ML20008F331
| ML20008F331 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 02/27/1981 |
| From: | Reid R Office of Nuclear Reactor Regulation |
| To: | Mattimoe J SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| References | |
| NUDOCS 8103120838 | |
| Download: ML20008F331 (3) | |
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o UNITED STATES 8'
7,,o NUCLEAR REGULATORY COMMISSION
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g WASH 6NGTON, D. C. 20555 k *...+ /
February 27, 1981 Docket No. 50-312
'e Mr. J. J. Mattimoe
,4 Assistant General Manager and 6
Chief Engineer
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Sacramento Municipal Utility District 6201 S Street
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Dear Mr. Mattimoe:
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n We have reviewed your report transmitted by letter dated January 14, which re.enonded to our request for infonnation dated August 21, 1979, concerning a benchmark analysis of sequential auxiliary feedwater flow to the steam generators following loss of main feedwater using a 3 node CRAFT 2 OTSG representation.
We have determined that a more rigorous assessment of the Babcock &
Wilcox (B&W) small break LOCA model is being performed under Item II.K.3.30 "Small Break LOCA Methods" of the TMI Action Plan requirements and, therefore, further code assessment under Item II.K.2.19 " Benchmark Analysis of Sequential f,uxiliary Feedwater Flow" of tr.2 TMI Action Plan requirements is unnecessary.
We have concluded that you have complied with the intent of Item II.K.2.19 and, therefore, Item II.K.2.19 is resolved for Rancho Seco Nuclear Generating Station.
Our associated Safety Evaluation Report is enclosed.
Sincerely,
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rbW.Reid, Chief Operating Reactors Branch #4 Division of Licensing
Enclosure:
Safety-Evaluation cc w/ enclosure:
See next page B103120
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Sacramento Municipal Utility '
Rancho Seco, Docket No. 50-312 District cc w/ enclosure (s):
David S. Kaplan, Secretary and Christopher Ellison, Eso.
General Counsel Dian Grueuich, Esq.
6201 S Street California Energy Comission P. O. Box 15830 1111 Howe Avenue Sacramento, California 95813 Sacramento, California 95825 Sacramento County Ms. Eleanor Schwartz Boar.d of Supervisors California State Office 827 7th Street, Room 424 600 Pennsyl"ania Avenue, S.E., P,m. 201 Sacramento, California 95814 Washington, D. C.
20003 Business and Municipal Department Docketing and Service Section Sacramento City-County Library Office of the Secretary 828 I Street U.S. Nuclear Regulatory Comission Sacramento, California 95814 Washington, D. C.
20555 Resident.Inspcctor/ Rancho Seco Director, Criteria and Standards c/o U. S. N. R. C.
Division 14410 Twin Cities Road Office of Radiation Prograns (ANR-460)
Herald, CA 95638 U. S. Environmental Protection Agency Washington, D. C.
20460 Dr. Richard F. Cole Atomic Safety & Licensing Board Panel U.S. Nuclear Regulatory Comission U. S. Environmental Protection Agency Washington, D. C.
20555 Region IX Office ATTN: EIS C00R]INATOR Mr. Frederick J. Shon 215 Fremont Street Atomic Safety and Licensing Board San Francisco, California 94111 Panel U.S. Nuclear Regulatory Comission Mr. Robert B. Borsum Washington, D. C.
20555 Babcock & Wilcox Nuclear Power Generation Division Elizabeth S. Bowers, Esq.
Suite 420, 7735 Old Georgetown Road Chairman, Atomic Safety end Bethesda Maryland 20014 Licensing Board Panel U.S. Nuclear Regulatory Comission Thomas taxter, Esq.
Washington, D. C.
20555 Shaw, Pittman, Potts & Trowbridge 1800 M Street, N.W.
Mr. Michael R. Eaton Washington, D. C.
20036 Energy Issues Coordinator Sierra Club Legislative Office Herbert H. Brown, Esq.
1107 9th Street, Room 1020 Lawrence Coe Lanpher, Esq.
Sacramento, California 95814 Hill, Christophce and Phillips, P.C.
1900 M Street, N.W.
Atomic Safety and Licensing Board Washington, D. C.
2rJ36 Panel U.S. Nuclear Regulatory Comission Helen Hubbard Washington, D. C.
20555 P. O. Box 63 Sunol, California 94586
SaGramento Municipal Utility -
District
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Atomic Safety and Licensing Appea' Board Panel
).S. Nuclear Regulatory Commission llashington, D. C.
20555 California Department -
'2alth ATTN: Chief, Enviro..;r.tal Radiation Control Unit Radiological Health Section 714 P Street, Roou 498 Sacramento, California 95814 4
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NUCLEAR REGULATORY COMMISSION g:4 '.n g
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION CONCERNING ITEM II.K.2.19 " BENCHMARK ANALYSIS OF SEOUEi'TIAL AUXILIARY FEEDWATER FL0r FOR BABC0CK & WILCOX REACTOR PLANTS DOCKETS NOS. 50-269, 50-270, 50-267, 50-289, 50-302,50-312, 50-313 AND 50-346 Introduction At a meetine in Bethesda, April 25, 1979, with the oners cf Babcock and Wilcox (S&W) reactor plants, we requested a bench. art analysis of sequen-tial auxiliary feedwater flow to the steam generators icilowing a less of main feedaater. This analysis was provided in a let er frem J. Taylor (B&W) to R. Mattson (HRC) dated June 15, 1979. However, in this analysis the TRAP-2 Code with 6 node steam generator model was utilized. All small break analysis presented to the NRC have been performed using the CRAFT-2 Code with a 3 node steam generator model. We require a senchmark analysis for sequential auxiliary feedsater flow also be performed using CFAFT-2 with a 3 mode steam generator representation. By letter dated August 21 ~,
1979 we requested such analysis.
Each licensee of E6U reactor plants
, responded with a report which presented analysis of sequential auxiliary feedwater flow to the steam cenerators for a less of main feedwater trans-ients using the CFAFT-2 Code!
This issue was later identified as Item II.K.2.19 of the TMI Action Plan reqLirements.
Discussion & Conclusions S&W utilizes the CRAFT-2 computer progran in perfc -ing less of coolant accident (LOCA) licensing evaluations for their nuclear steam supply systems (NSSS).
Subsequent to the Tlil-2 accident, this ccmputer pro; ram was used to confirm emergency operator guidelines for all power phnts with USSSs designed by S&W. Our review of these confirmatory antlyses have ied to questions re-garding the ability of the CRAFT-2 progr. to adequately predict steam generator performance anc its influence on the primsry system thermal-hydraulic behavicr.
In particular, we noted that the CFAFT-2 steam generator nodel did not contain the same degree of detail as the model used with the TRAF-2 Code.
TPAP-2 is a ccmputer code primarily used for non-LOCA transients by D&W.
In order to validate the TFAF-2 transient code cith actual plant data, an asyrostric cocidown test was. incorporated into the Crystal River Unit 3 poaer ascension progran.
Eecause cf the sinplified steam generator model in the CRAFT-2 Code, we also re;uested that the CFAFT-2
': S be assessed against the Crystai River Unit 3 asy :vric coolder:n detr..
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The comparative analyses of the startup test demonstrated that the simplified steam generator model used in the licensing code (CRAFT-2) predicted thennal-hydraulic behavior similar to the more detailed steam generator model utilized in the TRAP-2 Code. However, comparisons with data for both codes were poor. Further examination of the Crystal River Unit 3 asymnetric startup test has indicated the test to be inappropriate for assessing com-puter codes. This is attributed to inadequate instrumentation whercby key data required for code assessment were not obtained.
Reviews conducted by our B&O Task Force, following the TMI-2 accident, have concluded that further assessment of the CRAFT-2 Code would be required.
The najority of the concerns identified are documented in NUREG-0565.
In particular, the neglect of a mechanistic, regine-dependent heat transfer model and the use of a constant, steam generator heat transfer coefficient throughout the transient have been identified as requiring either revision or further justification.
This requiremer.t for further justification and/or revision of the small break ECCS nedels is being performed under TMI Action Plan Item II.K.3.30.
We believe that satisfactory resolution of code modeling ccncerns as part of the Action Iten II.K.3.30 will resolve the ecdeling concerns of II.K.2.19.
The conclusions of our review of Action Item II.K.2.19 are as follows:
(a) The intent of Item II.K.2.19 was accomplished, (b)
Results provided by CRAFT-2 were similar to those provided by the more detailed TRAP-2 program. However, both codes showed poor agreement when compared with the test data, (c) The poor agreement of the code prediction with test data has been attributed to the fact that the Crystal River ascension test data was not adequate for assessing thernal-hydraulic codes, and (d) A more rigorous assessment of the B&W small break LOCA model is being performed under TMI Action item II.K.3.30.
Further code essessment under TMI Action Item II.K.2.19 is therefore unnecessary.
Based on the above conclus' ions, we consider Item I'.K.2.19 completed by all rf censees with B&W NSSSs by issuance of this Safet) Evaluation Report.
l Moreover, we do not believe it necessary for Item II.K.2.19 to be addressed any further.
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Dated:
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