ML19322C567
| ML19322C567 | |
| 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 | |
| TASK-TF, TASK-TMR NUDOCS 8001170893 | |
| Download: ML19322C567 (3) | |
Text
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ADVISORY COMMITTEE ON REACTOR SAFEGUARDS
(.
UNil'ED SI'ATES ATOMIC ENERGY COMMISSION WAculNGToN, D.C.
July 17, 1969 lionorable Glenn T. Seaborg Chaiman U. S. Atomic Energy Counnission Washington, D. C. 20545
Subject:
REPORT ON TilREE MILE ISIAND NUCLEAR STATION UNIT 2
Dear Dr. Seaborg:
At its 111th meeting, July 10-12, 1969, the Advisory Comittee on Reactor Safeguards reviewed the proposal of the tktropolitan Edison Cotapany and the Jersey Central Power and Light Company to construct Unit 2 at the Three Mile Island Nucicar Station. A Subconnaittee also met to review this project on June 26, 1969. During its review, the Cou:nittee had the benefit of discus-sions with representatives and consultants of both applicants, the Babcock and Wilcox Company, Burns and Roe, Inc., Ceneral Public Utilities Corp.,
{i and the AEC Regulatory Staff. The Counittee also had available the docu-nents listed below.
The plant will be located adjacent to Unit 1 on Three Mile Island near the cast shore of the Susquehanna River, about 10 miles southeast of llarrisburg, Pennsylvania. The nuclear steam supply system, englueered safety features, reactor building, and aircraft hardening protection are similar to those of Unit 1, noted in our January 17, 1968, and April 12, 1968, reports. Unit 2 will be ope rated at a power level of 2452 IGt.
Review of Unit 2 has taken into account the similarities of the Three Mile Island units, new features, updating of the research and development prograas, and rther evaluations of the site. The review also included matters previ-identified that warrant careful consideration for all large, water-
.ed power reactors; the Coranittee 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. Army Corps of Engineers and will be larger than had been considered in the design cf Unit 1.
The applicant has stated that both units will be protected by measures which would assure a safe, orderly shutdown of the reactors in the event of the maxiunim flood, C
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C O
P Y July 17, 1969 Honorable Glenn T. Seaborg The applicant has conducted a test program in support of his proposal to grout the stranded tendons for the containment prestressing system. The Con:nittee believes that adequate grouting can be attained through proper and careful execution of the procedures develeped in this program. The applicant has 1157. of design pressure to proposed a program of periodic proof testing at monitor the integrity of the containnent, which has been designed conserva-tively to obviate any adverse ef fects of repeated proof testing at this high The Cocuittee believes that such a program, involving measurement of defornations and thorough inspection for cracking of the concrete during pressure.
each proof test, will provide reasonabic assurance of the continued integrity of the containa2nt.
Further review is necessary of the research and development being coap ated l
for the alkaline sodium thiosulfate spray additive to determine whether the spray systems as proposed need augmentation to achieve required performance Provisions will be incorporated f.n the design of in postulated accidents.
the containment system to permit equipment additions if nec essary to ensure accident to doses limiting the radiological consequences of a loss-of-coolant 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-Additional studies are needed to establish the accepta-coolant accident.
These studies bility of this system and to conaid9r alternative approaches.
should include allowance for levels of zircalay-water reaction which could occur if the ef fectiveness of the emerdency core cooling system were signifi-cantly less than predicted. The Committee believes that this matter can be resolved during construction of the reactor.
The Committee reiterates its belief that the instrumentation design should be reviewed for common failure modes, taking into account the possibility of s; stematic, non-randan, concurrent failures of redundant devices, not con-sidered in the single-failure criterion. The applicant should show that the proposed interconnection of control and safety instrumentation will not adversely af fect plant safety in a significant manner, considering the The Committee believes that possibility of systematic component failure.
this matter can be resolved during construction of the reactor.
for transients having a high probability of The Comm.ttee believes that, and for which action of a protective system or other engineered safety feature is vital to the public health and safety, an exceedingly high occurrenc<
Connon failure modes must be probability of successful action is needed.
The Conaittee considered in ascertaining an acceptable level of protection.
recommands that a study be made of the possibic consequences of hypothesized failures of protecr *ve systems during anticipated transients, and of steps l
f to be taken if neeled. The Conmittee believes that this matter can be resolved during construction of the reactor.
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C O
P July 17,1969 Y Honorable Glenn T. Seaborg f in-service plicant study possible means o ctor for the presence of loose parts in the reaof the pri The Cocanittee recon = ends that the ap monitoring for vibration or h r portions pressure vessel as well as in ot ed practical and appropriate.
the implement such means as are foun t retain its integrity tThe applicant hroughout bsequent cooling period.
The post-accident cooling system mus r-lant temperature, pH, radioactivity, coof the co course of an accident and the su should review the effects of coorosive materials other parts Degeneration of cen-h nisms lly abrasive slurries.llers, and seals by any of these stored chemicals), and potentia Particular attention should be paii ilar metals in t ponents such as filters, pump impe should be reviewed.
h design, arising from the use of diss m concerning the adequacy of t einspection i
recommends that detailsquality assurance, and in-serv ceresolved betw The Committee l
pump flywheels be the material characteristics, In this connection, and, in genera,
requirements of the main coolant-d and importance of quality Staff.
ll as con-applicant and the Regulatorycontinues to emphasize the neei n and moni the Connittee assurance, in-service incpect o design.
servative safety margins in f guards believes that the items men-if due co The Advisory Committee on Reac';or Sa eresolved during ce d that, d site
/c tiened can be is given to the foregoing, Unit assurance that VJ.
ble f the public.
can be constructed with reasonarisk to the health and safety o out undue Sincerely yours,
/s/ Stephen H. Hanauer Stephen H. Hanauer Chairman i
Unit 2, Preliminary Safety Analys s Three Mile Island Nuc1 car Station -6, Oyster Creek Nuclear
References:
Report, Volumes 1-4 (Amendment No.
1.
Unit 2, Docket No. 50-320).
i ses.
Amendments 7-10 to Application for L cendated July 3,1969 2.
Metropolitan Edison Company letter 3.
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