05000275/FIN-2008005-01
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Finding | |
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| Title | 500 Kv Offsite power Source Compliance With General Design Criterion 17 |
| Description | The 500 kV system provided the plant second Technical Specification required offsite power source. The 500 kV system is a delayed off-site source requiring the main generator disconnects to be removed before power can be back fed through the station service transformers to the safety related buses. During the initial NRC review of the system, PG&E stated that the 500 kV power source can be available by manual initiation in about 30 seconds. The NRC concluded (Safety Evaluation by The Directorate of Licensing U.S. Atomic Energy Commission in the Matter of Pacific Gas and Electric Company Diablo Canyon Nuclear Power Station, Units 1 And 2 San Luis Obispo County, California Docket Nos. 50-275 And 50-323) that the 500 kV off-site power source was acceptable because the circuits provided sufficient assurance that redundant and independent sources of offsite power are provided, as required by General Design Criterion 17. General Design Criterion 17 stated that the delayed power source must be available in sufficient time to assure that design conditions of the reactor coolant pressure boundary are not exceeded. In 1998, PG&E modified the Updated Final Safety Analysis Report to specify a 30-minute delay time before the 500 kV power source could be aligned to the safety related buses. On July 31, 2008, the inspectors observed that licensed operators took about 40-minutes to complete the 500 kV back feed during plant simulator requalification training (Course R08, Lesson R082S2, Loss of All Alternating Power and a Seismic Event). A loss of reactor coolant pump seal injection and cooling is anticipated during the delay time. Based on information provided by NRC Information Notice 2005-14, Fire Protection Findings on Loss of Seal Cooling to Westinghouse Reactor Coolant Pumps, and Westinghouse Technical Bulletin, TB-04-22, Reactor Coolant Pump Seal Performance, Appendix R Compliance and Loss of All Seal Cooling, Revision 1, the inspectors estimated that about 12 gallons per minute of reactor coolant inventory would be lost through the reactor coolant pump seals during the first 8 minutes of the delay time. After 8 minutes reactor coolant leakage would increase to about 88 gallons per minute until seal injection could be reestablished. Based on the simulator response, the inspectors estimated that about 2,800 gallons of reactor coolant inventory would be lost through the reactor coolant pump seals, resulting in pressurizer level dropping below the indicating range during the delay time. The inspectors also anticipated a low pressurizer pressure safety injection accident signal to occur within the first 20 minutes due to the combination of the loss of reactor coolant and the inability to throttle turbine drive auxiliary feedwater flow to the steam generators. On August 26, 2008, the inspectors requested the licensee provide the design measures demonstrating that the 500 kV power source met the General Design Criterion 17 design basis. On October 28, 2008 the licensee stated that the requested design basis was not retrievable. Plant engineers reevaluated the 500 kV off-site power system and concluded that General Design Criterion 17 compliance was demonstrated by a road map of pre-existing analysis created to support other plant design basis. The road map included the assumption that no excessive loss of reactor coolant would occur due to reactor coolant pump seal leakage during the delay time. The inspectors determined that the licensees assumption that the reactor coolant pump seals would remain intact during the delay time was incorrect. The licensee reevaluated the General Design Criterion 17 analyses with assumed reactor coolant pump leakage provided by the vendor. Pacific Gas and Electric again concluded that General Design Criterion 17 acceptance criteria were met because the reactor coolant system pressure and temperature would be maintained less than 110-percent of the of the design values during the delay time. The inspectors were unable to verify that the NRC had used 110-percent of the reactor coolant system design values as acceptance criteria for General Design Criterion 17 in the past. This issue is unresolved pending NRC review of the General Design Criterion 17 acceptance criteria applied by PG&E and basis and verification of 30-minutes assumed for the delay time. Unresolved Item 05000275/2008005-01, 05000323/2008005-01, 500 kilo-Volt Off-Site Power Source Compliance with General Design Criterion 17 |
| Site: | Diablo Canyon  |
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| Report | IR 05000275/2008005 Section 1R04 |
| Date counted | Dec 31, 2008 (2008Q4) |
| Type: | URI: |
| cornerstone | Mitigating Systems |
| Identified by: | NRC identified |
| Inspection Procedure: | IP 71111.04 |
| Inspectors (proximate) | L Ricketson G Guerra M Brown M Peck P Elkmann V Gaddy M Runyan R Kellar A Fairbanksg Guerram Brown M Peck V Gaddy |
| INPO aspect | |
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Finding - Diablo Canyon - IR 05000275/2008005 | |||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||||
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Finding List (Diablo Canyon) @ 2008Q4
Self-Identified List (Diablo Canyon)
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