ML25196A164
| ML25196A164 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 01/14/1977 |
| From: | Bender M Advisory Committee on Reactor Safeguards |
| To: | Rowden M NRC/Chairman |
| References | |
| Download: ML25196A164 (1) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, O. C. 20555 Honorable Marcus A. Rowden Chairman.
- u. s. Nuclear Regulatory Conmission Washington, DC 20555 January 14, 1977 SUBJEX:T:
REFORr 00 DOOALD C. 000K NUCLEAR PLANT UNIT NO. 1
Dear Mr. Rowden:
During its 201st meeting, January 6-8, 1977, the Advisory Conmittee on Reactor Safeguards CODPleted its review of the proposal to replace, dur-ing the first refueling of the Donald C. Cook Nuclear Plant Unit No. 1, 65 of the original Westinghouse Electric Corporation fuel assemblies with Exxon Nuclear Company (ENC) fuel assemblies and to operate the re-sulting core to produce rated reactor power of 3250 MWt.
The Comnittee has previously discussed this plant in its reports of Decenber 13, 1968, October 17, 1973, and March 11, 1976. A Subcomnittee meeting to consider the current proposal was held in Washington, D. C., on December 22, 1976.
During its review, the Comnittee had the benefit of discussions with representatives of Indiana and Michigan Power Conpany, American Electric Power Service Corporation, ENC, and the Nuclear Regulatory COilllli.ssion (NRC) Staff. The CODmittee also had the benefit of the docmnents listed.
The NRC Staff has concluded that the design of the ENC fuel assemblies propose(I for Donald C. Cook Nuclear Plant Unit No. 1 Cycle 2 is similar to that supplied by ENC for other pressurized water reactors (PWRs). The NRC Staff has indicated that its review of ENC fuel design analytical methods is not yet conplete but that the review has progressed sufficiently to indi-cate that the methods are adequate for application to Donald c. Cook Nuclear Plant Unit No. 1 Core 2. Approximately 1000 fuel bundles manufactured by ENC are in PWRs and in boiling water reactors with. burnups ranging fran first cycle to 25,000 megawatt-days per metric ton of uranium. Performance of these assemblies has been good.
Primarily because of the low back pressure produced by the ice-condenser type containment following a loss-of-coolant accident, the peaking factor required to satisfy the emergency core cooling system (ECCS) Acceptance Criteria of 10 CFR 50.46 is unusually low. The ENC analysis satisfied the ECCS Acceptance Criteria of 10 CFR 50.46 with an assumed peaking fac-tor of 1.95 at rated power. The Licensee proposes a peaking factor 297
Honorable Marcus A. Rowden 2 -
January 14, 1977 Technical Specification limit of 1.95 at rated power for Cycle 2. The Licensee proposes continued use of the axial power distribution ironi-toring system (AP'S) for determining conformance. Experience with APIM> during Cycle 1 operation at Donald C. Cook Nuclear Plant Unit No. 1 and in other reactors indicates this system can provide an appro-priate measurement of the core power distribution.
Although sufficient information and analyses exist to predict the per-formance of the Westinghouse fuel at the beginning of Cycle 2, further analyses may be appropriate with regard to both fuel pellet-clad inter-action and fission gas release rate, with operation near the end of the cycle. The Committee wishes to be kept informed.
During Cycle 1 operation, one or two fingers broke off a control rod dur-ing rod drop timing tests. The Licensee and Westinghouse Electric Corpo-ration have concluded that the observed failure is not indicative of generic failures and will not adversely affect reactor control rod scram times. The NRC Staff is requiring further examination and analyses by the Licensee. The ACRS wishes to be kept informed.
The ACRS believes that, subject to the foregoing and to matters discussed in its report of Marcll 11, 1976, the Donald C. Cook Nuclear Plant unit No. 1 can be operated with the proposed reload core up to the design power of 3250 MWt, under the proposed operating and ironitoring conditions, without undue risk to the health and safety of the public.
Sincerely yours, M. Bender Chairman Mditional Comments by Members David Okrent and Milton Plesset Jn connection with the Marcll 11, 1976 report on Donald C. Cook Nuclear Plant Unit No. 1, we made additional comments whid1 included the following:
"First, while there may be merit in the proposed cllanges in the Westing-house evaluation trodel, we believe further examination is warranted of several factors, including the scaling of experiments, the scatter in data, and the possible influence of super-plasticity on clad behavior during postulated loss-of-coolant accidents. Our reluctance to endorse these changes is also due, in large part, to signs of a continued pro-cess of cutting into the conservatisms built into the original evalua-tion models, without a concomitant build-up in our basic understanding or predictive ability for the overall LOCA-ECCS process. In this situation there are limits beyond whicll the use of best estimate heat transfer coefficients, etc., is no longer appropriate.
298
Honorable Marcus A. Rowden 3 -
January 14, 1977 "Second, even with application of the revised Westinghouse evaluation model which has been judged acceptable by the NRC Staff, Donald C. Cook Nuclear Plant Unit No. 1 requires a LOCA - limited maximum peaking factor (FQ} of 1.98 (plus the margin for bowing} at rated power. While this is somewhat higher than the FQ which can be expected at steady operation for the rest of the first fuel cycle for Donald C. Cook Nuclear Plant Unit No. 1, it still represents a very large reduction in the margin that has been available for most plants between LOCA - limited FQ and that value which would be present most of the time. This margin has been eroded until it is a small fraction of its earlier values. Further-more, if we accept this low FQ value for Donald C. Cook Nuclear Plant Unit No. l, a precedent will oe set by means of which all PWR's will be able to reduce what was a substantial safety margin only a few years ago. This previously available substantial safety margin could cover many of the existing uncertainties in the analysis of LOCA-ECCS. The uncertainty aspect is highlighted by the less than perfect record obtained by the experts in their pre-prediction of various separate effects experi-ments, by the recognized difficulties in a calculation from first principles, by the current unavailability of experiments to test all relevant effects, and by the lack of a meaningful test of Westinghouse predictive capability with experiment.
"Third, the ACRS has in the past been reluctant to accept proposed operation of reactors with Fa's less than 2.2. In part, such caution arose from the knowledge that, wit:h a more flattened power distribution, a much larger fraction of the fuel elements would be at or near peak temperatures, given a LOCA, and therefore potentially vulnerable to an "anomaly" in ECCS function (such as some three-dimensional flow effect or excessive steam generator leakage)."
We find that these comments apply equally to the proposed operation with Exxon Nuclear Company (ENC) fuel.
We believe that the proposed new ENC ECCS evaluation model is subject to considerable uncertainty, particularly with regard to flow blockage effects, the choice of FLECHT heat transfer coefficents, and steam cooling.
More importantly, as we suggested on March 11, 1976, the NRC Staff has continued to follow a legalistic approach in its interpretation of 10 CFR 50, Appendix K, accepting so-called best-estimate parameters and models in areas where conservatism is not explicitly required.
Since March 1976, a significant nmnber of operating ~ms have been granted authority to operate with peaking factors even less than 1.98; for example., Surry Units I and 2 were granted approved peaking factors of 1.80 and 1.82, respectively, on August 27, 1976.
299
Honorable Marcus January 14, 1977 In view of the current state of knowledge, we do not believe that the path currently being followed by the NRC Staff is prudent, and we recormnend that the Nuclear Regulatory Commission reexamine 10 CFR 50, Appendix K, including its actual implementation in evaluation models.
For Donald C. Cook Nuclear Plant Unit No. 1, we still believe that opera-tion with the present design of fuel assembies and ECCS, should be limited to about 92% of rated power.
References:
- 1. Revision 1 to Nuclear Reactor Regulation (NRR) Safety Evaluation Report on the Exxon Nuclear Company (ENC) WREM-Based Generic PWR:-ECCS Evalua-tion Model Update ENC-WREM-II, dated January 5, 1977
- 2. Letter, Indiana and Michigan Power Company (I and M) to NRR, dated Decei-nber 17, 1976, concerning reactor vessel overpressur ization events
- 4. Letter, I and M to NRR, dated December 13, 1976, concerning proposed changes to Technical Specifications on power distribution limits and surveillance requirements
- 6. Letter, I and M to NRR, dated November 23, 1976, concerning modifi-cations being made to valve control circuits and procedures
- 7. Letter, I and M to NRR, dated Nove.1tlber 23, 1976, forwarding responses to NRR questions concerning a permit to operate at full power during Cycle 2
- 8. Letter, ENC to NRR, dated November 19, 1976, forwarding XN-76-35 Supplement 1, "Assumptions Used in the Plant Transient Analysis for the Donald C. Cook Unit 1 Nuclear Plant"
- 9. Letter, I and M to NRR, dated November 17, 1976, forwarding XN-76-35, "Donald C. Cook Unit 1 u:x:A Analyses Using the ENC WREM-Based PWR ECCS Evaluation Model (ENC-WREM-II)"
- 10. Letter, I and M to NRR, dated November 17, 1976, forwarding the results of analyses of the effect of degraded grid voltage on the operability of safety-related eguipnent
- 11. Letter, I and M to NRR, dated November 11, 1976, concerning the loose-parts monitoring system
- 12. Letter, I and M to NRR, dated November 5, 1976, forwarding answers to NRR questions on the reload license application
- 13. Letter, American Electric Power Service Corporation to the Office of Inspection and Enforcement, dated October 29, 1976, forwarding a supplement to the Startup Test Report
- 14. Letter, I and M to NRR, dated October 27, 1976, concerning fire protection considerations 300
Honorable Marcus A. Rowden 5 -
January 14, 1977 References Cont'd
- 15. Letter, I and M to NRR, dated October 19, 1976, concerning sus-ceptibility to reactor vessel overpressurization events
- 16. Letter, I and M to NRR, dated October 1, 1976, forwarding answers to NRR questions on ENC reports XN-76-25 and XN-75-39
- 17. Letter, I and M to NRR, dated October 1, 1976, forwarding the
- report, 11Long Term Evaluation of the Ice Condenser System Results of the July 1976 and Septe.'llber 1976 Ice Weighing Programs"
- 18. Letter, I and M to NRR, dated August 27, 1976, concerning the evalua-tion of the adequacy of the reactor pressure vessel supports
- 19. Letter, I and M to NRR, dated August 26, 1976, forwarding XN-76-36, "Exxon Nuclear Company WREM-Based Generic PWR ECCS Evaluation Model (ENC-WREM-II)"
- 20. Letter, ENC to NRR, dated August 20, 1976, forwarding XN-76-35, "Plant Transient Analysis for the Donald C. Cook Unit 1 Nuclear Power Plant"
- 21. Letter, I and M to NRR, dated July 30, 1976, forwarding the report, "Long Term Evaluation of the Ice Condenser System - Results of the January 1976 and April 1976 Ice Weighing Programs"
- 22. Letter, I and M to NRR, dated July 20, 1976, concerning request to operate at full power during Cycle 2