ML20040H214
| ML20040H214 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 04/19/1978 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Eicheldinger C WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| Shared Package | |
| ML18023A006 | List: |
| References | |
| NUDOCS 8202170449 | |
| Download: ML20040H214 (3) | |
Text
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.s-APR IS 1978 Mr. C. Eicheldinger, Manager 4
Nuclear, Safety.Departnent Westinghouse Electric Corooration P. D. Box 355 Pittsburgh, Pennsylvania 15230
Dear Mr. Eicheldinger:
I
SUBJECT:
STAFF EVALUATION OF WCAP-7956, WCAP-8054, WCAP-8567, AND WCAP-8762 The Nuclear Regulatory Commission staff has completed its review.of the following Westinghouse Electric Corporation topical reports.
WCAP-7956 (flon-proprietary), "THINC-IV, An Imoroved Program for Themal-Hydraulic Analysis of Rod Bundle Cores" June 1973 i.
WCAP-8054 (Proprietary) and WCAP-8195 (Non-proprietary),
" Application of the THINC-IV Program to PWR Design,"
September 1973 WCAP-3567(Proprietary)andWCAP-8558(Non-proprietary),"I= proved Thermal Design Procedure," July 1975 WCAP-3762 (Proprietary) and WCAP-8763 (Non-proprietary), "New Westinghcuse Correlation WRS-1 for Predicting Critical Heat Flux in Rod Bundles with Mixing Vane Grids," July 1976.
Our evaluations of these reports are enclosed.
As a result of our review of WCAP-7956 and WCAP-8054, we have concluded that the THINC-lV computer code is acceptable for perfoming steady-state hydraulic calculations in reactor cores provided suitably conservative i
assumptions are used with respect to plant operating conditions, fuel fabrication tolerances, and power peaking uncertanties.
Fluid co. ditiens i
are limited to the single phase or the hemogeneous two phase flow regime. Limitations on the use of the THINC-IV cede are provided and I
fully discussed in our evaluation of these topical reports (Enclosure 1).
f WCAP-8567 describes a new procedure for calculating departure-from-nucleate-boiling ratio (DNSR) in reactor cores based on the statistical combination of uncertainties in plant parameters that affect DNBR.
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As a result of our review, of WCAP-3567, we have concluded that the new thermal design procidere is, acceptable for use in licensing applications subject to the imposition of certain restrictions and changes to the procedure. The sensitivity factors for plant parameters are valid only for the W-3 DNB correlation. THINC-IY computer code, and the range of pla,nt parameters considered in the report. Sensitivity factors, variances, and distributions for plant parameters ~ must be included and justified in each plant safety analysis report. Code urcertaint'ies specified by the staff (Tn DN3R analyses.
+4% for THINC-IV and +1% for.
transient analyses) must be included Fuel red bowing-effects 'cannot be treated a described in the report; however, addition of aiod-bew penalty to the DNBR limit is acceptable. A plant transient resulting from a loss of reactor coolant flow must satisfy the design -
basis DNBR limit for faults of moderate frequency (ANS Condition II event) rather than the design basis limit for infrequent accidents (ANS Condition III event) as specified in WCAP-8567. The improved themal design procedure is applicable only to DNBR analyses and cannot be used in other analyses, such as overpressure calculations.
The acceptable conditions for use of the improved thermal design
. procedure are fully described in our safety evaluation of WCAP-8567 (Enclosure 2).
As a result of our review of WCAP-3762, we have concluded that the WP3-1 critical heat flux correlation is acceptable for use in thermal-hydraulic calculations of pressurized water reactor cores provided a DNBR limit of 1.37 is used for cores composed of fuel assemblies with an "L" grid l
design. The proposed DNSR limit of 1.17 is acceptable for cores composed l
of fuel assemblies with an "R" grid design. The limit of 1,37 for "L" grid designs results from a large variation in data 7'e the limited number of tests run with this design. This limit may be reduced by obtaining additional test data or by additional analytical work to improve the correlation. Our evaluation, including an independent audit of calculations to determine local coolant. conditions for the tests and an independent audit of the statistical calculations used to establish the DNBR limit.is sumarized in Enclosure 3.
Accordingly, topical reports WCAP-7956, WCAP-3054,.WCAP-8567, and WCAP-8762 are acceptable for reference in license amplications. Topical reports WCAP-8195, WCAP-8568, and WCAP-6763 are acceptable non-prepiietary versions of the proprietary reports. When these reports are used as references, both the prcprietary report and the non-proprietary version must be referenced.
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'. Mr. C. Eicheldinge. APR 191978 In accordance with established procedure, it is requested that Westinghouse issue revised versions of these reports within three months of receipt of this letter to include the NRC acceptance letter, the enclosed evaluations, and any changes resulting from the review.
We do not intend to repeat our review of these reports witen they appear as references in,a particular licerse application except to assure that the material presented in these reports is applicable to the specific plant involved.
Should Nuclear Regulatory Connission criteria or regulations change, such that our conclusions concerning these reports are invalidated, you will be notified and given an opportunity to revise and resubmit your topical reports, a
Sincerely, John F. Stol:, Chief Light Water Reactors Branch No.1 Division of Project Management i
Enclosures:
1.
Topical Report Evaluation WCAP-7956 and WCAP-8054 2.
Topical P.eport Evaluation i
WCAP-8567 3.
Safety Evaluation Report WCAP-8762 cc:
D. Aawlins l
Westinghouse Electric Corporation P. O. Box 355 Pittsburgh, Pennsylvania 15230 i
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