ML19256D837
| ML19256D837 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 07/17/1969 |
| From: | Hanauer S Advisory Committee on Reactor Safeguards |
| To: | Seaborg G US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| ACRS-DR-2262, NUDOCS 7910230787 | |
| Download: ML19256D837 (3) | |
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FROM CONTROL NUMBER ACTION COMPLETION DEADLINE 60 4d Stephen 11. lianauer, Chat n
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ADVISORY COMMITTEE ON REACTOR SAFEGUARDS yg 7' UNITED STATES ATOMIC ENERGY COMMISSION WASHINGTON. D.C. 20545 July 17, 1969 Honorable Glenn T. Seaborg 0 9*
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Subject:
REPORT ON THREE MILE ISLAND NUCLEAR STATION UNIT 2 Dear Dr. Seaborg At its lilth meeting, July 10-12, 1969, the Advisory Coc:mittee on Reactor Safeguards reviewed the proposal of the Metropolitan Edison Company and the Jersey Central Power and Light Company to construct Unit 2 at the Three Mile _.__ _
Island Nuclear Station. A Subco=mittee also met to review this project on June 26, 1969. During its review, the Cc=:sittee had the benefit of discus-sions with representatives and consultants of both applicants, the Babcock and Wilcox Company, Burns and Roe, Inc., General Public Utilities Corp.,
and the AEC Regulatory Staff. The Cocznittee also had available the docu-ments listed below.
The plant vill br located adjacent to Unit 1 on Three Mile Is1'and near the east shore of the Susquehanna River, about 10 miles southeast of Harrisburg,'
Pennsylvania. The nuclear steam supply system, engineered safety features, reactor building, and aircraf t hardening protection are similar to those of Unit 1, noted in our January 17, 1968, and April 12, 1968, reports. Unit 2 vill be operated at a power level of 2452 W e.
Review of Unit 2 has taken into account the similarities of the Three Mile Island units, nca features, updating of the research and development progrs=s, and further evaluations of the site. The review also included matters previ-ously identified that warrant careful consideration for all largo, water-cooled power reactors; the Co::xnittee believes that resolution of these matters should apply equally to this reactor.
The estimate of probable maximum flood discharge in the Susquehanna River at the site is being revised upwards by the U. S. Ar=y Corps of Engineers and vill be larger than had been considered in the design of Unit 1.
The applicant has stated that both units will be protected by measures which would assure a safe, orderly shutdott:2 of the reactors in the event of the maxi =um flood.
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1454 046 Rec'd Off.j0ir.pf Reg.
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Eonorable Glenn T. Ser.borg
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July 17, 1969 The applicant has conducted a test program in support of his proposal to grout the stranded tendons for the containment prostressing system. The Co=mittee believes that adequate grouting can be attained through proper and careful execution of the procedures developed in this program. The applicant hea proposed a program of periodic proof testing at 1157. of desige pressure to conitor the integrity of the containment, which has been designed conserva-tively to obvisto any adverso effects of repeatri proof testin; at thic high pressuro. The Co=mittoo balin es that such a program, involving measur:=cnt of deformations and thorough inspection for cracking of'the concrete during each proof test, will provide reasonable assurance of the continued integrity of the containment.
Further revicu is necessary of the roscarch and developc:ent being co=pleted for the alkaline sodium thiosulfate spray additive to determino whether the spray systc=s as proposed need augmentation to achiovo required performaace in postulated accidents. Provisions will bo incorporated in the design of the containment system to permit equipment additions if necessary to ensure limiting the radiological consequences of a loss-of-coolant accide.nt to doses significantly below the 10 CFR 100 guideline values.
The applicant has been considering a purge system to cope with potential hydrogen buildup from various sources in the unlikely event of a loss-of-coolant accident. Additional studies are needed to establish the accopta-bility of this system and to consider altetuativo cpproaches. Those studies should include allowance for levels of zircaloy-water reaction which could occur if the effectiveness of the c=crgency core cooling system were signifi-cantiv less than predicted. The Committee believes thz.c this matter can be roso. icd during construction of the reactor.
The Co=mittee reiterates its belief that the instrumentation design should be revicwed for co=on failure modes, taking into account che possibility of systematic, non-random, concurrent failures of redundant devices, not con-sidered in the single-failure critorion, The applicant should show that the proposed interconnection of control and safety instrumentation will not adversely offect plant safety in a significant manner, considering the possibility of systematic component failure.- The Co=mitteo believes that this catter can be resolved during construction of the reactor.-
The Cc:=2ittee believes that, for transients having a high probability of occurrence, and for which action ot a protective system or other engineered safety feature is vital to the public health and safety, an exceedingly high probability of successful action is needed., Co==on failure modos =ust be considered in ascertaining an acceptable level of protection. The Co=mittee race =: ends that a study be made of tbo possible consequencea of hypothosiaed fa;.lurcs of protectivo systems during anticipated transients, and of steps to be taken if nocded. The Committee believes that this matter can be resolved during construction of the reactor..
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Honorable Glenn T. Seaborg July 17, 1969 The, Committee recommends that the applicant study possible means of in-service conitoring for vibration or for the presence of loose parts in the reactor pressure vessel as well as in other portions of the primary system, and imple=ent such means as are found practical and appropriate.
The post-accident' cooling system must retain its integrity throughout the course of an accident and the subsequent cooling period. The applicant should review the effects of coolant temperature, pH, radioactivity, cor-rosive materials from the core or other parts of the containment (12cluding stored chemicals), and potentially abrasive slurries. Degeneration of com-ponents such as filters, pump impellers, and seals by any of these mechanis=s should b2 reviewed. Particular attention should ba paid to potential problems arising from tha use of dissimilar metals in these systems.
The Committee reco::raends that details concerning the adequacy of the design, the material characteristics, quality assurance, and in-service inspection requirements of the main coolant-pump flyuheels be resolved between the applicant and the Regulatory Staff.
In this connection, and, in general, the Connittoo continues to emphasi:;e the need and iciportance of quality assurance, in-service inspection and monitoring programs, as well as con-servative safety margins in design.
The Advisory Co=mittee on Reactor Safeguards believes that the items men-tioned can be resolved during construction, and that, if due consideration is given to the foregoing, Unit 2 proposed for the Three Mile Island site can be constructed with reasonable assurance that it can be operated with-out undue risk to the health and safety of the public.
Sincerely yours, Original Signed b7' Stephen H. Hanauer Stephen H. Hanauer Chairman
References:
1.
Threc Mile Island Nucicar Station - Unit 2, Preliminary Safety Analysis Report, Volu=es 1-4 (A=endment No. 6, Oyster Creek Nuclear Station, Unit 2 Docket No. 50-320).
2 Amend =ents 7 - 10 to Application for Licenses.
3.
Metropolitan Edison Company letter dated July 3, 1969.
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