ML19220A453
| ML19220A453 | |
| 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) |
| Shared Package | |
| ML19220A449 | List: |
| References | |
| NUDOCS 7904180017 | |
| Download: ML19220A453 (3) | |
Text
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ADVISORY COMMITTEE ON REACTOR SAFEGUARDS UNITED STATCS ATOMIC ENERGY COMMISSION W ASHINGToN. O C.
20345 July 17, 1969 Honorable Glenn T. Seaborg Chairman U. S. Atomic Energy Commission Washington, D. C.
20545
Subject:
REPCRT ON THREE MILE ISLAND NUCLEAR STATION UNIT 2
Dear Dr. Sea'org:
o At its llita meeting, July 10-12, 1969, the Advisory Committee on Reactor Safeguards reviewed the proposal of the Metropolits Edison Company and the Jersey Central Power and Light Company to construct Unit 3 at the Three Mile Island Nuclear Station. A Subcommittee also met to review this project on June 26, 1969. During its review, the Committee had the beoefit of discus-sions with representatives and consultants of both applicanta, the Bal cock and Wilcox Company, Burns and Roe, Cnc., General Pub 1'. Utilities Corp.,
and the AEC Regulatory Staff. The Committee also had available the docu-ments listed below.
The plant will be located adjacent to Unit 1 on Three Mile Island 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 alt eraf t hardening protection are similar to those of Unit 1, noted in our January 17, 1968, and April 12, 1968, reports. Unit 2 will be operated at a power level of 2452 FMt.
Review of Unit 2 has taken into account the similarities of the Three Mile Island units. new features, updating of the research and development
- programs, and further evaluations of the site.
The review also included matters previ-ously identified that warrant careful consideration for all large, water-cooled power reactors; the Connittee believes that resolution of these matters should apply equally to this reactor.
The estimate of probable maximum flood discharge in the Susquehanna River the site is being revised upward; by the U. S. Army Corps of Engineers at and will be larger than had been considered in the design of dnit 1.
The applicant has sr,ted that both units will be protected by measures which would assure a sata, orderly shutdown of the reactors in the event of the maximum flood.
AB g,D,,
Honorable Glenn T. Seaborg July 17, 1969 The applicant has conducted a test program in support of his proposal to grout the stranded tendons for the containment prestressing system. The Committee believes that adequate grouting can be attained through proper and careful execution of the procedures developed in this program. The applicant has proposed a program of periodic proof testing at 1157. of design pressure to monitor the integrity of the containment, which has been designed cons,erva.
tively to obviate any adverse effects of repeated proof tescing at th4s high pressure. The Ccmmittee believes that such a program, involving measurement 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 review is necessary of the research and development being compierec for the alkaline sodium thiosulfate spray additive to determine whether the spray systems as proposed need augmentation to achieve required performance in' postulated accidents.
Provisions will be 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 accident to doses significantly belev the 10 CFR 100 guideline values.
The applicant has been considering a purge system to cooe with potential hydrogen buildup from various sources in the unlikely event of a loss-of-coolant accident. Additional studies ;ce needed to establish the accepta-bility of thio system and to consider aitarnative approaches. These studies should include allowance for levels of circaloy-water reaction which could occur if the ef fectiveness of the emergency core cooling system were signifi-cantly less than predicted. The Committee believes that this matter can be resolvet durin? construction of the reactor.
The Committee reiterates ito belief that the instrumentation design should be reviewed for common failure modes, taking inte account the possibility of systematic, non-randem, conc 2rrent failures of redundant devices, not con-sidered in the single-f ailur e criterion. The applicant should show that the proposed interconnection of control and safety instrumentation will not adversely affect plant safety in a significant manner, considering the possibility of systematic component failure.
The Committee believes that this matter can be resolved during construction of the reactor.
The Committee believes that, for transients having a high probability of occurrence, and for which action of 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. Common failure modes must be considered in ascertaining an acceptable level of protection. The Committee recommends that a study be made of the possible consequences of hypothesized failures of protective systems during anticipated transients, and of steps to be taken if needed. The Committee believes that this ratter can be resolved during construction of the reactor.
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Henorable Glenn T. Seaborg July 17, 1969 The Committee recommends that the applicant study possible means of in-service monitoring for vibration or for the pzesence of loose parts in the reactor pressure vessel as well as in other pertions of the primary system, and implement such means as are found practical and appropriate.
The post-accident cooling oystem must retain its integrity throughout the course of an accident and the subsequent cooling period. The applicant she 'J review the effects of coolant temperature, pH, radioactivity, cor-rosive materials from the core or other parts of the containment (including stored chemicals), and potentially abrasive slurries.
Degeneration of ccm-ponenr.s such as filters, pump impellers, and seals by any of these mechanisms should be reviewed.
Particular attention should be paid to potential problems arising from the use of dissimilar metals in these systems.
The Committee recommends that details concerning the adequacy of the design, the material characteristics, quality assurance, and in-service inspection requirements of the main coolant pump flywheels be resolved between the applicant and the Regulatory Staff.
In this connection, and, in general, the Committee continues to emphasi::e the need and importance of quality assurance, in-service inspecticn and monitoring programs, as well as con-servative safety margins in design.
The Advisory Committee on Reactor Safeguards believes that the.it emhen;,
tiened 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 nealth and safety of the public.
Sincerely yours, A.
x
-L 1 3 j,0 g
c r.O_s O_ttcute_. v ;
. ' ' ' Stephen H. Hancuer
^
Chairman
References:
1.
Three Mile Island Nuclear Station - Unit 2, Preliminary Safety Analysis Report, Volumes 1-4 (Amendment No. 6, Oyster Creek Nuclear Statica, Unit 2, Docket No. 50-320).
2.
Amendments 7 - 10 to Application for Licenses.
3.
Metropolitan Mison Co.30any letter dated July 3, 1969.
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