ML25197A051
| ML25197A051 | |
| Person / Time | |
|---|---|
| Issue date: | 07/15/1975 |
| From: | Kerr W Advisory Committee on Reactor Safeguards |
| To: | Anders W NRC/Chairman |
| References | |
| Download: ML25197A051 (1) | |
Text
ADVISORY COMMITTEE ON REACTOR SAFEGUARDS NUCLEAR REGULATORY COMMISSION Honorable Wiiliam A. Anders Chairman WASHINGTON, D. C. 20555 U.S. Nuclear Regulatory Commission Washirgton, D. C.
20555
Subject:
REPORT ON FAST FLUX TFST FACILITY
Dear Mr. Anders:
July 15~ 1975 Durirg its 183rd Meetirg, July 10-12, 1975, the Advisory Committee on Reactor Safeguards continued its review of the Energy Research and Development Administration's (ERDA) Fast Flux Test Facility (FF'fF).
The ACRS reported previously on this project on July 13, 1971, January 13, 1972, and May 18, 1973. Since the last report, Subcom-mittee meetirgs were held in Richland, Washirgton, on July 19, 1974, and in Denver, Colorado, on March 15, 1975. A site visit was made on July 20, 1974, and the project was considered by the Committee durirg its Special Meetir.g, October 31-November 2, 1974, and durirg its 180th Meetirg, April 3-5, 1975. Durirg its review the Committee had the benefit of discussions with representatives and consultants of the Division of Reactor Research and Development of the ERDA, the Hanford Er:gineerir.g Development Laboratory (HEDL) of the Westir.ghouse Hanford Company, the Advanced Reactors Division of the Westir.ghouse Electric Corporation and the NRC Staf.f. The Committee also had the benefit of the documents listed below.
At this stc:ge of its review, the Committee has been asked to consider and comment on two questions:
(1) whether sealir.g the head cavity in the manner proposed by HEDL would contribute s:igni.ficantly to safety; and (2) which of several alternate measures proposed by HEDL should be adopted for the space provided below the guard vessel.for use in connection with a possible ex-v~ssel post-accident core retention system.
These questions, in one form or another, have been of concern to the ACRS durirg its entire review of this facility and have been discussed in previous reports. In its report of May 18, 1973s the Committee recommended the development of extensive additional information on 515
Honorable William July 15, 1975 postulated accidents. The Committee recommended also that the necessary further rEgulatory review of the desjgn basis work energy release and the requirements with rEgard to post-accident heat removal be scheduled and accomplished in timely fashion so that additional features, if necessary, could be provided prior to the scheduled reactor startup.
Additional information on postulated accidents has been developed by the contractor and extensive :further review has been carried out by the NRC Staff. Al thotgh uncertainties remain, the NRC Staff has been able to make certain recommendations with which the ACRS generally concurs.
The NRC Staff has concluded that sealir:g the head cavity in the manner proposed by HEDL would not contribute sjgnificantly to safety. The ACRS agrees with this conclusion.
With regard to the space beneath the guard vessel, the contractor has indicated that a large amount of additional research and development would be required to desjgn an ex-vessel post-accident core retention device and that the FFTF schedule would be delayed a matter of years if this course were followed. After considerir:g various alternatives, the contractor recommended that this space be filled with concrete. The NRC Staff has concluded that the need for an ex-vessel core-retention device is small but cannot be jgnored. The Staf.f has reconmended that the existir:g space should be retained so as not to make impossible the future installation of such a device if :further studies or charges in the nature or use of the facility should indicate its desirability.
Thd ACRS concurs in the recommendation of the NRC Staff.
The NRC Staff has recommended also that hot liners should be installed wherever sodium could accumulate followir:g a release into the reactor cavity, and that cold liners in this cavity be vented. The ACRS concurs in these recommendations.
The NRC Staff has recommended further that emergency plans be prepared for the FFTF pursuant to 10 CFR Part 50, Appendix E.
The ACRS agrees with this recommendation.
For those postulatea accident sequences.for which ventir:g of the contain-ment mjght be desirable in order to prevent it from beir:g overpressured the ACRS sqmests that consideration be given to the possible usefulness of sand-and-gravel filters for the removal of airborne particulates.
The Committee wishes to point out that the FFTF is a special test facility, and that both the positive and nEgative aspects of this circumstance have been considered tbrotghout the review of this project_. The ACRS believes that the desjgn of the FFTF and the review of its safety aspects should not be used as a precedent for establishir:g the safety criteria for commerical liquid metal fast breeder reactors.
516
Honorable William July 15, 1975 The Final Sa:fety Analysis Report (FSAR) for the'FFTF is now in pre-paration.
The Committee expects the FSAR to provide a comprehensive treatment or the accident considerations, the contaimnent capability, and the supportir:g research, development, analysis, and erg ineerir:g.
The Committee cautions that, because the prcgram plans for the FFTF call for its use to perform a wide rarge of experiments usir:g new fuels under a variety of conditions, the safety aspects of which have not yet been examined, there will be a continuir:g need to review the adequacy of the safety :features provided.
The Advisory Committee on Reactor SafEguards believes that, if due rEgard is given to the matters mentioned above, and in previous reports, it is acceptable for contruction of the FFTF to proceed. The ACRS expects to continue to review this project after the Final Safety Analysis Report has been received.
Sincerely,
./JJ1~
W. Kerr Chairman ADDITIONAL COMMENTS BY MEMBER D. OKRENT I generally concur with this report.
I would like specifically to note the NRC Staff recommendation that it is an ERDA contractor's obljgation to show that the likelihood of a core discruptive accident is small and if it occurs, the energetics would not exceed the capability of the containment system either by penetration or overpressurization.
Also, I would like to observe that in view of the orjginal and remain-irg continuir:g uncertainties with rEgard to the possible energy yield and mechanical work yield in low probability, postulated core-disruptive accidents, the behavior and disposition of core material followirg postulated accidents leadirg to gross fuel meltir:g, and the efficacy of in-vessel lorg-tenn core coolir:g followirg possible accj.dents, a quantitative assessement of the adequacy of the currently desjgned con-tainment system of FFTF is difficult.
I believe that, had the safety desjgn philosophy pursued by the contractor been one of achievement, within practical considerations, of a near-maximization of the primary containment capability to withstand the mechanical effects of postulated core-disruptive accidents, and one of the early and timely development of an ex-vessel core retention system, an awkward and possibly undesirable situation mjght have been alleviated.
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Honorable William July 15, 1975 REFERENCES TO FFTF LETTER:
Westir:ghouse Advanced Reactors Division Report, FRT 1561 Rev. 1, entitled:
"Ex-Vessel Core Catcher Design Requirements and Preliminary Concepts Evaluation," dated June 14, 1974 Hanford Ergineerir:g Development Laboratory letter dated September 16, 1974 (W/FFTF 7410545) conce~ evaluation of FFTF Head Compartment Argonne National Laboratory report entitled: "Summary Report on the Analysis of a Loss-of-flow (without scram) Accident in the FFTF," dated November 1974 Hanford Er:gineerir:g Development Laboratory report entitled: "Post-Accident Heat Removal Assessment for the FFTF," dated November 1974 Hanford Er:gineerir:g Development Laboratory Preliminary Report entitled:
"Post-Accident Heat Removal Containment Transients," dated November 18, 1974 Hanford Ergineerir:g Development Laboratory Preliminary Report entitled:
"Radiolcgical Evaluation of a Postulated FFTF Core Melt-Thro~h Accident,"
dated November 19, 1974 Westir:ghouse Advanced Reactors Division Report, WARD-2171-46, entitled:
"EK-Vessel Core Catcher (EVCC) Design Study for FFTF.," dated December 1974 Directorate of Licensir:g, US Atomic Energy Commission, Supplement No. 1 to the Safety Evaluation of the Fast Flux Test Facility, Issued:
December 13, 1974 Division of Reactor Licensir:g, US Nuclear Regulatory Commission, Supplement No. 2 to the Safety Evaluation of the Fast Flux Test Facility, Issued:
March 7, 1975 518