.1 (c)).
05000275/FIN-2010004-05Inadequate Design Control for the Auxiliary Saltwater System2010Q3The inspectors identified a Green noncited violation of I O CFR Part 50, Appendix B, Criterion Ill, Design Control, involving the failure to maintain adequate design control measures associated with the auxiliary saltwater system. The inspectors identified that the auxiliary saltwater system design did not comply with the plant design bases as described the Final Safety Analysis Report Update. Specifically, an auxiliary saltwater vent line did not meet the requirements established of General Design Criteria 1, Quality Standards and Records, and Regulatory Guide 1.26, Quality Group Classifications and Standards for Water, Steam, and Radioactive Waste Containing Components of Nuclear Power Plants. The licensee entered the performance deficiency into the corrective action program as Notification 50328942. This performance deficiency is greater than minor because the design control attribute of the mitigating systems cornerstone and the cornerstones objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences were affected. Using the Significance Determination Process (SDP) Phase 1 Screening Worksheet for the Mitigating Systems Cornerstone, the inspectors concluded the finding was of very low significance (Green) because it was a design deficiency confirmed not to result in the loss of operability or functionality. The inspectors concluded that the finding does not have a crosscutting aspect since the performance deficiency is not reflective of current plant performance.
05000275/FIN-2010005-01Failure to Maintain a Fire Barrier2010Q4The inspectors identified a noncited violation of Diablo Canyon Facility Operating License Condition 2.C (5), Fire Protection, after Pacific Gas and Electric failed to maintain the integrity of Door 155 in the rated condition. On December 9, 2010, the inspectors identified that the fire door was inoperable. Equipment Control Guideline 18.7, Fire Rated Assemblies, required the licensee to maintain Door 155 in a configuration that would provide at least a 112-hour rated fire barrier. The inspectors previously identified that Door 155 was degraded as a fire barrier in 2009. The licensee entered the violation into the corrective action program as Notification 50367381 and took immediate corrective actions to restore the fire barrier to the rated condition and to implement weekly plant fire door walkdowns The inspectors concluded that the finding was more than minor because the degraded fire barrier affected the Mitigating Systems Cornerstone external factors attribute and objective to prevent undesirable consequences due to fire. The inspectors determined that the finding was within the fire confinement category and that the fire barrier was moderately degraded. The inspectors concluded that the finding was of very low safety significance (Green) because there was a non-degraded automatic full area water-based suppression system in the exposed fire area. This finding had a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee did not take effective corrective actions to following the previous occurrence of the violation (P.1(d)).
05000275/FIN-2010005-02Inadequate Transient Combustibles Procedure2010Q4The inspectors identified a noncited violation of Diablo Canyon Unit 2 Facility Operating License Condition 2.C.(5), Fire Protection, after Pacific Gas and Electric failed to ensure procedures for controlling flammable and combustible materials adequately incorporated requirements of the fire hazard analysis. On October 18, 2010, the inspectors identified that transient combustible materials staged in the Unit 1 12 kilovolt switchgear room did not have an approved transient combustibles permit. The licensee stated that the combustibles permit procedure did not require a permit for the room while Unit 1 was shutdown. However, the plant fire hazards design basis described safe shutdown equipment in the room that would be needed to support a safe shutdown of the operating unit, specifically the Unit 2 startup bus located in the room. The inspectors determined that the licensees transient combustibles permit procedure was inadequate because the procedure did not require a permit for the Unit1 12 kilovolt switchgear room when Unit 2 was operating. The licensee entered the issue into the corrective action program as Notification 50366302 and performed an evaluation of the transient combustibles stored in the area. The inspectors concluded that this finding was more than minor because it affected the Mitigating Systems Cornerstone external factors attribute objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions. The inspectors determined that the finding was within the fire prevention and administrative controls category and represented a low degradation level due to the minimal impact on the effectiveness and reliability of the affected systems. The inspectors concluded that the finding was of very low safety significance (Green) based on a qualitative screening, the low degradation rating, and only equipment needed to reach and maintain cold shutdown conditions was affected. This finding had a crosscutting aspect in the area of human performance associated with the resources component because the licensee failed to ensure that the design documentation adequately identified the Unit 2 startup bus as equipment required for safe shutdown for Unit 2 (H.2(c)).
05000275/FIN-2010005-03Corrosion of Containment Fan Cooler Unit Cooling Coil Casings2010Q4The inspectors identified an unresolved item concerning the degradation of the containment fan cooler unit cooling coil casings due to corrosion. Specifically, the issue concerns the licensees actions to verify the heat removal capability of the containment fan coolers under degraded conditions and the failure to take corrective actions for the repair or replacement of the corroded cooling coil assemblies. The containment fan cooler units function during normal plant operation to maintain the containment atmosphere at design conditions. During accident conditions, the cooler units automatically initiate to maintain containment operability. Diablo Canyon Units 1 and 2 each have five cooler units installed inside the containment building. Each cooler unit has two cooling coil banks with six coils stacked in each bank. Each of the coils is mounted on sheet metal casings and the casings are mounted within the cooler unit frame. The casings act to prevent air bypass between the coils in the banks and as structural support for the coil tubes and fins. The inspectors reviewed Diablo Canyon Power Plant Health Issue Reports 2002-S023-002 and 2002-S023-003 which identified corrosion of the containment fan cooler unit cooling coil casings. The power plant health issue reports acknowledged that continued casing corrosion would decrease the available design margin of the heat removal capacity of the cooler unit cooling coils. Containment Fan Cooler Unit Coil Study, Phase 1, Revision 0, recommended that the corroded coil casing be repaired or replaced whenever possible to avoid impacting the cooler unit heat removal capacity. The licensees initial recommended corrective action, as described in the power plant health issue reports, was to replace the cooling coil assemblies in Units 1 and 2 beginning in Refueling Outage 1R13 (Fall 2005). The replacement plans were not implemented and the licensee currently plans to begin replacement of the cooling coil assemblies during upcoming Refueling Outages 1R18 and 2R18. All cooling coil assemblies in Units 1 and 2 are scheduled to be replaced by Refueling Outage 1R20 and 2R20, respectively. The inspectors determined that additional information was needed to resolve this issue. The inspectors were unable to clearly determine the design basis function of the cooling coil casings based upon documentation provided by the licensee during the inspection. Additionally, the licensee has not quantified the effect of the corrosion to verify that the cooling coil casing functions would be maintained under the current degraded conditions and has not provided a technical justification for the acceptability of the proposed coil assembly replacement schedule. The licensee has agreed to provide this information for NRCs review. Because more information is necessary to resolve this issue, it is considered an unresolved item pending further NRCs review. The NRC will review the licensees evaluation to determine If the licensees failure to verify the heat removal capability of the containment fan cooler units is a performance deficiency If the licensees decision to delay taking corrective actions for repair or replacement of the corroded cooling coils constitutes a violation of NRC requirements The licensee entered this issue into its corrective action program as Notification 50364991. This unresolved item is identified as URI 05000275; 323/2010005-03, Corrosion of Containment Fan Cooler Unit Cooling Coil Casings.
05000275/FIN-2010005-04Inadequate Operability Determinations2010Q4The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criteria V, Instructions, Procedures, and Drawings, after Pacific Gas and Electric failed to adequately evaluate two nonconforming conditions for operability as required by Procedure OM7.ID12, Operability Determination. On October 15, 2010, the inspectors identified a less than adequate technical evaluation supporting Prompt Operability Assessment 50350918, Unit 2 - Insulation in Bio-Wall Penetration. Engineering personnel failed to adequately evaluate the extent of condition after technicians identified about 632 pounds of Temp-Mat and 60 pounds of Min-K fibrous insulation in the Unit 1 reactor coolant loop biological shield wall penetrations. This fibrous material could have potentially been transported and plugged the emergency core cooling containment sump screen. The licensee performed the prompt operability assessment for Unit 2, which was operating at the time. The inspectors concluded that the engineering personnel inappropriately applied the leak-before-break methodology to exclude about 87 percent of this material from the extent of condition review in the prompt operability assessment. The second example involved Prompt Operability Assessment Notification 50355265, RHR Sump Margin, which was completed by the licensee on October 23, 2010. In this example, engineering personnel failed to identify and demonstrate that the specified safety function of the refueling water storage tank could be maintained as required by the plant operability procedure. The inspectors identified that the post accident flow path from the reactor cavity to the containment sump was blocked by a large shield plug. This blockage reduced the amount of post accident inventory available at the containment sump at the time of transition from injection to recirculation mode of emergency core cooling operation. Engineering personnel failed to demonstrate that the safety function to ensure full sump submergence was maintained with the blocked flow path. Full submergence of the sump was used by the NRC as the basis for approval of Technical Specification 3.5.4, Refueling Water Storage Tank, inventory requirements. The licensee entered the violation into the corrective action program as Notification 50369117 and revised the prompt operability assessments using assumptions consistent with the current licensing bases. The inspectors concluded that the performance deficiency was more than minor because the finding affected the Mitigating Systems Cornerstone initial design control attribute and objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors concluded that the finding was of very low safety significance (Green) because the finding was confirmed not to result in the loss of operability or functionality. This finding had a crosscutting aspect in the area of human performance associated with the decision making component because Pacific Gas and Electric did not use conservative assumptions in decisions to demonstrate component operability in either example (H.1(b)).
05000275/FIN-2010005-05Less than Adequate Containment Recirculation Sump Design Control2010Q4The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, after Pacific Gas and Electric failed to ensure Calculation STA-255, Minimum Required Refueling Water Storage Tank Level for GE Sumps, Revision 2, demonstrated adequate available refueling water storage tank inventory. On October 19, 2010, the inspectors identified that emergency core cooling post accident flow path from the reactor cavity to the containment sumps was blocked by a large steel plug on Unit 1. The accident analysis assumed this 35 square foot path was open to allow coolant from a pipe break inside the biological shield to communicate with containment sumps during the recirculation mode of emergency core cooling. The licensee credited the inventory from the reactor cavity when determining the minimum required refueling water storage tank volume in Calculation STA-255. Pacific Gas and Electric used Calculation STA-255 as the basis for determining the minimum required refueling water storage tank volume specified by Technical Specification 3.5.4, Refueling Water Storage Tank. The inspectors identified that the recirculation flow path was also blocked on Unit 2. The inspectors concluded that the most significant contributor to the violation was inaccurate plant drawings used by plant engineers during the performance of Calculation STA-255. The licensees corrective actions included completion of a prompt operability assessment justifying continued operation of Unit 2 and replacement of the shield plug with a movable platform on Unit 1 prior to plant restart The inspectors concluded that the performance deficiency was more than minor because the finding affected the Mitigating Systems Cornerstone plant modification design control attribute and objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors concluded that the finding was of very low safety significance (Green) because the performance deficiency involved a design deficiency confirmed not to result in the loss of operability or functionality. This finding had a crosscutting aspect in the area of human performance associated with the resources component because Pacific Gas and Electric failed to use complete, accurate and up-to-date drawing for Calculation STA 255 (H.2(c)).
05000275/FIN-2010005-06Inadequate Emergency Diesel Generator Surveillance Testing2010Q4The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, after Pacific Gas and Electric failed to develop and implement an adequate testing program for the emergency diesel generators that met design requirements and recommendations. Specifically, in December 2008, the inspectors identified that the diesel generator loading calculations were inadequate to demonstrate that the design bases were met. Pacific Gas and Electric updated the load calculations, but failed to make the necessary revisions to Surveillance Test Procedure STP M-9D1, Diesel Generator Full Load Rejection Test. As a result, Pacific Gas and Electric failed to test several of the emergency diesel generators at the complete load as required by Regulatory Guide 1.108, Revision 1, which is part of the current licensing bases. The licensee entered this into the corrective action program as Notification 50368801, determined there was no loss of safety function for the affected components, and applied the provisions of Surveillance Requirement 3.0.3 for a missed surveillance test. The inspectors concluded the most significant contributor to the finding was less than adequate diesel generator loading evaluations to support corrective action from previous violations associated with the emergency diesel generator testing. The inspectors concluded that the performance deficiency was more than minor because the finding affected the equipment control 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 inspectors determined that the finding was of very low safety significance (Green) because it did not represent an actual loss of safety function of a single train for greater than its technical specification allowed outage time. This finding had a crosscutting aspect in the area of problem identification and resolution, associated with the corrective action program component because the licensee failed to perform an adequate evaluation of the nonconservative surveillance test such that the resolution addressed the fundamental basis for the surveillance (P.1(c)).
05000275/FIN-2010005-07Inadequate Quality Verification Audits2010Q4The inspectors identified a noncited violation of 10CFR Appendix B, Criterion XVIII, Audits, which required that a comprehensive system of planned and periodic audits be carried out to verify compliance with all aspects of the quality assurance program and to determine the effectiveness of the program as well as follow up action, including re-audit of deficient areas, where indicated. Contrary to this requirement, Pacific Gas and Electric failed to ensure that a comprehensive system of planned and periodic audits were carried out to verify compliance with all aspects of the quality assurance program, determine the effectiveness of the program, and perform necessary follow up actions. Specifically, the 2008 Quality Verification audit of the corrective action program failed to adequately address an adverse trend in the problem evaluation process documented in NRC Inspection report 2008005, which identified eleven examples of an adverse trend in problem evaluation. The licensee entered this into their corrective action program as Notification 50365083 and determined there was no loss of safety function for the affected components. The inspectors concluded the most significant contributor to the finding was a less than adequate evaluation of the corrective action trending program. This finding was more than minor because it was associated with the equipment control 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 inspectors determined the performance deficiency was of very low safety significance (Green) it was a deficiency confirmed not to result in the loss of operability or functionality. This finding had a crosscutting aspect in the area of problem identification and resolution, associated with the corrective action program component, because the licensee failed to coordinate and communicate the results from assessments to affected personnel, and track the corrective actions to address issues commensurate with their significance (P.3(c)).
05000275/FIN-2010005-08Licensee-Identified Violation2010Q4Technical Specification 5.4.1.a required that the licensee establish, implement, and maintain written procedures covering the applicable activities recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Regulatory Guide 1.33 recommended procedures required for shutdown of pressurized water reactors. Contrary to this, Pacific Gas and Electric failed to properly implement procedures for shutdown, as required by Technical Specification 5.4.1.a. Specifically, on October 5, 2010, operators failed to follow Operating Procedure OP L-6, Cold Shutdown/Refueling, Step 5.9.6 and establish a reactor coolant system vent path before reducing reactor coolant system temperature below 90F. As a result, operators declared the Low Temperature Overpressure Protection System inoperable in accordance with Technical Specification 3.4.12 and restored temperature above 90F. Pacific Gas and Electric entered the issue into the corrective action program as Notification 50347713. Using Appendix G of Inspection Manual Chapter 0609, Shutdown Operations Significance Determination Process, the inspectors concluded this finding was of very low safety significance because the licensee maintained an adequate mitigation capability during shutdown and the issue did not require a quantitative assessment.
05000275/FIN-2010005-09Licensee-Identified Violation2010Q4Title 10 CFR 50.59 c(2)(vi) required that a licensee obtain a license amendment pursuant to Sec. 50.90 prior to implementing a proposed change, test, or experiment if the change, test, or experiment would create a possibility for a malfunction of a structures, systems, and components important to safety with a different result from any previously evaluated in the FSARU. Contrary to this, Pacific Gas and Electric failed to obtain a license amendment prior to implementing revisions to plant operating procedures. Specifically, on June 29, 2010, Pacific Gas and Electric identified that revisions to Operating Procedure OP L-4, Normal Operation at Power, and OP L-5, Plant Cooldown from Minimum Load to Cold Shutdown allowed operators to control pressurizer level above the program control band during brief periods. The licensee concluded that operation above the program control band could result in a pressurizer overfill condition that could challenge the performance of the pressurizer power-operated relief valves if a spurious safety injection occurred. The licensee entered the issue into the corrective action program as Notification 50320032. The inspectors concluded this finding was of very low safety significance because it is a design deficiency confirmed not to result in the loss of operability or functionality.
05000275/FIN-2010005-10Licensee-Identified Violation2010Q4Title 10 CFR Part 50, Appendix B, Section III, Design Control, required, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. Contrary to this, Pacific Gas and Electric failed to establish measures to ensure that applicable regulatory requirements and the design bases were correctly translated into specifications, drawings, procedures, and instructions. Specifically, on May 14, 2010, Pacific Gas and Electric discovered a design deficiency during the final preparation of the Diablo Canyon Power Plant Unit 1 design change package for the safety injection test line optimization modification. The package lacked the required flow restrictors for the reactor coolant pressure boundary isolation valves, resulting in the potential to create loss of reactor coolant system inventory in Mode 4 which would exceed normal charging makeup capability. On June 14, 2010, the licensee determined that this design deficiency impacted the Unit 2 safety injection modification that had been implemented during the most recent refueling outage. Pacific Gas and Electric entered this issue into the corrective action program as Notification 50316384 and initiated corrective actions to install travel limiting stops on all of the new safety injection test line manual valves to limit the opening size prior to valve testing. The inspectors concluded this finding was of very low safety significance because it is a design deficiency confirmed not to result in the loss of operability or functionality.
05000275/FIN-2010005-11Licensee-Identified Violation2010Q4Title 10 CFR Part 50, Appendix B, Section XI, Test Control, required, in part, that the test program demonstrate that systems and components will perform satisfactorily in service and is performed in accordance with written test procedures that incorporate the requirement and acceptance limits contained in applicable design documents. Contrary to this, Pacific Gas and Electric failed to ensure that testing demonstrated that systems and components would perform satisfactorily in service and perform in accordance with written test procedures which incorporated the requirements and acceptance limits into applicable design documents. Specifically, on November 14, 2010, the licensee performed comprehensive pump testing of Unit 1 turbine-driven auxiliary feedwater pump following replacement of the turbine governor during the most recent refueling outage. The licensee adjusted governor speed settings to meet the pump speed criteria of Procedure STP P-AFW-A11, Comprehensive Testing of Turbine-Driven Auxiliary Feedwater Pump 1-2, Revision 6. Testing personnel identified that the turbine speed settings differed from the previous governor. Following an engineering evaluation, operators performed a surveillance test of the pump and discovered that turbine speed exceeded the test acceptance criteria. The licensee concluded that the new governor had different performance characteristics from the previous governor. Pacific Gas and Electric failed to revise the test procedures to reflect the new governor performance characteristics. As a result, the licensee inappropriately adjusted the governor settings. Pacific Gas and Electric entered this issue into the corrective action program as Notification 50316384 and initiated corrective actions to re-perform comprehensive testing to the pump and adjust the governor to meet pump acceptance criteria. The inspectors concluded this finding was of very low safety significance because it did not result in an actual loss of safety function of the pump.
05000275/FIN-2010005-12Licensee-Identified Violation2010Q4Title 10 CFR Part 50, Appendix B, Section XVI, Corrective Action, required measures be established to ensure that conditions adverse to quality are promptly identified and corrected. Contrary to this, Pacific Gas and Electric failed to ensure corrective actions to identify and correct fibrous material inside containment, as described in licensee Letter DCL-08-059, Supplemental Response to Generic Letter 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents at Pressurized Water Reactors (July 10, 2008). Subsequently, on October 14, 2010, the licensee identified approximately 632 pounds of Temp-Mat and 60 pounds of Min-K fibrous material installed in the Unit 1 reactor coolant loop biological shield wall penetrations. Pacific Gas and Electric entered this issue into the corrective action program as Notification 50355265 and initiated corrective actions to remove the fibrous insulation material. The inspectors concluded that this finding was of very low safety significance because it was a design deficiency confirmed not to result in the loss of operability or functionality.
05000275/FIN-2010006-01Inadequate Design Control for the Emergency Diesel Generator2010Q3The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, involving the failure to maintain adequate design control measures associated with the emergency diesel generating air system. Specifically, failure of non-seismically qualified air compressor unloader sensing lines during a seismic event could impact the safety function of the emergency diesel generators. Subsequent analysis of the nonconforming condition performed by the licensee determined the piping would not fail during a postulated seismic event. The licensee entered this issue into the corrective action program as Notifications 50307496, 50307497, 50307504, 50307670, 50308204, and 50308824. The finding was more than minor because it affected the mitigating systems cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Significance Determination Process (SDP) Phase 1 Screening Worksheet for the Initiating Events, Mitigating Systems, and Barriers Cornerstones the finding was potentially risk significant for a seismic initiating event requiring a Phase 3 analysis. The analyst estimated the nonrecovery probabilities for operators failing to isolate air between the receiver and the compressor prior to air pressure depletion, and operators failing to manually open fuel transfer valves to makeup to the diesel day tank. The final quantitative result was calculated to be 1.06 x 10-6. However, using a qualitative evaluation of the bounding assumptions, the analyst determined that the best available information indicated that the finding was of very low risk significance (Green). The team determined that the finding was reflective of current plant performance because it had been recently identified during the license renewal inspection and had a human performance crosscutting aspect related to decision making because the licensee did not use conservative assumptions when evaluating this nonconforming condition in previous evaluations
05000275/FIN-2010006-02Failure to Maintain Proficiency of Operators to Meet the Time Critical Operator Actions2010Q3The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, involving the failure to ensure that operators are able to implement specified actions in response to operational events and accidents. Specifically, operators could not achieve actions within the analysis time estimates for the cold leg recirculation phase of a loss of coolant accident response and the steam generator tube rupture response as described in the licensees safety analysis report. The finding is more than minor because it affected the Mitigating Systems Cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding represented a potential loss of a safety function requiring a Phase 2 analysis. Because the probability of human error is not effectively addressed by a Phase 2 analysis, a Phase 3 analysis was performed. The senior reactor analyst reviewed the actual timing of the walkdowns associated with the steam generator tube rupture time critical actions. The analyst determined that, while the licensee failed to meet the specific cooldown timing documented in the Final Safety Analysis Report, the total time to start cooling the reactor was well within the total critical timing of the event. The analyst found no impact on safety in delaying the cooldown of the reactor for one minute given that the other time critical actions were performed more quickly than required. Therefore, the analyst determined that this portion of the finding was of very low safety significance because it does not represent an actual loss of safety function (Green). The senior reactor analyst reviewed the issue related to the assumed action times associated with switching over to containment sump recirculation lineup for their emergency core cooling system pumps during a large break loss of coolant accident. The analyst noted that this time critical action was only required if a large-break loss of coolant accident occurred simultaneously with the failure of an residual heat removal pump to stop automatically, requiring local isolation of the pump. Given that the frequency of the initial conditions for the time critical action are below the Green/White threshold, the change in core damage frequency associated with this finding must be of very low safety significance (Green). The team determined that the finding was reflective of current plant performance because the licensee participated in a recent industrywide study on time critical operator actions, but did not implement any of the groups recommendations. The finding had a crosscutting aspect in the area of human performance, decision making, because the licensee did not use conservative assumptions in the decision making process related to verifying the validity of the underlying assumptions used to evaluate the feasibility of operators implementing time critical operator actions
05000275/FIN-2010006-03Failure to Submit Complete and Accurate Information for a Requested License Amendment2010Q3The team identified a noncited violation of 10 CFR 50.9(a), Completeness and Accuracy of Information with multiple examples. Specifically, information supplied to the NRC in License Amendment Request 01- 10, dated February 24, 2010, related to the revision of Technical Specification 3.8.1, AC Sources - Operating, were not complete and accurate in all material respects. Following NRC questioning of the discrepancies the licensee withdrew the amendment request. The finding is more than minor because the inaccurate information was material to the NRC. Specifically, this information was under review by the NRC to evaluate specific changes to the surveillance requirements associated with the emergency diesel generators. Following management review, this violation was determined to be of very low safety-significance because the amendment request was withdrawn before the NRC amended the facility technical specifications. Because this issue affected the NRC\'s ability to perform its regulatory function, it was evaluated with the traditional enforcement process. Consistent with the guidance in Section IV.A.3 and Supplement VII, paragraph D.1, of the NRC Enforcement Policy, this finding was determined to be a Severity Level IV noncited violation. The finding had a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program because the licensee did not adequately evaluate the extent of condition and take appropriate corrective actions after the NRC identified a similar violation
05000275/FIN-2010006-04Untimely and Inadequate Corrective Actions for the Emergency Diesel Generators2010Q3The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, with two examples for the failure of the licensee to promptly identify and correct nonconforming conditions related to the emergency diesel generators meeting the design basis. The first example resulted from the failure to identify that instrument inaccuracies were not accounted for in the bounding calculations. The second example involved the failure to identify that the worst case loading calculations exceeded the emergency diesel generator operating load limit. The failure to promptly identify and correct the design deficiencies associated with the emergency diesel generators was a performance deficiency. This finding is greater than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and affected the cornerstones objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. In accordance with Inspection Manual Chapter 0609, Significant Determination Process, the team performed a Phase 1 analysis to analyze the significance of this finding and determined the finding is of very low safety significance because the condition was a design or qualification deficiency confirmed not to result in loss of operability or functionality, did not represent an actual loss of safety function of the system or train, did not result in the loss of one or more trains of nontechnical specification equipment, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The finding had a crosscutting aspect in the area of human performance, decision making, because the licensee did not use conservative assumptions in the decision making process or conduct an adequate effectiveness review to verify the validity of the underlying assumptions for a safety-significant decision
05000275/FIN-2010006-05Failure to Appropriately Evaluate Failed Residual Heat Removal Surveillance Test2010Q3The team identified a noncited violation of Technical Specification 5.4.1.a for failure to appropriately evaluate and correct a condition adverse to quality, as instructed by Surveillance Test Procedure P-RHR-A22, Comprehensive Testing of Residual Heat Removal Pump. Specifically, the licensee failed to recognize a deviation in differential pressure towards the alert range, following the February 9, 2008, comprehensive surveillance test of the 2-2 residual heat removal pump. Continued degradation of the 2-2 residual heat removal pump resulted in failure of the October 9, 2009, comprehensive surveillance test due to the differential pressure exceeding the action limit. The licensee entered this issue into the corrective action program as Notification 50308225. The finding is more than minor because it was associated with the equipment reliability attribute of the Mitigating Systems Cornerstone and it adversely affected the associated cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team evaluated the finding in accordance with Inspection Manual Chapter 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Table 4a for the Mitigating Systems Cornerstone. The finding was determined to be of very low safety significance (Green) because: (1) it was a design or qualification issue confirmed not to result in a loss of operability or functionality; (2) did not represent an actual loss of safety function of the system or train; (3) did not result in the loss of one or more trains of nontechnical specification equipment; and (4) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The team determined that this finding had a crosscutting aspect in the area of problem identification and resolution, corrective action program, because the licensee failed to appropriately evaluate the 2009 residual heat removal surveillance test failure such that the resolution identified and corrected the cause of the failure
05000275/FIN-2010007-01Less than Adequate Change Evaluation to the Facility as Described in the Final Safety Analysis Report Update2010Q2The inspectors identified a Severity Level IV noncited violation of 10 CFR 50.59 for the licensee\'s failure to demonstrate that prior NRC approval was not required prior to making changes to the facility degraded voltage protection scheme as described in the Final Safety Analysis Report Update. In response to this violation, the licensee re-performed the corresponding safety analysis to demonstrate that the subject change to the facility degraded voltage protection scheme was consistent with General Design Criteria 17, \"Electric Power Systems.\" The violation is in the licensee\'s corrective action program as Notification 50306053. The failure of Pacific Gas and Electric to perform a 10 CFR 50.59 evaluation of modifications to the offsite power protection scheme, in accordance with NEI 96 07, was a performance deficiency. The violation was more than minor because of a reasonable likelihood the change to the facility would require Commission review and approval prior to implementation. The violation screened as very low safety significance (Green) because the finding was not a design or qualification deficiency confirmed not to result in loss of operability or functionality, did not represent a loss of system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The finding has a crosscutting aspect in the area of human performance associated with the decision-making component because the licensee did not adopt the requirement to demonstrate that the proposed action was safe in order to proceed rather than a requirement to demonstrate that the proposed action was unsafe in order to disapprove the action, in that the Plant Safety Review Committee did not require that a 50.59 evaluation be performed to demonstrate that the proposed action was safe in order to proceed.
05000275/FIN-2010007-02Failure to Adequately Evaluate Changes to the Diesel Testing as Described in the Final Safety Analysis Report Update2010Q2The team identified two examples of a Severity Level IV noncited violation of 10 CFR 50.59 after the licensee failed to perform an adequate evaluation to demonstrate that prior NRC approval was not required before making changes to the frequency and voltage recovery criteria and to the diesel testing commitments as described in the Final Safety Analysis Report Update. Specifically, the 1998 Final Safety Analysis Report Update identified a change from Safety Guide 9 to Regulatory Guide 1.9, Revision 2. The scope involved the removal of the KWS relay and included new requirements for voltage and frequency response. This resulted in a reduction in acceptance criteria. The team also identified a second example where the licensee failed to evaluate the 2005 Final Safety Analysis Report Update change from Regulatory Guide 1.9, Revision 2 to Revision 3 for diesel testing and interval frequency. Using NEI 96-07, \"Guidelines for 10 CFR 50.59 Evaluations,\" Revision 1, the team concluded that these changes resulted in a departure from a method of evaluation described in the Final Safety Analysis Report Update establishing the facility design bases. In addition, the licensee\'s 50.59 evaluation, for DCP E-049424, Revision 0, \"EDG Starting, and Loading Capability\" was less than adequate to conclude that prior NRC approval was not required for the changes. The licensee has entered these issues into their corrective action program as Notifications 50302467 and 50302481. The failure of Pacific Gas and Electric to perform an adequate 10 CFR 50.59 evaluation prior to changing the facility as described in the Final Safety Analysis Report Update is a performance deficiency. The violation was more than minor because of a reasonable likelihood the change to the facility would require Commission review and approval prior to implementation. The violation screened as very low safety significance (Green) because the finding was not a design or qualification deficiency confirmed not to result in loss of operability or functionality, did not represent a loss of system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. Because this violation is of very low safety significance and was entered into the licensee\'s corrective action program, this violation is being treated as a noncited violation, consistent with Section VLA.1 of the NRC Enforcement Policy. The finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component. In License Amendment Request 10-01, dated February 24, 2010, the licensee did not thoroughly evaluate the original problem of using the 10 CRF 50.59 evaluation process to justify using Regulatory Guide 1.9, Revision 2, Section C, Position 4, as an exception to meeting the frequency and voltage criteria identified in Safety Guide 9.
05000275/FIN-2010007-03Non-Conservative Decision Making resulted in a Violation of Technical Specification2010Q2On March 10, 2010, the inspectors identified a noncited violation of Technical Specification 3.7.7, \"Vital Component Cooling Water System,\" after both Unit 2 component cooling water loops were inoperable longer than permitted during power operations. On March 9, 2010, Pacific Gas and Electric identified that the degraded voltage protection scheme was inadequate to ensure minimum required voltage would be availab!e to operating engineered safety feature pumps during a degraded offsite power grid. The licensee concluded that operating pumps could trip and lock out on over current before the protection scheme would automatically transfer power to the emergency diesel generators. The licensee declared the 230kV offsite power systems inoperable and took compensatory actions to enable the automatic transfer of busses with operating engineered safety feature pumps directly to the diesel generators following a unit trip. On March 10, 2010, the inspectors identified that operating component cooling water pump 2-3 was still aligned to automatically transfer to 230kV offsite power source following a unit trip. The licensee had previously removed component cooling water pump 2-2 from service for maintenance on March 7, 2010. Technical Specification 3.7.7, \"Vital Component Cooling Water System,\" required a minimum of two operable component cooling water pumps to establish operability of a vital component cooling water loop. Contrary to Technical Specification 3.7.7, on March 10, 2010, the licensee operated Unit 2 without an operable vital component cooling water loop for greater than 14 hours. The licensee has entered this issue into their corrective action program as Notification 50304802. Either the failure of Pacific Gas and Electric to restore at least two operable component cooling water pumps or to have placed Unit 2 in Mode 3 within six hours, as required by plant Technical Specification 3.7.7, was a performance deficiency. The performance deficiency is more than minor because it is associated with the Mitigating Systems Cornerstone attribute of equipment performance, of ensuring the availability, reliability, and capability of safety systems that respond to initiating events to prevent undesirable consequences (i.e., core damage), and it was within the licensee\'s ability to correct this problem. The inspectors used Inspection Manual Chapter 0609, Appendix A, \"Determining the Significance of Reactor Inspection Findings for At-Power Situations,\" to analyze the finding because the violation represents the actual loss of safety function for greater than the technical specification allowed outage time. The finding was of very low safety significance (Green) based on a bounding qualitative evaluation using Inspection Manual Chapter 0609, Appendix M, \"Significance Determination Process Using Qualitative Criteria.\" The inspectors based this conclusion on the low probability of an actual degraded grid condition coincidental with an accident or anticipated operational occurrence during the 14hour exposure that the vital component cooling water loops were unavailable due to the performance deficiency. The inspectors concluded that this finding had a crosscutting aspect in the area of human performance associated with the decision-making component because the licensee did not use conservative assumptions in their decision to implement compensatory actions following the inoperability of the degraded voltage protection scheme.
05000275/FIN-2010007-04Failure to Update the Final Safety Analysis Report Update with the Current Plant Design Bases Section2010Q2The inspectors identified a Severity Level IV noncited violation of 10 CFR 50.71 after Pacific Gas and Electric failed to include the current plant design basis for the 230kV degraded voltage protection scheme in the Final Safety Analysis Report Update. Title 10 CFR 50.71.(e) states in part, \"Each person licensed to operate a nuclear power reactor shall update periodically, as provided in paragraphs (e) (3) and (4) of this section, the Fina! Safety ,ll.na!ysis Report Update originally submitted as part of the application for the operating license, to assure that the information included in the report contains the latest information developed. Contrary to the above, on March 14, 2010, the inspectors identified that Pacific Gas and Electric failed to update the Final Safety Analysis Report Update to include complete design basis information for the offsite degraded voltage protection scheme. The inspectors identified that Final Safety Analysis Report Update did not include the design basis for the allowable time delay or the limiting voltage setpoints. The licensee has entered this issue into their corrective action process as Notification 50313763. Failure to include the current plant design basis for the 230kV degraded voltage protection scheme in the Final Safety Analysis Report Update is a performance deficiency. Using inspection Manual Chapter 0612, Appendix B, the team determined that this issue was to be evaluated using the traditional enforcement process because the performance deficiency was a failure to meet a requirement or standard, had the potential for impacting the NRC\'s ability to perform its regulatory function, and the concern was within the licensee\'s ability to foresee and correct and should have been prevented. The team used the General Statement of Policy and Procedure for NRC Enforcement Actions, Supplement I \"Reactor Operations,\" dated January 14,2005, to evaluate the significance of this violation. The team concluded that the violation is more than minor because the incorrect Final Safety Analysis Report Update information had a potential impact on safety and licensed activities. Using Supplement I, Section D, Item 6, of the NRC Enforcement Policy, this performance deficiency will be treated as a Severity Level IV violation because the erroneous information was not used to make any unacceptable change to the facility or procedures. Using Inspection Manual Chapter 0609.04, \"Phase 1 Initial Screening and Characterization of Findings,\" the team concluded that the issue screened as having low safety significance (Green) under the significance determination process. Because this violation is of very low safety significance and was entered into the licensee\'s corrective action program, this violation is being treated as a noncited violation, consistent with Section VI.A.1 of the NRC Enforcement Policy. Because the violation included a performance deficiency, the inspectors also concluded the issue was a finding under the Reactor Oversight Process and had a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee did not adequately evaluate the extent of the condition and take appropriate corrective actions after the NRC identified a similar violation.
05000275/FIN-2010007-05Inadequate Operability Determination Associated With the Offsite Degraded Voltage Protection Scheme2010Q2On February 27, 2010, the inspectors identified a noncited violation of 10 CFR 50, Appendix B, Criteria V, \"Instructions, Procedures, and Drawings,\" after Pacific Gas and Electric failed to complete an adequate operability evaluation, as required by Procedure OM7.ID12, \"Operability Determination,\" Revision 14. The inspectors identified that the offsite power degraded voltage protection scheme time delay was inconsistent with key assumptions in the accident analysis. The licensee entered this nonconforming condition into the corrective action program as Notification 50301167 on February 24,2010. Plant operators subsequently requested plant engineering to perform an operability determination of the nonconforming condition per Operability Determination Procedure OM7.ID12. On February 27, 2010, the plant operating authority concluded that the protection scheme was operable based on the information provided in the operability determination. Contrary to the above, on March 2, 2010, the inspectors concluded that the licensee\'s operability determination was inadequate to demonstrate protection scheme operability and was not performed as required by Operability Determination Procedure OM7.ID12. Plant engineering only addressed the capability of the protection scheme at normal grid voltage following a mechanical failure of the 230kV load tap changer. Operability Determination Procedure OM7.ID12, Section 5.3, \"Write the Prompt Operability Assessment (POA),\" required that the licensee address the potential effect of the nonconforming condition to perform the specified safety function. The licensee has entered this finding into the corrective action program as Notification 50319258. Failure to complete an adequate operability evaluation, as required by Procedure OM7.ID12, \"Operability Determination,\" Revision 14, is a performance deficiency. Using Inspection Manual Chapter 0612, Appendix 8. The performance deficiency is more than minor because it is associated with the Mitigating Systems Cornerstone attribute of procedure quality. and the failure to perform an adequate operability evaluation affects the ability to ensure operability of the protection scheme at normal grid voltage following a mechanical failure of the 230kV load tap changer. The inspectors used Inspection Manual Chapter 0609. Appendix A, \"Determining the Significance of Reactor Inspection Findings for At-Power Situations,\" to analyze the finding because the violation represent the actual loss of safety function for greater than the Technical Specification 3.3.5 allowed outage time. Using Appendix M. of the \"Significance Determination Process Using Qualitative Criteria.\" the inspectors concluded that the finding was of very low safety significance (Green) based on a bounding qualitative evaluation. The inspectors based this conclusion on the low probability of an actual degraded grid condition coincidental with an accident or anticipated operational occurrence during the exposure time that protection scheme was available due to the performance deficiency. This finding had a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because Pacific Gas and Electric did not thoroughly evaluate the nonconforming condition for operability and reportability.
05000275/FIN-2010007-06Second Level Undervoltage Relay Time Delay to Initiate Load Shed and Sequencing Upon the Diesel Generator is Adequate to Assure Plant Safety2010Q2The team identified a noncited violation of 10 CFR Part 50. Appendix B. Criterion III, \"Design Control.\" for failure to ensure that plant conditions were consistent with design calculation inputs and assumptions. The licensee failed to assure and verify that Technical Specification 3.3.5 (SR3.3.5.3) pertaining to the second level undervoltage relay time delay to initiate load shed and sequencing upon the diesel generator was adequate to assure plant safety. Supplemental Safety Evaluation Report 09, Section 8.1. requires that a second level of under voltage protection for the onsite power system be provided. Subsection (1)(c)(i), reads: \"The allowable second level undervoltage relay time delay, including margin. shall not exceed the maximum time delay that is assumed in the Final Safety Analysis Report Update accident analyses.\" Contrary to the above, as of March 4. 2010. the licensee failed to adequately implement the requirements of Supplemental Safety Evaluation Report 09. The second level undervoltage relay time delay setpoint for the emergency diesel generator of less than or equal to 20 seconds, assuming a safety injection signal concurrent with a degraded off site power source, exceeded the Final Safety Analysis Report Update accident analysis. This item identified in the licensee\'s corrective action document Notification 50301167. Failure to ensure that plant conditions were consistent with design calculation inputs and assumptions is a performance deficiency. Using Inspection Manual Chapter 0612, Appendix E, Section 3 Example j, the vioiation was determined to be more than minor because the engineering calculation error results in a condition where there is now a reasonable doubt on the operability of a system or component. These deficiencies represented reasonable doubt regarding the mitigation of an accident by being in an unanalyzed condition. Using Inspection Manual Chapter 0609, \"Significance Determination Process,\" Phase 2, the finding was determined to have very low safety significance (Green), did not represent an actual loss of a system safety function, did not result in exceeding a technical specification allowed outage time, and did not affect external event mitigation. The team reviewed the finding for crosscutting aspects and none were identified.
05000275/FIN-2010007-07Nonconservative Inputs into Motor Operated Valve Calculation2010Q2The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, \"Design Control\" involving multiple mechanical and electrical errors in documentation associated with torque, thrust, and operation of certain safety related motor operated valves. Specifically, on March 3, 2010, the team identified seven different numerical inputs to the operation, margin, and functionality of certain safety related motor operated valves, which had nonconservative inputs for the mechanical and electrical settings of the motor operated valves. The composite of these nonconservative inputs could have a detrimental effect on the operation of these motor operated valves. Contrary to above, on March 3, 2010, per 10 CFR Part 50, Appendix B, Criterion III, \"Design Control\", the licensee did not establish measures to assure that applicable regulatory requirements and the design basis, for those structures, systems, and components were correctly translated into specifications, drawings, procedures, and instructions. The licensee has entered this issue into their corrective action process as Notification 50302437. The failure of the licensee to correctly translate design basis information into specifications, drawing, procedures, and instructions is a performance deficiency. Using Inspection Manual Chapter 0612, Appendix E, Section 3, Example j, the violation was determined to be more than minor because the engineering calculation error results in a condition where there is now a reasonable doubt on the operability of a system or component. Using the Inspection Manual Chapter 0609, \"Significance Determination Process,\" Phase 1 Worksheets, the finding was determined to have very low safety significance (Green) because it was not a design issue resulting in loss of function, did not represent an actual loss of a system safety function, did not result in exceeding a technical specification allowed outage time, and did not affect external event mitigation. The inspectors reviewed the finding for crosscutting aspects and none were identified.
05000275/FIN-2010007-08Inadequate Drawings and Procedures to Align Emergency Makeup Water Supply from Diablo Canyon Creek to Support the Auxiliary Feedwater System2010Q2The team identified a noncited violation of Diablo Canyon Technical Specification 5.4.1. \"Procedures,\" for failure to have a procedure. The Diablo Canyon Final Safety Analysis Report Update, Revision 18, Section 6.5.2.1.1 documents the design of the auxiliary feedwater system and credits eight sources of water that can provide backup means of supply in the event that its primary source of water, the condensate storage tank, becomes exhausted. One of the sources included is the Diablo Canyon Creek. Diablo Canyon Technical Specification 5.4.1 states: \"Written procedures shall be established, implemented, and maintained covering the following activities: (a.) The applicable procedures recommended in NRC Regulatory Guide 1.33, Revision 2, Appendix A, February 1978\". NRC Regulatory Guide 1.33, Revision 2, Appendix A, describes procedures under Items 31 (instructions for shutdown, startup, and operation, including system filling, of the auxiliary feedwater system) and 6j (loss of feedwater system or feedwater system failure). Contrary to Technical Specification 5.4.1, on March 4, 2010, the team identified that the licensee did not have an established procedure to accomplish the task identified in the Final Safety Analysis Report Update, Section 6.5.2.1.1 for taking water from the Diablo Canyon Creek to be a supply for the auxiliary feedwater system. This item identified in the licensee\'s corrective action document Notification 50298563. Failure to provide a procedure or instructions and acceptance criteria to perform an emergency makeup water alignment to the auxiliary feedwater system is a performance deficiency. Using Inspection Manual Chapter 0612, Appendix B, the performance deficiency is more than minor because it is associated with the Mitigating Systems Cornerstone attribute of procedure quality, and the lack of having this procedure affects the ability to ensure the availability, reliability, and capability of the auxiliary feedwater system to respond to initiating events to prevent undesirable consequences, (i.e., core damage, and it was within the licensee\'s ability to correct this problem.) Using the Inspection Manual Chapter 0609, \"Significance Determination Process,\" Phase 1 Worksheets, the finding was determined to have very low safety significance (Green) because it was not a design issue resulting in loss of function, did not represent an actual loss of a system safety function, did not result in exceeding a technical specification allowed outage time, and did not affect external event mitigation. The team concluded that this finding had a crosscutting aspect in the area of problem identification and resolution, in that the licensees\' corrective action program thoroughly evaluates problems such that the resolutions address causes and the extent of conditions, as necessary. Per licensee Notification 50298563, changes were made to pumping systems associated with the Diablo Canyon Creek in 2007, which affected the ability to pump water through the discussed credited lineup supporting the auxiliary feed\\,v8ter system. This effect was not identified as part of the changes, so no review of procedures related to the emergency auxiliary feedwater system alignment in question was performed. Since these actions occurred within the last three years, this performance characteristic reflects current performance.
05000275/FIN-2010007-09Failure to Update Feedwater Rupture Accident Analysis in the Final Safety Analysis Report Update2010Q2The team identified a Severity Level IV noncited violation of 10 CFR 50.71, \"Maintenance of records, making of reports.\" Paragraph (e) states, \"Each person licensed to operate a nuclear power reactor shall update periodically the final safety analysis report originally submitted as part of the application for the license, to assure that the information included in the report contains the latest information developed.\" in the Diablo Canyon Final Safety Analysis Report Update section addressing the feedwater line break accident, it states that operator actions are credited with precluding the operation of pressurizer safety valves based on determinations in Westinghouse study WCAP-11667 (1998) (Final Safety Analysis Report Update, Section 15.4.2.2.2). Review of this study, and associated correspondence on the topic during 2006 indicated that the Westinghouse study did not state that operator actions could be credited for this event, but analysis of the worst case pressurizer overfill accidents by the licensee may show that this is the bounding case for such accidents, and that it did not need to be addressed in the feedwater line break analysis. In 2006, the licensee indicated that they would revise the Final Safety Analysis Report Update text to remove this reference to the Westinghouse study, which had been in the Final Safety Analysis Report Update since Revision 16. Contrary to the above, since 2006 (Final Safety Analysis Report Update Revision 16), the licensee failed to update Final Safety Analysis Report Update, Section 15.4.2.2.2. The licensee has entered this issue into their corrective action process as Notification 50301747. Failure to periodically update the Final Safety Analysis Report Update with a known error is a performance deficiency. Using Inspection Manual Chapter 0612, Appendix B, the team determined that this issue was to be evaluated using the traditional enforcement process because the performance deficiency was a failure to meet a requirement or standard, had the potential for impacting the NRC\'s ability to perform its regulatory function, and the concern was within the licensee\'s ability to foresee and correct and should have been prevented. The team used the General Statement of Policy and Procedure for NRC Enforcement Actions, Supplement I, \"Reactor Operations,\" dated January 14, 2005, to evaluate the significance of this violation. The team concluded that the violation is more than minor because the incorrect Final Safety Analysis Report Update information had a potential impact on safety and licensed activities. Using Supplement I, Section 0, Item 6, of the NRC Enforcement Policy, this performance deficiency will be treated as a Severity Level IV violation. Because this violation is of very low safety significance and entered into the licensee\'s corrective action program, this violation is being treated as a noncited violation, consistent with Section VI.A.1 of the NRC Enforcement Policy. The inspectors reviewed the finding for crosscutting aspects and none were identified.
05000275/FIN-2010007-10Failure to Update Text to Reflect Credited Design Class I Makeup Flowpath to Component Cooling Water Expansion Tank in the Final Safety Analysis Report Update2010Q2The team identified a Severity Level IV noncited violation of 10 CFR 50.71, \"Maintenance of records, asking of reports.\" Title 10 CFR 50.71, paragraph (e) states, \"Each person licensed to operate a nuclear power reactor shall update periodically the final safety analysis report originally submitted as part of the application for the license, to assure that the information included in the report contains the latest information developed.\" In the Final Safety Analysis Report Update Table 9.2-7, Component 5, it states, \"This 250 gpm, Design Class I, makeup water flowpath, described under Makeup Provisions in Subsection 2.3.3 (Section 9.2.2.3.3), can be started within 10 minutes.\" Final Safety Analysis Report Update, Section 9.2.2.3.3 states, \"All piping and valves in the makeup path from the condensate storage tank (including their crossconnections) and the firewater tank, through the makeup water transfer pumps up to and including the makeup valves on the component cooling water system lines, are Design Class I.\" Text later in the section implied that the flow path from the firewater tank was not Design Class I. Review by the licensee staff revealed that the only Design Class I flow path to provide makeup to the component cooling \'vvater expansion tank is via the condensate storage tank. This revealed that the text provided in Final Safety Analysis Report Update, Section 9.2.2.3.3 stating that both the condensate storage tank and firewater tank makeup paths are credited is incorrect. Contrary to above, since 1984 (Final Safety Analysis Report Update, Revision 0), the licensee did not update Final Safety Analysis Report Update, Section 9.2.2.3.3 to correct the error of including firewater as a possible makeup path to the component cooling water expansion tank. The licensee has entered this issue into their corrective action process as Notification 50301884. Failure to periodically update the Final Safety Analysis Report Update with a known error is a performance deficiency. Using Inspection Manual Chapter 0612, Appendix B, the team determined that this performance deficiency was to be evaluated using the traditional enforcement process because the performance deficiency had the potential for impacting the NRC\'s ability to perform its regulatory function. Using General Statement of Policy and Procedure for NRC Enforcement Actions, Supplement I, Reactor Operations, dated January 14, 2005, to evaluate the significance of this violation, the team concluded that the violation is more than minor because the incorrect Final Safety Analysis Report Update information had a potential impact on safety and licensed activities. Using Supplement I, Section D, Item 6, of the NRC Enforcement Policy, this performance deficiency will be treated as a Severity Level IV violation. Because this violation is of very low safety significance and entered into the licensee\'s corrective action program, this violation is being treated as a noncited violation, consistent with Section VI.A.1 of the NRC Enforcement Policy. The team reviewed the finding for crosscutting aspects and none were identified.
05000275/FIN-2010007-11Licensee-Identified Violation2010Q2Title 10 CFR Part 50, Appendix B, Criterion XI, \"Test Control\", required Pacific Gas and Electric to establish a test program to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is performed in accordance with written test procedures, which incorporate the requirements, and acceptance limits contained in applicable design documents. Contrary to the above, in February 2010, Pacific Gas and Electric discovered that they had failed to incorporate into Procedure STP M-15, \"Integrated Test of Engineered Safeguards and Diesel Generators,\" requirements and acceptance limits contained in Regulatory Guide 1.9, Revision 2, regarding voltage and frequency dips and recovery during design basis loading sequencing onto the emergency diesel generators. After selfidentification of the issue, the licensee entered this issue in their corrective action program as Notification 50294602. This finding is of very low safety significance because the condition did not screen as potentially risk significant, in that it was a design deficiency that did not result in actual loss of safety function.
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05000275/FIN-2011002-01Inadequate Design Control for the Preferred Offsite Power System2011Q1The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, after Pacific Gas and Electric failed to ensure that the preferred offsite power system design basis was correctly translated into electrical dynamic loading Calculations 357A-DC, Units 1 and 2 Load Flow, Short Circuit and Motor Starting Analysis, Revision 12 and 359-DC, Offsite Power Dynamic Analysis, Revision 8. The licensee did not include the limiting load flow cases representing the largest total onsite demand for both units as required by the plant design basis. On July 7, 2010, the NRC clarified that the Diablo Canyon current licensing basis required the preferred offsite power system to have adequate capacity and capability to supply the most limiting loading requirements, including a dual unit trip. The licensee subsequently entered the condition into the corrective action program as Notification 50289590 and revised the station dynamic loading analysis to reflect the increased onsite power demand. The inspectors concluded that the failure to ensure that the dynamic loading analysis included all design basis requirements was a performance deficiency. This performance deficiency is more than minor because the finding was associated with the Mitigating Systems Cornerstone initial design control attribute and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Because the inspectors were unable to conclude that the preferred offsite power system had not been inoperable for greater than the allowed Technical Specification outage time, a senior reactor analyst performed a bounding Phase 3 analysis. The Phase 3 analysis demonstrated that the subject finding was of very low safety significance (Green), because of the small increase of probability of a loss of offsite power that the finding represented. This finding had a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee did not thoroughly evaluate the current licensing basis requirements to ensure that resolutions addressed causes and extent of conditions, as necessary (P.1(c)).
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