ML20210E173
| ML20210E173 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 02/03/1987 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Julie Ward SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| References | |
| NUDOCS 8702100243 | |
| Download: ML20210E173 (4) | |
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gs n wi b(L Docket No. 50-312 Mr. John E. Ward Sacramento Municipal Utility District Rancho Seco Nuclear Generating Station P.O. Box 15830, Mail Stop No. 291 Sacramento, California 95852-1830
Dear Mr. Ward:
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION RELATED TO THE TECHNICAL SPECIFICATION CHANGE REQUEST REGAPDING MAIN STEAM SAFETY VALVES By letter dated November 14, 1985, SMUD submitted an application (Proposed Amendment No. 127) to amend the Rancho Seco Technical Specifications (TSs) concerning Main Steam Safety Valve (MSSV) operability.
More specifically, changes were proposed for TS 3.4.2.1 which would permit full power operation of Rancho Seco with as many as three inoperable MSSVs per steam generator.
During the NRC review of this amendment request, several questions and/or concerns have evolved that can not be resolved solely from information available in the November 14, 1985 application. As such, enclosed is a request for additional information necessary for the NRC to complete its review of SMUD's proposed MSSV TS changes.
The reporting and/or recordkeeping requirements contained in this letter affect fewer than ten respondents; therefore, OMB clearance is not reouired under P.L.96-511.
Sincerely, "GHUi DES L5/t' John F. Stolz, Director PWR Project Directorate #6 Division of PWR Licensing-B
Enclosure:
Request for Additional Information cc w/ enclosure:
See next page Distribution Copies:
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Mr. John F. Ward Pancho Seco Meclear Generating Sacramento Municipal Utility District Station CC*
Pr. David S. Kaplan, Secretary Sacramento County and General Counsel Board of Supervisnrs Sacramento Municipal Utility 827 7th Street, Room 374 District Sacramento, California 95814 6201 S Street P. G. Box 15830 Ps. Helen Hubbard Sacramente, California 95813 P. O. Fox 63 Sunol, California 94ff6 Thomas A. Paxter, Esq.
Shaw, Pittner, Potts & Trowbridge 2300 N Street, N.V.
Washinoton, D.C.
20037 Mr. Ron Columbo Sacramento Municipal Utility District Rancho Secc Nuclear Generatino Station 4440 Tvin Cities Road Herald, California 9563C-9799 Mr. Robert B. Borsum Babcock & Wilcox Nuclear Power Generation Division Suite 220, 7910 Woodmont Avenue Bethesda, Maryland P0814 Resident Inspector /Pancho Seco c/o U. S. N. R. C.
14410 Twin Cities Road Herald, California 95638 Pegional Administrator, Pegion V U.S. Nuclear Regulatory Commission l
1450 Paria Lane, Suite 210 Walnut Creek, California 94596 Director i
Energy Facilities Siting Division
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Energy Resources Conservation 8 Development Commission 151C - 9th Street Sacramento, California 95814 Mr. Joseph 0. Ward, Chief Padiological Health Branch State Department of Pealth Services 714 P Street, Office Building #8 Sacramento. Califernia 95814
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ENCLOSURE 1 RANCHO SECO UNIT 1 REQUEST FOR ADDITIONAL INFORMATION RELATED TO THE PROPOSED TECHNICAL _ SPECIFICATION CHANGES REGARDING MAIN STEAM SAFETY VALVES 1.
Sacramento Municipal Utility District (SMUD) has indicated in its letter dated November 14, 1985 that the assumptions used in B&W Document 86-1153322-00, " Rancho Seco: Main Steam Safety Valve (MSSV)
Analysis," is consistent with the assumptions used in the B&W Topical Report BAW 10043, " Overpressure Protection For B&W Pressurized Water Reactors." The former report concluded that Rancho Seco could meet the requirement of overpressure protection of the secondary system when the plant is operated at 112 percent of rated power with three out of nine MSSVs per steam generator inoperable. However, the latter report concluded that the capacity margin for the main steam safety valves is only 6 percent. Provide discussion on the above discrepancy in capacity margin for the MSSVs and justify that the proposed change to the technical specification will provide reasonable assurance of adequate overpressure protection of the secondary system.
2.
SMUD presented, in B&W Document 86-1153322-00, the results of a bench-mark analysis which compared the results of an analysis using RELAP 5 with the Davis Besse turbine trip transient data obtained in November, 1982..However, the RELAP 5 model does not account for any turbine bypass relief capability wh.ile the Davis Besse transient includes the actuation of the turbine bypass and the atmospheric dump valves. The staff does not consider the above stated benchmark analysis valid.
Provide justification, including supporting code verification and appropriate sensitivity analysis demonstrating the conservatism of the methodology used, for the use of the RELAP 5 code for calculating secondary system overpressure transient.
3.
SMUD stated in a letter dated November 14, 1985 that the proposed changes of the technical specification regarding the MSSV availability do not involve t
l a significant reduction in a margin of safety. Justify this determination in light of the reduction in pressure relieving capacity margin of the l
secondary system due to the proposed technical specification change to Section 3.4.2.1, which will permit a 33% reduction in secondary system pressure relieving capacity during full power of operation at Rancho Seco.
4 In the IE Information Notice No. 86-05 regarding MSSV test failures and ring setting adjustments, NRC has expressed its concern on the MSSV problems.
Describe the testing and maintenance program used at Rancho Seco to pre-clude similar problems from occurring and thereby provide assurance of proper MSSV operation.
. ~.
I' 5.
In light of the industrial experiences on the use of MSSV identified in the IE Information Notice No. 86-05, provide justification on how the proposed technical specification will provide secondary system over-pressure protection with reliable facilities and high confidence level.
6.
Section 1.1 of the B&W Document 86-1153322-00, " Rancho Seco: Main Steam Safety Valve (MSSV) Analysis," states that primary system pressure is not of major concern in this analysis as events which require MSSV actuation results in primary side pressurization to the Reactor Protection System (RPS) high pressure setpoint and consequent reactor trip within the first ten seconds of the event. For design basis event, a turbine trip from overpressure, an anticipatory reactor trip prevents any significant FCS pressure increase. This statement is not consistent with the assumptions listed in Section 3.1 of the report which states that no direct trip is assumed to result from turbine trip. Also, the report does not provide sufficient discussions to address the effect of the primary system pressure transient due to a reduction of the secondary system pressure relieving capacity. Provide the results of such analysis.
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