4.0 Summary of Major DCISC Review Topics, 18th Annual Report - July 1, 2007 thru June 30, 2008

4.17 Outage Management

4.17.1 Overview and Previous Activities

The DCISC monitors PG&E’s outage plans, actions, and results in the following ways:

PG&E completed its thirteenth Unit 1 & 2 refueling outages (1 R13 & 2R13) during the 2005-2006 reporting period. Since the DCISC began review of this subject in 1990, outage management performance has steadily improved as shown in the table below. PG&E expects its outages can routinely run in the high-twenty to low-thirty day range.

Other outage indicators also are showing continuous improvement. With the exception of anomalous 1R9 radiation levels, radiation exposure and personnel injuries have been generally declining in the last three outages as follows:

Outage indicators table
Outage Duration (days) Collective Radiation Exposure
(person-Rem)		 Personnel Safety
(recordable injuries)
Outage Unit 1 Unit 2 Unit 1 Unit 2 Unit 1 Unit 2
R10 40.4 29.5 162 108 2 1
R11 30.3 33.5 138 125 4 2
R12 77.5 57.0 149 98 5 0
R13 41.0 38.8 116 74 5 3
R14 29.8 68.9 103 226 6 3

The DCISC reviewed the following during the previous reporting period.

The DCISC concluded in the previous period that the DCPP 1R14 Outage Refueling Outage results were good with relatively low collective personnel radiation dose and human error, and no significant nuclear safety concerns or events.

4.17.2 Current Period Activities

The DCISC reviewed the following outage-related topics during the current period:

Outage 1R14 Safety Schedule Changes

The DCISC Fact-finding Team met with Dennis Peterson, DCPP Outage Director, at the August 2007 Fact-finding Meeting (Volume II, Exhibit D.2, Section 3.5) to discuss the three outage safety plan changes in Outage 1R14. The DCISC initially became aware of these changes at its June 13-14, 2007 Public Meeting (Reference 6.6).

DCPP prepares an Outage Safety Plan for each refueling outage. The purpose of the plan is to provide information on outage safety requirements and risk areas for the plant staff. In order to assess outage safety impact, the Outage Safety Plan is referred to prior to making major schedule changes. Plants can be susceptible to variety of situations challenging safety during shutdown conditions. PG&E’s outage safety program is designed around three major concepts:

The Outage Safety Plan provides background information for the logic contained in the Outage Safety Checklists. The checklists provide the logic used to develop the Outage Safety Schedule. The Outage Safety Schedule, along with the checklists, ensures the equipment and plant conditions assumed in the shutdown abnormal procedures are met. The shutdown abnormal procedures provide guidance for providing passive core cooling and key system restoration.

There were five potential changes from the 1R14 Outage Safety Schedule as follows:

  1. Offsite start-up power was lost for approximately 55 minutes to both units due to a failed Morro Bay-to-DCPP transmission line. Unit 1 was shutdown in No Mode with all fuel offloaded to the Spent Fuel Pool (SFP). Auxiliary power had been cleared for maintenance, and 230 kV offsite power was supplying start-up power. Emergency Diesel Generator (DG) (EDG) 1-3 was cleared for maintenance. Upon loss of 230kV/start-up power, EDGs 1-1 and 1-2 auto-started as designed and powered their loads. SFP cooling, not automatically loaded on the EDGs, was manually started by the operators within five minutes per procedures. Unit 2 continued to operate normally. There was a six-hour schedule impact on the outage but no safety impact on the plant.
  2. The Outage Safety Schedule included all three EDGs as available for core reload; however, due to emergent maintenance, EGD 1-3 was not available. All other on-site and off-site power sources were available and energized with required cross-tie capability. There were no impacts on outage activities or plant safety.
  3. The Outage Safety Schedule had all three EDGs and both off-site power sources required for reduced Reactor Coolant System (RCS) inventory Refueling Outage procedure in which, after shutdown and a cooling period, reactor coolant is lowered below the hot and cold legs, permitting work to be performed in a relatively dry environment. The operation is a relatively high-risk condition due to the potential for loss of cooling.">mid-loop operation; however, EGD 1-3 had not been returned to service. Except for EGD 1-3, all power sources were available and energized. Station management approved continuing outage activities with EGD 1-3 out-of-service. There were no impacts on outage activities or plant safety.
  4. EGD 1-2 was declared inoperable for approximately 85 minutes due to an oil leak on a pressure gauge fitting on the lube oil system. The fitting was tightened and the EGD declared operable following a satisfactory test run. All other required power sources were available and energized. There were no impacts on outage activities or plant safety.
  5. A Component Cooling Water (CCW) System isolation valve was found to be leaking, temporarily adversely affecting the ability to provide clearance on CCW Pump 1-3. The valve was repaired and the system returned to service. There were no impacts on outage activities or plant safety.

Each of the changes was documented, evaluated, and approved by plant staff and management prior to proceeding with work affected by the change. The evaluation includes comparison with the Outage Safety Checklists for required operable and available components and systems. Except for the first change above (loss of off-site start-up power), none of the changes reduced Defense-in-Depth (DID) below the checklist minimum. The first change was evaluated as reducing DID below checklist values but having little or no safety impact. The event will be added to the Outage 2R14 Safety Plan as operating experience.

DCPP performed a Root Cause Analysis (RCA) of the first change (loss of off-site power). The root cause was identified as the 230kV loop relay protection scheme not being designed for an unanticipated voltage transient in the output signal supplying the protection relay for a breaker. A contributory cause was an in-service failure of a non-ceramic insulator near the Morro Bay Switchyard. The Corrective Action to Prevent Recurrence (CAPR) was to reconfigure the breaker protection relays in the Mesa and DCPP switchyards to establish a 16 millisecond (one cycle) time delay in their protection scheme. This has been completed.

The unanticipated voltage transient has been determined to be applicable only to the DCPP and Mesa switchyards. The cause does not apply to the 500kV system.

The DCISC Fact-finding Team believes these evaluations and corrective actions were appropriate.

The five potential Outage 1R14 Safety Schedule Changes were documented, evaluated and dispositioned properly by DCPP staff and management as required by DCPP procedures. The evaluations appeared to be satisfactorily analyzed for safety impact. One of the five changes actually reduced Defense-in-Depth below pre-determined conservative checklist minimums, but it had no actual safety impact on the plant.

Refueling Outage 2R14 Major Work, Safety Plan & Results

The DCISC Fact-finding Team (FFT) met with DCPP and received a presentation on DCPP Refueling Outage 2R14 at the following meetings:

The 2R14 overall outage results were as follows:

2R14 Goals 2R14 Results
Zero Nuclear Safety Events Zero
Zero Disabling Injuries Zero
> 2 Recordable Injuries Three
Zero Human Performance Clock Resets Zero
Zero FME Significant Events Zero
> 66 Days Breaker to Breaker Schedule Duration 68 Days 22.5 Hours
> 5 Day Power Ascension @lt; 5 Days
100 Power < 90 Days Met
Dose Goal 256.7 Person-Rem 226 Person-Rem

Although DCPP did not attain its outage duration or recordable injury goals, the overall outage was considered by DCPP and the FFT to be successful. Notably, there were no nuclear safety events. There were 18 changes to the Outage Safety Schedule, one of which was considered notable but not a safety concern.

Major work items in the outage were as follows:

The Outage Safety Schedule Change Process is a controlled method to handle changes to the pre-established outage schedule for having safety equipment and sources available. The one notable change was isolation of Component Cooling Water to the Containment Fan Cooler Units (CFCUs) which could impact containment temperatures affecting personnel working conditions; however, no safety problems were identified. Examples of other changes were:

The DCISC believed this process was effectively utilized and was in part responsible for DCPP’s maintaining nuclear safety throughout the outage.

As is typically done following each outage, DCPP developed a list of “lessons-learned” to enhance performance in future outages. The 2R14 list included the following major lessons learned:

Approximately 200 days into Cycle 14 prior to the outage, Chemistry identified and monitored a fuel failure according to DCPP procedures. Radiochemistry results indicated one or two open fuel defects. This resulted in ramp rate restrictions to protect the fuel, plans for outage in-mast fuel sipping and fuel repair activities, and enhanced 2R14 foreign material search provisions. Following the sipping, one once-burned core peripheral assembly was found to have a single fuel rod with a defect. The defect consisted of a corner rod separated into two pieces; however, field inspections could not identify the failure mechanism. Because of the nature of the defect, the assembly could not be re-constituted and was removed from service. This required a minor core redesign and the use of six additional assemblies to maintain core symmetry with the affected assembly removed. The redesign resulted in a small impact on full power capability and minor adverse changes to core operating limits (peaking factor margin reduction by 1.7 and Axial Flux distribution initially more negative by 0.6).

The fuel vendor, Westinghouse, and DCPP are performing a root cause analysis expected to be complete by June 2008. In addition to the fuel defect, DCPP discovered unexpected internal wear in an instrument guide tube which was also under investigation. The DCISC should request a presentation on this analysis and investigation at its June 25-26, 2008 Public Meeting.

DCPP’s Outage 2R14, an outage of unusually large scope with four Steam Generator replacements, was performed successfully overall. As expected by DCPP, a single fuel rod defect was identified as well as an instrument guide tube with unexpected internal wear. Both are under investigation which the DCISC should continue to monitor.

Outage 2R14 Containment Integrated Leak Rate Test

The DCISC Fact-finding Team met with Meagan Wilson, Associate Engineer and Containment Integrated Leak Rate Test (ILRT) Day Shift Test Lead,at the May 20-22, 2008 Fact-finding Meeting (Volume II, Exhibit D.10, Section 3.12) to review the ILRT completed in Outage 2R14. This is the DCISC’s first review of this test.

The DCPP Containment, like all nuclear power containments, is designed to provide near-leak-tight containment of any radiological materials in the building which might otherwise escape during normal, off-normal or accident conditions. NRC regulation 10CFR50, Appendix J contains requirements for containment leak rate testing generally on a ten-year basis; however, certain exceptions are permitted. The last test of DCPP Unit 2 was 15 years ago due to a five-year extension by NRC. NRC is considering going to a standard 15-year period based on the good performance of the industry in maintaining containment leak tightness. The test was performed in accordance with industry standard ANSI/ANS (American National Standards Institute/American Nuclear Society) 56.8-1994, “Containment system leakage Testing Requirements (Absolute Method, Mass Point Analysis, Leakage Stabilization Criteria, Termination Criteria).”

The DCPP Containment contains a net free volume of 2.55 million cubic feet and has a design pressure of 47 psig. The Containment has a Technical Specification maximum design basis leak rate of 0.1 weight /day used for accident calculations.

The ILRT was performed by ILRT, Inc., a specialist in ILRT testing, and required 42 hours (vs. a projected 36 hours) and included the following steps:

Step Time Required (hours)
Pressurization 7.5
Stabilization/Troubleshooting 16.08
ILRT itself 10.00
Verification Test 4.00
Depressurization delay 0.75
Depressurization 3.00
Unrestricted access restored 0.66

Pressurization was begun at 1810 hours on April 2, 2008 at an average pressurization rate of 8 psi/hour using 17 compressors with a rated capacity of 27,500 cfm. The test was performed at almost 46 psig (end of test) with the following results:

The DCISC Fact-finding Team received a copy of the test report “Periodic Reactor Containment Building Integrated Leakage Rate Test Final Report,” dated April 3&4, 2008. The report was thorough and informative. The test team generated many lessons-learned to improve the Unit 1 ILRT in early 2009.

The DCPP Outage 2R14 Unit 2 Containment integrated Leak Rate Test (ILRT) was performed successfully. All test acceptance criteria were met. The measured leak rate was approximately one-sixth of the acceptance criterion.

1R15 Outage Plan

DCPP presented its Outage 2R14 lessons-learned and its plans for Outage 1R15 at the DCISC June 25 & 26, 2008 Public Meeting.

Outage 1R15, scheduled for January 26 – April 16, 2009, is similar to Outage 2R14, except for the following major scope areas which differ:

The Safety Plan for 1R15 is essentially the same as that during 2R14; however, the 1R15 Safety Plan will incorporate the lessons learned form 2R14. The infrequently performed evolutions during 1R15 are the same as during 2R14 and include:

Key 1R15 challenges include:

Outage 1R15 has somewhat less scope than 2R14 due to no sump modification work being required, that work having been previously performed.

4.17.3 Conclusions and Recommendations:

Conclusion:
DCPP’s Outage 2R14, an outage of unusually large scope with four Steam Generator replacements, was performed successfully overall. As expected by DCPP, a single fuel rod defect was identified as well as an instrument guide tube with unexpected internal wear. Both are under investigation which the DCISC will continue to monitor.
Recommendations:
None

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