Diablo Canyon Independent Safety Committee

Diablo Canyon Independent Safety Committee

3.0 Nuclear Regulatory Commission (NRC) Assessments and Issues, DCISC 17th Annual Report - July 1, 2006 thru June 30, 2007

This section of the DCISC Annual Report describes the DCISC review of PG&E’s interface with the US Nuclear Regulatory Commission (NRC). The NRC is the Federal regulatory entity charged with assuring the safety and security of domestic nuclear power plants; by agreement with the State, NRC also performs these functions for the State of California. As regulator, the NRC employs two full-time Resident Inspectors at the plant (and other specialist inspectors at its US headquarters and regional locations), performs and reports on its inspections at DCPP on matters of nuclear safety and security, investigates significant plant events, maintains a set of plant performance indicators, and performs an annual assessment of DCPP regulatory performance which it reports at a public meeting in the plant vicinity. The NRC also must approve significant changes, additions and deletions to plant designs, procedures and Technical Specifications.

PG&E is required to submit routine, periodic reports to the NRC on selected activities and submit special reports when triggered by off-normal plant incidents, events or occurrences

The DCISC monitors the aforementioned activities and resulting documents in the following ways: (1) receipt and review of correspondence and reports between PG&E and the NRC, (2) on-site review (at Fact-finding meetings at the plant) of selected NRC inspections, investigations and reports, (3) meetings with the NRC Resident Inspectors, and (4) presentations by PG&E at DCISC public meetings on NRC matters.

3.1 Summary of License Event Reports

3.1.1 Discussion and Required LERs

License Event Reports (LERs) are reports required of the nuclear power plant licensee by Nuclear Regulatory Commission (NRC) regulations when an off-normal event occurs. These events include operations or conditions outside of or in violation of station Technical Specifications (TS), procedures or NRC regulations. Events are to be promptly reported by telephone and by written report within 60 days of the event or initial knowledge of the event. Voluntary LERs are submitted for events which NRC should know about or are significant but are not specifically required by NRC. Each of these reports is reviewed in DCISC public meetings and is mailed to each DCISC Member and Consultant.

The LER is the responsibility of the Licensee, in this case PG&E. Therefore, it is the Licensee who makes the determination of the level of risk or significance to safety of the event. The NRC has a Significance Determination Process which sets forth its rules for making these determinations; however, events may be complex or may not easily fit the rules.

The NRC may concur or it can question or challenge the Licensee’s determination. Discussions or meetings may be required to reach understandings between the parties.

Three LERs were reported during this time period and corresponding corrective action was as follows:

1. On August 15, 2006 DCPP discovered approximately 100 dead cormorants (diving sea birds common to the California central coast) in the plant intake cove. In accordance with its Facility Operating License Environmental Protection Plan (non-radiological), DCPP made a non-emergency report to the NRC. DCPP believed the birds died of natural causes and not plant operation. Some of the cormorants remains were sent to the University of California at Davis for testing and analysis. Because the deaths were not related to plant operation, no corrective actions were identified.
2. On December 10, 2006 DCPP licensed operators manually tripped the Unit 2 reactor while it was subcritical. Prior to the trip, Unit 2 was operating at 100% power when Reactor Coolant Pump (RCP) 2-2 stator temperature indication increased. Operators initiated an unplanned reactor shutdown in accordance with plant procedures. When the temperature indication reached 300°F, operators manually tripped the reactor as required by procedures. All rods inserted fully and all systems functioned as required. The three remaining RCPs maintained forced circulation and cooling of the reactor coolant system.
The DCPP investigation revealed that a stator resistance temperature detector (RTD) had failed causing a false high temperature indication. The failed RTD input was replaced with an installed spare that was verified to be operating satisfactorily.
Although the cause was random failure of an RTD, DCPP initiated three corrective actions.
  1. The RTD stator temperature alarm response procedure was revised to provide additional guidance about information to be gathered for a formal operational decision making prior to initiation of a reactor trip.
  2. A program for RTD health monitoring and performance of periodic assessment of critical RTDs identified as single point vulnerabilities was developed to aid in identifying potentially-failing RTDs.
  3. The Plant Process Computer alarm setpoint was revised for the RCP status RTDs to provide additional margin between alarm initiation and required operator action.
3. On December 12, 2006 with Unit 2 at approximately 25 percent power and increasing, an electrical failure (phase-to-phase internal fault) of a motor surge capacitor occurred in the Circulating Water Pump (CWP) 2-1 motor enclosure. The electrical transient experienced on the 12kV non-vital bus actuated an undervoltage protection relay tripping the load breakers for CWP 2-1 as well as Reactor Coolant Pumps (RCPs) 2-2 and 2-4 motor breakers. This caused a reactor trip signal as programmed into the Reactor Protection System trip logic. All control rods inserted fully, all plant systems functioned as required, and operators responded appropriately. A small fire was extinguished. The root cause of the capacitor failure was an in-service insulation breakdown, considered a single random failure. Contributing causes were (1) the use in 1993 of single three-phase capacitors in lieu of the more reliable single-phase capacitors and (2) an inadequate surge capacitor maintenance capacitor replacement program.
Corrective actions were (1) replacing the capacitor and repairing the associated electrical circuitry, (2) replacing the three 12kV three-phase surge capacitors with more reliable single-phase capacitors, and (3) enhancing the surge capacitor preventive maintenance program to include periodic replacement of the 12kV and 4kV surge capacitors based upon testing and/or industry data.
There were no actual safety consequences due to the event as both the non-safety and safety-related systems and components responded as designed.

3.1.2 Special Report LERs

PG&E submitted one special LER during the reporting period.

1. DCPP provided

  1. The 90-day reporting of results of the Unit 2 2R13 steam generator (SG) Wstar (W*) alternate repair criteria (ARC) tubesheet inspections and calculated steam line break leakage from application of all ARC and non-ARC. There were 65 tubes (containing 70 single axial PWSCC indications) categorized as W* candidates left in service. The 120-day reporting of results of the Unit 2 2R13 SG Primary Water stress Corrosion Cracking (PWSCC) ARC inspections at dented tube support plate (TSP) intersections. There were 59 axial and 16 circumferential PWSCC indications detected at dented TSP intersections. These tubes were left in service with no corrective actions required because all satisfied conditioning monitoring burst margin requirements and leakage margin requirements.
  2. The 90-day reporting of results of Unit 2 2R13 SG voltage-based ARC inspections for TSP Outside Diameter Stress Corrosion Cracking (ODSCC). There were 65 tubes with ODSCC plugged during 2R13.

These indications were considered to be consistent with expectations.

3.1.3 Voluntary LERs

There were no voluntary LERs submitted by PG&E during this period.

3.1.4 Reactor Trips Reported in LERs

During the reporting period, there were two manual or automatic reactor trips reported in LERs.

In the past five DCISC reporting periods the following numbers of trips have occurred:

LERs table
Reporting Period Number of Trips
Automatic Manual
2002/2003 0 1
2003/2004 0 0
2004/2005 0 0
2005/2006 0 0
2006/2007 1 1

Although increasing, the number of trips continues to be low.

3.1.5 LER Trends

The following table depicts the LER history for DCPP for the last five DCISC reporting periods:

LER history
Time Period Number of LERs Submitted
7/1/02 – 6/30/03 11 (plus 0 voluntary LERs)
7/1/03 – 6/30/04  2 (plus 0 voluntary LERs)
7/1/04 – 6/30/05  3 (plus 0 voluntary LERs)
7/1/05 – 6/30/06  2 (plus 0 voluntary LERs)
7/1/06 - 6/30/07  3 (plus 0 voluntary LERs)

The DCISC considers this to be a good performance trend.

During the current reporting period, the reported events were reported within the requirement of within 60 days of event discovery. All three LERs were self-identified by PG&E. None was initially found by NRC.

The root cause of the LER was as follows:

Two LER root causes
Root Cause Number of Causes Percent of Total
Random single failure component failure 2 67%
Natural causes (dead birds) 1 33%

3.1.6 DCISC Evaluation and Conclusions

The DCISC recognizes that events will occur in any large complex system. The goal is to identify them and understand them, and take action to minimize the consequences and likelihood of any significant increase in risk. The design basis for nuclear power plants involves defense-in-depth. This recognizes that in real systems, unanticipated events will occur, so protective systems are designed to provide protection even if systems do not always perform as anticipated. For this reason, it is important to investigate events and to share information about them with other plants.

Each of the three Licensee Event Reports was investigated by PG&E to determine the plant conditions before and during the event, background and detailed event description, root cause and contributory causes, immediate and preventive corrective action, and previous LERs on identical or similar problems. No LER was significant enough to seriously affect operational safety. No significant cause code trends were observed. LER investigation reports were submitted to all DCISC Members and Consultants for review; PG&E reported on each LER at DCISC public meetings.

In previous reporting periods the largest contributor to LERs has been personnel error. The table below shows five-year LER personnel error history. The specific personnel errors have been as follows:

Personnel errors
Reporting Period Number of LERs No. Personnel Errors Percent Personnel Error
2002-2003 11 4  37%
2003-2004  2 2 100%
2004-2005  3 1  33%
2005-2006  2 1  50%
2006-2007  3 0 0

The number of DCPP License Event Reports (LERs) has remained consistently low at 2-to-3 per year. The DCISC considers this to be good performance.

DCPP LER investigations appeared generally adequate and corrective actions appeared to be appropriate for all LER events. There appears to be little or no recurrence of reportable events. The DCISC will continue to monitor LERs, their causes, and PG&E’s actions to correct and prevent them in future Fact-finding and public meetings.

3.2 NRC Inspection Reports

3.2.1 Discussion

The NRC performs inspections at each nuclear power plant. The purpose is to determine how well the plant operators are implementing and following NRC regulations, plant Technical Specifications, and other requirements, procedures, or commitments. Generally, better regulatory performance results in fewer inspections. NRC meets with the nuclear plant operator twice per year to review plant safety performance under the NRC Reactor Oversight Process (see Section 3.4 below). These meetings are usually public.

Inspections are performed by the plant Resident NRC Inspectors, inspectors from the NRC Region Office, experts from other NRC organizations, and NRC consultants. The bulk of inspections are routine, announced visits focusing on one or more specific areas of operation such as As Low As Reasonably Achievable (ALARA) radiation dose minimization program, maintenance, chemistry, security, operator examinations, or corrective actions. Special inspections are often made for investigation into previous events affecting plant safety and into special programs, such as NRC Generic Letter 89-10, Testing of Motor-Operated Valves.

Each inspection usually concludes with an exit meeting with licensee personnel, followed by a written inspection report. Inspections can result in the following categories of findings:

Unresolved Items
are items for which information is not yet available or awaiting licensee response or action.
Individual strengths
are used to point out good practices and weaknesses for the licensee’s attention for improvement and/or to prevent future problems.
Deviations
are variances from NRC regulations and/or licensee procedures or other requirements or commitments which are not as severe as outright violations.
Concerns,
typically including more than one individual weakness in a single area, are to alert the licensee to situations which could become violations if not corrected.
Non-cited Violations
are violations for which NRC credits the licensee for identifying the violation and/or for prompt, effective corrective action completed before or taken during the inspection. These are usually non-recurring, non-safety-significant items.
Violations
of NRC regulations, plant Technical Specifications, and other commitments, procedures, etc. require a formal response and corrective action. Violations carry four severity levels as described in Section 3.3, NRC Enforcement Actions.

Fewer violations generally mean better performance. Some in the industry think having a significant number of non-cited violations indicates an effective, aggressive regulatory program, meaning the licensee quickly finds and corrects its own problems/violations rather than the NRC finding them.

During the period July 1, 2006 - June 30, 2007, there were ten inspection reports received from the NRC for DCPP. This compares with 4, 9, 5, 8 and 7 in the previous five periods, respectively. PG&E’s regulatory performance with NRC has been good, and this generally means fewer inspections. The ten inspections during this period were as follows (the reported non-cited violations (NCVs) and findings are listed below separately in Section 3.3.1:

  1. Integrated Inspection (August 9, 2006) – the inspection covered problem identification and resolution (DCPP’s Corrective Action Program [CAP]). NRC concluded that DCPP CAP performance had improved in all areas and no findings of significance were identified; however, NRC noted finding a number of preventable repeat problems and of previously-documented NRC-identified and self-revealing findings.
  2. Inspection (August 10, 2006) – the inspection covered the preparation and construction of Independent Spent Fuel Storage Installation (ISFSI) components, namely the Cask Transport Facility, ISFSI pad, ISFSI cut slope stabilization and transport route. There were no violations identified during the inspection.
  3. Integrated Inspection (August 14, 2006) – the inspection covered radiation protection, emergency preparedness, operator requalification, and in-service inspections. There were four NRC-identified NCVs and one self-revealing findings – all were Green or very low safety significance.
  4. Security Inspection (September 13, 2006) – the inspection covered the DCPP security plan. There were no findings of significance identified.
  5. Integrated Inspection (October 31, 2006) – the inspection covered winter storm surge response, equipment alignments, fire protection, flood protection measures, licensed operator re-qualification, maintenance effectiveness, operability evaluations, post-maintenance testing, surveillance testing, temporary plant modifications, emergency plan changes, radiation safety, performance indicator verification, and problem identification and resolution. There were two NRC-identified findings of very low safety significance and two DCPP-identified findings. Both were considered to be non-cited violations.
  6. Radiation Safety Inspection (January 18, 2006) – the inspection covered radiation safety. NRC identified one self-revealing and one-DCPP identified NCVs which were Green or very low safety significance.
  7. Integrated Inspection (February 13, 2007) – the inspection covered fire protection, biennial heat sink performance, licensed operator re-qualification, maintenance effectiveness, maintenance risk assessments and emergent work control, operability evaluations, plant modifications, post-maintenance testing, surveillance testing, emergency plan exercise, identification and resolution of problems, and event follow-up for Unit 2 manual reactor shutdown and automatic reactor trip.
  8. Component Design Basis Inspection (February 23, 2006) – this inspection was performed on selected DCPP components to “. . . verify the initial design and subsequent modifications and operator actions to perform their design basis functions.” The NRC identified two non-cited violations of very low safety significance.
  9. Integrated Inspection (May 8, 2007) – the inspection covered adverse weather, equipment alignments, fire protection, flood protection measures, licensed operator re-qualification, maintenance effectiveness, maintenance risk assessments and emergent work control, operability evaluations, plant modifications, post-maintenance testing, surveillance testing, identification and resolution of problems, and event follow-up of wildland fire north of plant. The NRC identified two non-cited violations of very low safety significance.
  10. Material Control and Accounting Program Inspection (May 11, 2007) – the inspection covered the adequacy of DCPP measures taken to control the risk of less, theft, or diversion of Special Nuclear Material (SNM). NRC mentioned one DCPP-identified violation of very low safety significance.

3.2.2 DCISC Evaluation and Conclusions

The DCISC noted that there were no individual items or apparent significant new trends in Nuclear Regulatory Commission (NRC) inspections which would warrant additional recommendations or actions. Although the DCISC routinely follows-up on inspection report items in Fact-finding and public meetings, the DCISC plans no particular actions on NRC inspection reports, except as noted below in the discussion in Section 3.3, NRC Enforcement Actions.

3.3 NRC Enforcement Actions

3.3.1 Discussion

NRC considers items not in compliance with its regulations or with the licensee’s commitments or procedures to be violations. Corrective action is required for all violations. NRC identifies five severity levels for violations.

Level I is the most severe, representing the most significant regulatory concern which usually involves actual or high potential impact on the safety of the public. Level IV violations are more than minor concern and should be corrected so as to prevent a more serious concern.

Civil penalties (monetary fines) are usually imposed for Level I and II violations, are considered for Level III, and usually not imposed for Level IV violations. Most low-level violations are reported as Non-cited Violations provided the licensee places the violation into its corrective action program and provided the violation is not willful or repetitive. NRC has increased its scrutiny of corrective action programs. The categorization of violations in this report follows NRC’s actual classification in each notice of a violation.

During the period July 1, 2006 - June 30, 2007, NRC cited no Level I, II, III or IV violations and identified 20 non-cited violations. The identification breakdown for these is as follows:

14 NRC-identified
2 self-revealing
4 DCPP-identified

The history of violations for this and the last four DCISC reporting periods is as follows:

History of non-cited violations
DCISC Reporting Period Number of Inspections Violation Severity Level Violations
III IV Non-cited Total
7/1/01 – 6/30/02  4 - -  9  9
7/1/02 – 6/30/03  9 - - 12 12
7/1/03 – 6/30/04  5 - - 21 21
7/1/04 – 6/30/05  8 - - 31 31
7/1/05 – 6/30/06  7 - - 14 14
7/1/06 - 6/30/07 10 - - 20 20

PG&E has not received any Level I or II violations since the inception of the DCISC in 1990.

NRC Non-Cited Violations (NCVs)

During the period July 1, 2006 – June 30, 2007, NRC reported 20 non-cited violations (NCVs) at DCPP, including four identified by DCPP. These were considered "non-cited" because they satisfied the criteria specified in the NRC Enforcement Policy that either (1) PG&E identified the problem and corrected the root cause as a normal part of its Corrective Action Program before or during the NRC inspection visit, (2) the violation was minor enough to not warrant full violation status, or (3) they were part of NRC’s policy (see above) that it did not normally cite Level IV violations. They are reported, tabulated and analyzed, and any significant findings are disseminated for information. The 20 non-cited violations were as follows:

  1. A self-revealing NCV of 10CFR 50 Appendix B, Criterion XVI was identified for the failure of operations personnel to promptly identify a condition adverse to quality. Operators failed to document in the Corrective Action Program (CAP) an unexpected level drop in Accumulator 1-3. Failure to enter the occurrence into the CAP precluded actions that would have addressed similar conditions that resulted in a subsequent event involving an unexpected level drop and water hammer associated with Accumulator 2-3. NRC considered this a Problem Identification and Resolution Cross-cutting aspect and considered this NCV to be “Green” or of very low safety significance because there was no loss of actual safety function or exceedence of Technical Specification allowed outage time. (IR 05000275/2006003 and 05000323/2006003)
  2. An NRC-identified Technical Specification 5.4.1.a was identified for an inadequate reactor draindown procedure in which prior operating experience had not been incorporated that demonstrated that the level instruments would read correctly. The NRC considered this NCV to be “Green” or of very low safety significance because an optional set of instrumentation provided accurate level indication and there was no loss of reactor coolant. The violation had a crosscutting aspect in the area of human performance because operators failed to insure the adequacy of procedures. (IR 05000275/2006003 and 05000323/2006003)
  3. An NRC-identified NCV for the failure to follow the procedure for ensuring that welding preheat temperatures were verified prior to welding. The NRC considered this NCV to be “Green” or of very low safety significance because the plant was defueled at the time and the condition was evaluated prior to the plant entering Mode 5. (IR 05000275/2006003 and 05000323/2006003)
  4. NRC Inspectors identified an NCV of 10CFR50, Appendix B, Criterion XVI for the failure to prevent recurrence of similar failures of valve actuators in the Auxiliary Feedwater System. DCPP failed to adequately troubleshoot and provide for timely corrective actions regarding AFW control valves that failed due to high actuator torque switch resistance. This finding had cross-cutting aspects in the area of problem identification and resolution. The NRC considered this NCV to be “Green” or of very low safety significance because it did not represent an actual loss of safety function, represent an actual loss of safety function for a single train for greater than the Technical Specification allowed outage time, or screen as potentially risk significant due to seismic, fire, flooding, or severe weather initiating events. (IR 05000275/2006003 and 05000323/2006003)
  5. The NRC identified a NCV of 10 CFR 20.1501(a) because DCPP failed to survey to determine the extent and magnitude of radiation levels and evaluate the radiological hazards near two Chemical Volume Control System valves. The NRC considered this NCV to be “Green” or of very low safety significance because it was not an “as low as reasonably achievable” finding. There was no over-exposure or substantial potential for an over-exposure, and the ability to assess dose was not comprised. The violation had crosscutting aspects associated with human performance because resources were not established for the survey requirements. (IR 05000275/2006003 and 05000323/2006003)
  6. The NRC identified a NCV 10CFR50, Appendix B, Criterion XVI, "Corrective Action" for the failure to promptly identify a condition adverse to quality in that DCPP had prestaged the wrong equipment necessary to cross-connect the Fire Main Water System to the Auxiliary Feedwater System during a loss of core cooling event. The finding had cross-cutting aspects associated with human performance because DCPP did not ensure that the equipment needed to perform an Emergency Operating Procedure was available and adequate to assure nuclear safety. The NRC considered this NCV to be “Green” or of very low safety significance because the condition did not represent a loss of safety function of a single train for greater than its TS allowed outage time, did not represent an actual loss of one or more risk-significant non-TS trains of equipment for greater than 24 hours, and did not screen a potentially risk-significant due to seismic, flooding, or severe weather. (IR 05000275/2006004 and 05000323/2006004)
  7. NRC identified a NCV of 10 CFR 50.65(b) for the failure of engineering to include the Auxiliary Feedwater Pump (AFWP) room floor drains within the scope of DCPP’s program for monitoring the effectiveness of maintenance. The floor drains are credited in the flood analysis. The NRC considered this NCV to be “Green” or of very low safety significance because the condition did not represent a loss of system safety function, did not represent an actual loss of safety function of a single train for greater than its TS allowed outage time, and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. The finding had a crosscutting aspect in the area of problem identification and resolution associated with operating experience because engineering personnel did not effectively incorporate pertinent industry operating experience into their program for evaluating the effectiveness of maintenance performed on AFWP room floor drains. (IR 05000275/2006004 and 05000323/2006004)
  8. The NRC identified a self-revealing NCV of 10CFR520.1501(a) that resulted in an unconditional release of radioactive material from the radiologically controlled area (RCA). The contents of a vehicle cab were not removed and surveyed, resulting in the release of a contaminated safety harness from the RCA; however, the harness remained in the RCA. DCPP determined the inadequate survey was caused by an inadequate procedure which was revised. The finding had human performance cross-cutting aspects in that DCPP did not have sufficiently detailed procedures. The violation had very low safety significance because (1) the finding was a radioactive material control finding, (2) it was not a transportation finding, (3) it did not result in public does greater than 0.005 Rem, and (4) radioactive material was not released. (IR 05000275/2006013 and 05000323/2006013)
  9. The NRC cited a self-revealing NCV of 10CFR50, Appendix B, Criterion III, “Design Control,” for the failure to apply adequate design control measures. Specifically, DCPP engineering failed to account for thimble tube chrome-plated bands at the fuel assembly bottom nozzle/lower core plate interface, resulting in mispositioning of a tube thereby increasing through-wall wear. A tube rupture occurred. The finding was considered to be of very low safety significance because the worst-case leakage from a thimble tube was less than the RCS identified leakage limit. The finding had a crosscutting aspect in the area of problem identification and resolution because DCPP removed a corrective action to prevent recurrence of significant thimble tube wear. (IR 05000275/2006005 and 05000323/2006005).
  10. The NRC identified a non-cited violation of 10CFR50, Appendix B, Criterion XVI, “Corrective Actions,” for the failure to promptly correct a condition adverse to quality. Specifically, DCPP implemented a temporary modification to a vital battery contrary to requirements of ANSI/IEEE standards. The modification adversely affected battery surveillance tests. The NRC determined that the finding was of very low safety significance because it did not represent an actual loss of safety function of a single train for greater than its TS allowed outage time. The finding had a crosscutting aspect in the area of problem identification and resolution associated with the DCPP CAP in that engineering did not thoroughly assess the operability of the battery and correct a condition adverse to quality in a timely manner. (IR 05000275/2007005 and 050003232006005).
  11. The NRC identified a non-cited violation of 10CFR50, Appendix B, "Design Control," for failure of engineering to apply adequate design control measures in that the acceptance criteria for the Auxiliary Saltwater Pump surveillance test were changed from greater than zero packing leak-off to zero packing leak-off contrary to vendor documentation. The NRC determined that the finding was of very low safety significance because it did not represent an actual loss of safety function of a single train for greater than its TS allowed outage time. (IR 05000275/2006005 and 05000323/2006005).
  12. An NRC-identified non-cited violation of 10CFR50, Appendix B, Criterion XVI, “Corrective Actions” was reported for failure of engineering to promptly identify and correct a condition adverse to quality. Specifically, on two occasions engineering and operations (1) failed to address operability when using manual controls in place of automatic controls associated with the Auxiliary Building Ventilation System and (2) failed to fully address the impact of debris between the circuit card and the panel connections of the Auxiliary Building Ventilation System. The finding was considered to be of very low safety significance because it only represents a degradation of the radiological barrier function provided for the Auxiliary Building. The finding had a crosscutting aspect in the area of problem identification and resolution. (IR 05000275/2006005 and 05000323/200605).
  13. NRC identified a NCV for a violation of 10CFR50, Appendix B, “Design Control,” for the failure to translate design basis information into specifications and procedures in that a non-conservative flow rate was used as input in engineering design calculations resulting in the potential for choked flow at the discharge valves for the Unit 1 Auxiliary Service Water System. The NRC considered this NCV to be “Green” or of very low safety significance because the condition did not represent a loss of system safety function, did not represent an actual loss of safety function of a single train for greater than its TS allowed outage time, and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. (IR 05000275/2006011 and 05000323/2006011)
  14. The NRC identified a non-cited violation of 10CFR50, Appendix B, “Design Control” for the failure to demonstrate that the acceptance criteria for surveillance tests had conservatively accounted for instrument uncertainties in determination of the minimum allowed ultimate heat sink temperature. The NRC considered this NCV to be “Green” or of very low safety significance because the condition did not represent a loss of system safety function, did not represent an actual loss of safety function of a single train for greater than its TS allowed outage time, and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. (IR 05000275/2006011 and 05000323/2006011)
  15. A non-cited violation of 10CFR50, Appendix B, "Design Control" was identified by NRC for failure of engineering to appropriately update the heat dissipation calculation for the Vital 480V switchgear rooms. Specifically, since 1994, the heat dissipation calculation had not been updated with changes in analyzed bus electrical loading. The calculation was input to other ventilation calculations to determine air flow balancing to 480V switchgear and inverter rooms. The NRC considered this NCV to be “Green”’or of very low safety significance because the condition did not represent a loss of system safety function, did not represent an actual loss of safety function of a single train for greater than its TS allowed outage time, and did not screen as potentially risk-significant due to seismic, flooding, or severe weather initiating events. (IR 05000275/2007002 and 05000323/2007002)
  16. An NRC-identified non-cited violation of 10CFR50, Appendix B, “Design Control” was determined for the failure of maintenance personnel to make modifications to the Control Room condenser filter mount consistent with the component’s design documentation and applicable procedure. Specifically, maintenance personnel used vice-grips, C-clamps and plastic tie-wraps to secure in-place the filter mount, which was significantly corroded. The modification had not been documented or analyzed at the time it was placed in service, and after subsequent engineering reviews, the condenser was considered inoperable due to the loss of seismic qualification. NRC considered the finding to be of very low safety significance because it did not represent degradation of the barrier function of the control room against radiological hazards, smoke, or toxic atmosphere. The finding has a crosscutting aspect in the area of problem identification and resolution associated with the DCPP CAP in that maintenance personnel failed to adequately identify the degraded condition of the control room condenser when it was initially discovered. (IR 05000275/2007002 and 05000323/2007002).

All of the non-cited violations were classified “Green” by the NRC, meaning that they were of very low safety significance (see Section 3.4 below for the NRC color ratings system).

A tracking spreadsheet is maintained by DCPP Quality Verification (QV) for all NRC violations, NOVs and NCVs, to ensure the issue identified by the NRC is adequately addressed. An AT-NCV Action Request (AR) is initiated for each potential NCV at the exit NRC inspection interview, and appropriate Corrective Action Program (CAP) documents are initiated and their status is reviewed and verified periodically, typically biweekly, through the resolution period. PG&E believed that the NRC’s implementation of its Reactor Oversight Process (ROP) has increased the numbers of NCVs, which do not require a formal response, and reduced the numbers of NOVs, which are reserved under the new Reactor Oversight Process for risk-significant issues.

NRC violations are included in the CAP Trending Program and are not trended separately. An Event Trend Record (ETR) is issued for each NCV associated with an AT-NCV AR. Periodic evaluation of the ETRs is undertaken to identify adverse trends.

The DCISC notes that NRC has increased its application of the cross-cutting aspects of NCVs. The highest number of NCVs is in the Mitigating Systems Cornerstone with Occupational Radiation Safety second. The following table shows DCPP NCV performance compared with other NRC Region IV plants:

Seven cornerstones of safety table
Year* DCPP NCVs Region IV Average NCVs
2002/2003 16 11
2003/2004 22 12
2004/2005 25 12
2005/2006 18 11
2006/2007 16 16

*DCISC Reporting periods

PG&E-Identified Violations

The following four items were initially identified by PG&E and captured in their Corrective Action Program. NRC reviewed the items and determined they met the guidance for Green (very low safety significance) non-cited violations (NCVs).

  1. The NRC identified a DCPP-identified violation of 10CFR50, Appendix B, Criterion XVI, “Corrective Action,” for failure to initiate a prompt operability assessment (POA) following the failure of a Hydrazine mixing check valve (Class 1 boundary valve) to reposition to the closed position following a surveillance test. Upon discovery, Operations closed the upstream manual isolation valve and issued a POA declaring the valve operable. The NRC considered this NCV to be “Green” or of very low safety significance because the condition did not represent a loss of system safety function, did not represent an actual loss of safety function of a single train for greater than its TS allowed outage time, and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. (IR 05000275/2006004 and 05000323/2006004).
  2. A DCPP-identified NCV of 10CFR50, Appendix B, Criterion III, “Design Control,” was identified for the failure of engineering personnel to recognize that the NRC had accepted the inner Containment sump screen for redundancy for trapping a fiber mat that may build on the screen. Engineering later discovered the error and assessed the current configuration as operable because the lack of materials actually present that could penetrate the outer sump screen. The NRC considered this NCV to be “Green” or of very low safety significance because it was a design deficiency confirmed not to result in a loss of function. (IR 05000275/2006004 and 05000323/2006004).
  3. The NRC cited a DCPP-identified violation of NRC NUREG-1600 for failure to control a radioactively-contaminated wrench and six Thermo-Luminescent Dosimeters (TLDs) in that the items were found outside the RCA but did not leave the Protected Area. The violation had very low safety significance because (1) the finding was a radioactive material control finding, (2) it was not a transportation finding, (3) it did not result in public dose greater than 0.005 Rem, and (4) radioactive material was not released. (IR 05000275/2007005 and 05000323/2006005).
  4. The NRC reported as a non-cited violation a 2004 DCPP-identified finding that an adequate required annual physical inventory of Special Nuclear Material (SNM) had not been performed. Specifically, the Spent Fuel Pool Fuel Rod Storage Container (FRSC) had been considered a single item rather than a collection of rods in the container. The procedure was revised, and subsequent inventories had been performed properly. The NRC considered this to be of very low safety significance because at the time of the inspection, there was reasonable assurance that all SNM was accounted for and stored in approved safe locations. Because the violation was self-identified and corrected and because all physical inventories of SNM contained in fuel were adequately performed, the NRC exercised enforcement discretion to not issue enforcement action. (IR 05000275/2007201 and 05000323/2007201).

NRC Findings

The NRC issued no findings to DCPP during the period.

3.3.2 DCISC Evaluation and Conclusions

The numbers of NRC inspections in this period and the previous three periods have been eight, seven, and ten respectively. This low number is a result of good regulatory performance as measured primarily by NRC Performance indicators (see Section 3.5 below). The number of NRC-cited violations remains zero as in the previous periods (see table of five-year inspection violation history in Section 3.3.1). The number of non-cited violations for this period is lower than past trends at 16. In addition to the 16 NCVs, there were 4 PG&E-identified violations (not issued as NCVs by NRC) making a total of 20 violations. The annual amount of 20 appears to the DCISC to be a typical number for DCPP, which compares favorably to the average 12-month rolling total of TBD for NRC Region IV plants at the end of June 2007. The DCISC plans to review the number and nature of these items during the next period.

The DCISC noted that NRC considered the violations have cross-cutting aspects in two areas: (1) human performance (8 occurrences) and (2) problem identification and resolution (8 occurrences). NRC identified more than three Green inspection findings with documented crosscutting aspects involving the adequacy of [engineering] design documentation and procedures. NRC stated that DCPP recognized this and implemented actions to address it and concluded that a substantive cross-cutting issue did not exist.

The DCISC noted that there were ten violations in the area of design control:

  1. Failure to include flood analysis for drain design on the Auxiliary Feedwater Pump room.
  2. Failure to account for In-core Instrument thimble tube chrome-plated bands in tube placement, resulting in a tube rupture.
  3. Failure to incorporate Institute of electrical and Electronics Engineers (IEEE) standards in a vital battery calculation.
  4. Failure to apply adequate design control measures in the shaft packing leak-off acceptance criteria in an Auxiliary Saltwater Pump surveillance test.
  5. Failure of engineering to promptly identify two conditions adverse to quality: control valve operability and circuit card contract debris.
  6. Failure to translate design basis information into specifications and procedures, resulting in non-conservative flow rates used in calculations.
  7. Failure to incorporate instrument uncertainty into acceptance criteria for surveillance tests.
  8. Failure to appropriately update a heat dissipation calculation for vital 480V switchgear rooms.
  9. Failure to make modifications to Control Room condenser filter mounts consistent with the component design basis document and applicable procedure.
  10. Failure of engineering personnel to recognize that the NRC had accepted the inner Containment sump screen for redundancy for trapping fibrous material. (DCPP-identified)

The DCISC has concerns about the high percentage of design control NCVs and will follow up during the next reporting period.

The DCISC heard presentations by PG&E on each non-cited violation, finding and LER at public meetings and has reviewed each cited violation and PG&E’s corrective actions, where applicable. PG&E corrective actions appeared adequate. There were no individual items of significance to warrant DCISC recommendations or actions.

The DCISC notes there were no NRC violations during this period, a continuing positive trend. The number of NRC non-cited violations (NCVs) continues to trend lower and was lower than the average for NRC Region IV. The ratio of NRC- to DCPP-identified violations (16:4) was higher than desirable. Half (10 of 20) of the NRC- and DCPP-identified violations involved design control or engineering-related design control failures. This is a matter of concern to the DCISC. The DCISC agrees with the very low safety significance determinations of the individual NCVs but would like to see the number continue to drop.

3.4 NRC Performance Evaluations

The Nuclear Regulatory Commission (NRC) inspection, assessment, and enforcement programs for commercial nuclear power plants takes into account improvements in the performance of the nuclear industry over the past 25 years and improved approaches of inspecting and assessing safety performance at NRC-licensed plants.

The NRC Revised Reactor Oversight Process (RROP) monitors licensee performance in three broad areas (called strategic performance areas):

  1. Reactor Safety (avoiding accidents and reducing the consequences of accidents if they occur)
  2. Radiation Safety (protecting plant employees and the public during routine operations)
  3. Safeguards (protecting the plant against sabotage or other security threats).

The process focuses on licensee performance within each of “Seven Cornerstones” of safety in the three areas:

Cornerstones of Safety
Reactor Safety Radiation Safety Safeguards
Initiating Events
Mitigating Systems
Barrier Integrity
Emergency Preparedness		 Occupational
Public		 Physical Protection

To monitor these Seven Cornerstones of safety, the NRC uses two processes that generate information about the safety significance of plant operations:

  1. Inspections
  2. Performance Indicators

Inspection findings are evaluated according to their potential significance for safety, using the significance determination process, and assigned colors of GREEN, WHITE, YELLOW, or RED.

  • GREEN findings are indicative of issues that, while they may not be desirable, represent very low safety significance.
  • WHITE findings indicate issues that are of low to moderate safety significance.
  • YELLOW findings are issues that are of substantial safety significance.
  • RED findings represent issues that are of high safety significance with a significant reduction in safety margin.

Performance Indicator data are compared to established criteria for measuring licensee performance in terms of potential safety. Based on prescribed thresholds, the indicators will be classified by color representing varying levels of performance and incremental degradation in safety: GREEN, WHITE, YELLOW, or RED.

  • GREEN indicators represent performance at a level requiring no additional NRC oversight beyond the baseline inspections.
  • WHITE corresponds to performance that may result in increased NRC oversight at the Resident Inspector or Regional level.
  • YELLOW represents performance that minimally reduces safety margin and requires even more NRC oversight at the NRC Region level.
  • RED indicates performance that represents a significant reduction in safety margin but still provides adequate protection to public health and safety. NRC response at the Agency level could include public meeting, utility-developed performance improvement plan, and/or special inspection team.

The assessment process integrates performance indicators and inspection so the agency can reach objective conclusions regarding overall plant performance. The NRC uses an Action Matrix to determine in a systematic, predictable manner which regulatory actions should be taken based on a licensee’s performance. The NRC’s actions in response to the significance (as represented by the color) of issues will be the same for performance indicators as for inspection findings. As a licensee’s safety performance degrades, the NRC will take more and increasingly significant action, which can include shutting down a plant, as described in the Action Matrix.

The NRC Performance Indicators (PIs) for DCPP through the second quarter 2007 are depicted in Table 3.1 at the back of Section 3.0.

The redesigned NRC inspection program uses a risk-informed approach to select areas of the plant to inspect within each cornerstone. The selection is based on potential risk, past operational experience, and regulatory requirements.

Each calendar quarter, NRC inspectors and the regional office review plant performance indicators and inspection findings. Each year, NRC regional and headquarters offices make a final review, to include a more detailed assessment of plant performance over the 12-month period, preparation of a performance report, and preparation of a six-month inspection plan. The report is sent to each plant and discussed in a public meeting.

NRC End-of-Cycle Report for 2006

NRC generated two performance review and assessment letters for DCPP as follows:

1. Annual Assessment Letter (March 2, 2007)

NRC reported that for the period January 1 through December 31, 2006
"Overall, Diablo Canyon Power Plant, Units 1 and 2, operated in a manner that preserved public health and safety and fully met all cornerstone objectives. Plant performance for the most recent quarter, as well as for the first three quarters of the assessment cycle, was within the Licensee Response Column of the NRC’s Action Matrix, based on all inspection findings being classified as having very low safety significance (Green) and all PIs [Performance Indicators] indicating performance at a level requiring no additional NRC oversight (Green). Therefore, we plan to conduct reactor oversight process baseline inspections at your facility."

The DCISC understands this to mean acceptable regulatory performance and no increased inspections above baseline.

In its March 3, 2004 Annual Assessment Letter the NRC identified a substantive cross-cutting issue in the area of Problem Identification and Resolution. This meant that DCPP’s Corrective Action Program (CAP) was not meeting NRC effectiveness expectations. The issue was eliminated in 2005 based on improvements in DCPP’s CAP. In the August 30 letter NRC noted:
“The problem identification and resolution substantive cross-cutting issue was of concern because inspection findings, particularly those involving long-standing degraded conditions and the adequacy of evaluations continued to be identified subsequent to the implementation of corrective action program improvements. Based on our review of your recent performance in this area, we have noted improvements in your staff implementation of the corrective action program. Specifically, we noted that the quality of evaluations and the effectiveness of resulting corrective actions taken to address equipment reliability issues have improved. We also noted your continuing efforts to sustain improvement in the implementation of the site corrective action program.”
NRC also reported the following:
The NRC staff identified that there were more than three Green inspection findings for the current 12-month assessment period with documented crosscutting aspects in the human performance area. There is a crosscutting theme associated with these findings involving the adequacy of design documentation and procedures. In evaluating the scope of efforts and progress in addressing the crosscutting theme, we determined that your staff previously recognized this theme and are continuing to implement effective actions to address it. In March 2005, your staff initiated a comprehensive 5-year plan to improve the quality of plant procedures. Most of the findings identified during this assessment period were associated with procedures that had not been enhanced as part of this long-term improvement initiative. In addition to the procedure upgrade program, your staff initiated an apparent cause evaluation of the most recent findings. At this time, we have determined that your current and planned corrective actions are effective in mitigating this crosscutting theme. Therefore, the NRC staff has concluded that a substantive crosscutting issue in human performance does not exist at this time.

2. Annual Security Assessment Letter (March 2, 2007)

NRC reported that for the period January 1 through December 31, 2006
With respect to physical protection and security, Diablo Canyon Power Plant, Units 1 and 2, were operated in a manner that preserved public health and safety, promoted the common defense and security, and fully met the cornerstone objective. Plant performance for the fourth quarter, as well as for the first three quarters of the assessment cycle, was within the Licensee Response column of the NRC’s security action matrix because all inspection findings had very low security significance (Green) and all performance indicators indicated that no additional NRC oversight was required (Green). Therefore, we will conduct security baseline and temporary instruction inspections at your facility through September 30, 2008.

The letter also reported that NRC identified more than three Green inspection findings with documented crosscutting aspects involving the adequacy of [engineering] design documentation and procedures. NRC stated that DCPP recognized this and implemented actions to address it and concluded that a substantive cross-cutting issue did not exist.

NRC Public Meeting to Discuss DCPP Annual Assessment

On May 17, 2007 the NRC provided notice of a June 26, 2007 public meeting to present to the public the above NRC assessments of Diablo Canyon safety performance, to describe other NRC focus areas for the period January 1 through December 31, 2006, and to receive public questions and comments.

The DCISC concurs with the NRC assessment that, overall, DCPP “. . . operated in a manner that preserved public health and safety. . .” and shares its concerns regarding the problem identification and resolution remarks, noting, however, that their impact had very low safety significance.

3.5 DCISC Meeting with NRC Resident Inspectors

The DCISC Fact-finding Team met with Terry Jackson, NRC Senior Resident Inspector, on March 21-22, 2007 to discuss items of mutual interest. The DCISC last met with Tim McConnell, the new NRC Resident Inspector, in November 2005. At that time the DCISC concluded the following:

DCISC’s meeting with DCPP’s new NRC Resident Inspector revealed no new significant issues or concerns. The two significant NRC cross-cutting issues on Human Performance and Problem Identification & Resolution had been resolved prior to this meeting.

Mr. Jackson has been a Resident at DCPP since 2002 and Senior Resident since 2005. He reported that he had noticed some weaknesses during and following Outage 2R12 regarding long-standing equipment issues. There were issues in the Problem Identification and Resolution (P&IR) area which led to NRC’s decision to issue a Substantive Cross-cutting issue in that area. This led DCPP to issue a Nonconformance Report (NCR) to investigate the causes.

Mr. Jackson has been looking at check valve back leakage of some second-off valves from the Reactor Coolant System (RCS). The valves have had little maintenance until recently. The leakage permitted Nitrogen gas to enter lower pressure systems, come out of solution and create voids. DCPP is replacing the valves in 1R14, possibly with soft-seat valves.

In its recent period assessment letter NRC reported six non-cited violations because of procedure quality problems. In some cases DCPP did not incorporate operational experience in the procedures.

During 2002 through 2005, there were many self-revealing violations which DCPP first identified. It appeared to Mr. Jackson that there are not now so many self-revealing violations and that NRC is identifying most of them first, but this is not yet a negative trend.

He reported that operations evaluations are spotty in quality, varying much from evaluator-to-evaluator. Sometimes there is a less than desirable basis for Operations operating decisions. He believed that this is not yet a big problem, but that it should be improved. Mr. Jackson reported that human performance improvements are being made. DCPP is reducing human errors to a low level of significance. Mispositioning performance has improved. He has seen a positive change in culture due to new management challenging the status quo. NRC is observing safety culture activities and results at the plant more.

Conclusion:
In meeting with the DCISC, the DCPP Senior Resident Inspector had no new concerns of significance to report. He had seen positive changes in the DCPP culture with the new management team.

3.6 DCISC Evaluation and Conclusion

The Committee notes that the Nuclear Regulatory Commission (NRC) concluded that, “Overall, Diablo Canyon Power Plant, Units 1 and 2, operated in a manner that preserved public health and safety and fully met all cornerstone objectives. Plant performance for the most recent quarter was within the Licensee Response column of the NRC’s Action Matrix, based on all inspection findings being classified as having very low safety significance (Green) and all PIs [Performance Indicators] indicating performance at a level requiring no additional NRC oversight (Green).”

The DCISC received regular reports on the NRC Performance Indicators, DCPP License Event Reports (LERs) sent to NRC, and NRC Inspection Enforcement Actions (violations) at each of its Public Meetings as well as copies throughout the reporting period. The DCISC will continue to monitor LERs, NRC inspection reports and the NRC Performance Indicators at both Fact-finding and public meetings.

Half (10 of 20) of the violations involved design control or engineering-related design control failures. This is a matter of concern to the DCISC, and the DCISC will follow up during the next reporting period.

The DCISC notes there were no NRC-cited violations during this period and fewer NRC non-cited violations (NCVs) than in some previous periods, both positive trends. The ratio of NRC- to DCPP-identified violations (16:4) was higher than desirable. The DCISC agrees with the very low safety significance determinations of the individual NCVs.

For more information about DCISC contact:

Diablo Canyon Independent Safety Committee
  Office of the Legal Counsel
857 Cass Street, Suite D, Monterey, California 93940
Telephone: in Califonia call 800-439-4688; outside of California call 831-647-1044
Send E-mail to: dcsafety@dcisc.org