ML20151G652
| ML20151G652 | |
| Person / Time | |
|---|---|
| Issue date: | 07/20/1988 |
| From: | Kerr W Advisory Committee on Reactor Safeguards |
| To: | Stello V NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| References | |
| ACRS-R-1311, NUDOCS 8807290062 | |
| Download: ML20151G652 (3) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION ppg n
.l ADVISORY COMMITTEE ON REACTOR SAFEGUARDS g
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WASHINGTON, D. C. 20555 4 +...+ $"
July 20, 1988 Mr. Victor Stello, Jr.
Executive Director for Operations U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Cear Mr. Stello:
SUBJECT:
COMMENTS ON THE NPC STAFF'S DRAFT SAFETY EVALUATION OF THE WESTINGHOUSE TOPICAL REFORT, WCAP-10924. "WESTINGHOUSE LARGE-BREAK LOCA BEST-ESTIMATE METHODOLOGY" During the 339th meeting of the Advisory Comittce on Reactor Safe-guards, July 14-16, 1988, we met with members of the NRC Steff and their consultants as well as representatives of the Westinghouse Electric Corporation ard licensees of the Prairie Island and Point Beach nuclear pcwer plants and reviewed the subject safety evaluation.
Cur Subcom-mittee on Thermel Hydraulic Phenomena mct on May 27 and June 21, 1988 to discuss this matter. We also bad the benefit of the dccuments listed as references for this letter.
This review concerns an improved method for predicting the performance of the er,ergency core cooling system (ECCS) during a large-break loss-of-coolant accident (LOCA) in two-loop pressurized water reactor (FWR) plants of Westinghcuse design.
The ECCS in these plants is different from those in the mejority of PWR plants in that low-pressure injection flow is provided to the reactor vessel through special nozzles entering the upper plenum rather than through the cold leg piping.
Several years ago, the NRC Staff recomerded that certain thermal-hydraulic conditions unique to the upper plenum injection (UPI) config-uretion were not adequately modeled in the ECLS codes being used.
Accordingly, licensees at UPI plants were instructed to develop improved models, to demonstrate that these models provided compliance with the ECCS rule, and to use these models in future licensing submittals.
Westinghouse has developed such an improved model for the utilities operating the Prairie Island and Point Beach nuclear power plants.
That is the subject of the present review.
Madels for two other UPI plants are being developed elsewhere and are to be given a separate review.
The improved analysis is provided by a code called WCOBRA/ TRAC that has been developed from existing codes TRAC-FD2 and COBRA-TF.
COBRA-TF is
' 't the reactor vessel part of the analysis while TRAC-PD2 models the gp/
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4 Mr. Victor Stello, Jr. July 20, 1988 s
overall system.
COBRA /TF describes three flow fields in three dimen-sions.
It thus has the capability of modeling thermal-hydraulic phenom-ena important in the UPI configuration, including details of counter-current flows of vapor and liquid.
In addition to the improved phenomenological modeling, the overall methodology presented in WCAP-10924 incorporates a so-called "best-estimate" approach to calculation of limiting plant parameters.
The general approach described in the referenced NRC document SECY-83-472, is used.
The staff's review has concluded that the WCOBRA/ TRAC code provides adequate tredeling to represent the UPI plant configuration for large-break LOCA analysis.
They have also concluded that the best-estimate methodology, including allowances for uncertainty, adequately conform to the provisions outlined in SECY-83-472. However, they have not accepted the best-estimate methodology, as presented, for other plants or for use with the proposed new ECCS rule when that becomes available. We find no reason to disagree with these conclusions of the staff.
A cautionary word about the so-called "best-estimate" approach: we have previously expressed our approval of, and, in fact, applauded, thi3 approach to analysis of reactor transients.
This applies to both the SECY-83-472 approach and the proposed new rule for large-break LOCA analysis.
Best-estimate analysis, in this sense, has two parts:
(1)a realistic analysis, with no purposeful biases, to provide a central estimate of the parameters cf interest and (2) a conscious and explicit estimate of the nargin that should be provided from this central esti-mate to achieve a desired level of confidence in conclusions to be drawn from the analysis.
In most practical engineering sitrations, including LOCA analysis, the relationships are so complex and the data so sparse that mathematical I
rigor in defining the desired confidence level and necessary allowance is impossible. However, the method is still of value even though it may involve what are largely engineering judgments about confidence level and the magnitude of allowances.
The problem is that, too often, practitioners of the best-estimate analysis or users of the results describe their analysis and the results in terms that imply mathematical rigor and give an impression that statistical relationships have been i
developed with great precision, whereas, the actual data and methods of analysis are approximate.
Terms such as "95% confidence interval" are used when only a term such as "a reasonably high confidence level" is justified.
In addition, distinctions among variability, uncertainty, and confidence level are not observed, and statistical relationships are often used carelessly and inaccurately.
We recomend that, in the future, the NRC Staff should involve professional statisticians in the review of these ratters.
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3 July 20, 1988 Mr. Victor.Stello, Jr.
Also, there will be a greater technical challenge for the staff in reviewing best-estimate analyses compared with evaluation-model reviews carried out in the past. We believe that agency management will have to make a special effort to provide appropriate resources.
r We hope that our comnents will prove useful.
Sincerely, l
W. Kerr Chairman l
1
References:
1.
Memorandum dated July 12, 1988, from M. W. Hodges, NRC, to P.
- Boehnert, NRC, transmitting draft "Safety Evaluation of the Westinghouse Electric Corporation Topical Report, WCAP-10924,
' Westinghouse Large-Break LOCA Best-Estimate Methodology'"
(Proprietary) 2.
U.S. Nuclear Regulatory Comission Staff Document, "Emergency Core
)
Cooling System Analysis Methods," SECY-83-472, dated Novembe, 17, 1983 3.
Westinghouse Electric Corporation. WCAP-10924-P:
"Westinghouse Large Break LOCA Best Estimate Methodology - Volume 1:
Model Description and Validation," June 1986 (Proprietary); and WCAP-10924-P, Volume 2 Revision 1:
"Application To Two-Loop PWRs Equipped With Upper Plenum Injection," April 1988 (Proprietary) 4.
Westinghouse Electric Corporation:
"Responses to NRC Questions on Westinghouse large Break LOCA Best Estimate Methodology, WCAP-l 10924-P Voluxe 1," Octot'er 1987 (Proprietary) 5.
Westinghouse Electric Corporation: WCAP-10924-P, Volume 2, Adden-dum 1.
Responses to NRC Questions on WCAP-10924-P. Volume 2 l
l (Addendum to Westin house largt Break LOCA Best-Estimate Method-ology, Vclume 2:
A lication to Two-Lnop PWRs Equipped With Upper f
PlenumInjection)," pril 1988 (Proprietary)
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