ML19220A948
| ML19220A948 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 07/17/1969 |
| From: | Hanauer S Advisory Committee on Reactor Safeguards |
| To: | Seaborg G NRC COMMISSION (OCM) |
| Shared Package | |
| ML19220A944 | List: |
| References | |
| NUDOCS 7904250123 | |
| Download: ML19220A948 (3) | |
Text
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ADVISORY COMMITTEE ON REACTOR SAFEGUARDS UNITED STATES ATOMIC ENERGY COMMISSION WASHINGTON, D.C. 20545 July 17, 1969 Honorable Glenn T. Seaborg Chair =an U. S. Atomic Energy Cocznission Washington, D. C.
20545
Subject:
RE?CRT CN THREE MILE ISLAND NUCLEAR S~ATION UNIT 2
Dear Dr. Seaborg:
At its lilth =eeting, July 10-12, 1969, the Advisory Concittee on Reactor Safe 3uards revicved the proposal of the :letropolitan Idison Company and the Jersey Central Fever and Light Company to construct Unit 2 at the Three Mile Island Nucicar Station. A Subco=sittee also =ct to revie:i this project on June 26, 1969. During its review, tha Ces::itteo had the benefit of discus-siens with representatives and consultants of both applicanta, the F.acock and Wilcox Ccmpany, Burns and Rce, Inc., General Public Utilities Corp.,
and the AEC Regulatory Staff. The Cc=sittee also had available the docu-
=ents listed below.
The plant will be located adjacent to Unit 1 on Ihree 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 aircraft hardenin; protection are similar to those of Unit 1, noted in our January 17, 1963, and April 12, 1963, reports. Unit 2 vill be operated at a power lerel of 2452 Mit.
Review of Unit 2 has taken into account the similarities of the Three Mile Island units, new features, updating of the research and developacnt progrn=s, and further evaluations of the site. The review also included natters previ-ously identified that warrant careful consideration for all large, water-ecoled pcuer reactors; the Cocznittee believes that resolution of these matters should apply equally to this reactor.
The esti= ate of probable cax4-flood discharge in the Susquehanna River at the site is being revised upwards by the U. S. Ar=y Corps of Engineers and will be larger than had been considered in the design of Unit 1.
Se applicant has stated that both units will be protected by nessures which sculd assure a safe, orderly shutdevn of the reactors in the event of the maxi =um flood.
7904U *l23
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July 17, 1969 Honorable Glenn T. Seaborg t of his proposal to grout The applicant has conducted a te',,A 7 gram in suppor The Co=mittee ar a nc prestressing system. attained through l
the stranded tendons for the conto The applicant has believes that adequate grouting can beexecution of the p ram.
proposed a program of periodic proof testing a hich has been designed conderva-
=enitor the integrity of tha contain=ent, wf repeated proof testing at this hig l i g measure =ent tively to obviate any adverse effects oThe Cc==ittee believes that suc of deformations and thorough inspection for cracce of the continued integrity pr e s sur e.
each proof test, will provide reasonable assuran of the contain=ent.
development being cc=pleted Further review is necessary of the research andfor determine whether the f
spray syste=s as proposed need aus:entation toProvisions will be incorporat in postulated accidents.systes to' permit equip =ent additions if necessart l
liniting the radiological consequences of a loss-of-coo an the contain=ent lues.
significantly below the 10 CFR 100 guideline va ope with potential The applicant has been considering a purge system to c nt of a loss-of-l hydrogen buildup from various sources in the unlike y eveA ta-These studies aches.
bility of this system and to consider alternative approter reaction which coolant accident.
l should include allowance for levels of =irco oy-wa oling syste= vere signifi-occur if the ef fectiveness of the emergency core coThe Cc=mittee belie cantly less than predicted.
resolved during construction of the reactor.
tion design should be The Committee reiterates its belief that the instrumentaunt the possib reviewed for ce==en f ailure modes, taking into accofailures of redundant d the The applicant should show that syste=stic, non-random, concurrent sidered in the single-f ailure criterion.
instrumentation will not proposed interconnection of control and safetyadve er, consr iering the possibility of systematic component feilure. tion of the reactor.
this matter can be resolved during consecuc for transients h;ving a high probability ofvst The Committee believes that, i
occurrence, and for which action of a protect ve.d safety, an exceedingly hig safety feature is vital to the public health anCc==cn failur. = odes =ust be is needed.
The Co=aittee probability of successful action level of protection.
ble consequences of hypothesized l
considered in ascertaining an acceptab e recc=: ends that a study be =ade of the possif ailur and of steps this =atter can be The Co=sittee believes that to be taken if needed.
resolved during construction of the reactor.
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m 3-July 17, 1969 Honorable Glenn T. Seaborg The Co=cittee reco==cnds that the applicant study possible = cans of in-service
- onitoring for vibration or for the presence of icose parts in the reactor pressure vessel as well as in other portions of the pri=ary syste=, and i=ple=ent such =eans as are found practical and appropriate.
The post-accident cooling system =ust retain its integrity throughout the course of an accident and the subsequent cooling period. The applicant should review the effects of coolant te=perature, pH, radi% ctivity, cor-rosive catorials frc= the core or other parts of the contai=ent (including stored che=icals), and potentially abrasive slurries. Jegeneration of co -
ponents such as filters, pump impellers, and seals by any of these =echanis=s should be reviewed. Particular attention should be paid to potential problems arising frc= the use of dissimilar metals in these systc=s.
Ihe Co:nnittee reco::cends that details concerning the adequacy of the design, the :aterial characteristics, quality assurance, and in-service inspection require =ents of the = sin coolant-pu=p flywheels be resolved between the applicant and the Regulatory Staff. In this connection, and, in general, the Co==ittee continues to e=phaaire the need and i=portance of quality assurance, in-service inspection and c:enitoring programs, as well as con-servative safety margins in design.
The Advisory Co==ittee on Reactor Safeguards believes that the ita=s =en-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 health and safety of the public.
Sincerely yours, Crigin2; Sir.ed W Stephen n. Eazagg Stephen H. Hanauer Chairman
References:
1.
Three Mile Island Nuclear Station - Unit 2, Preli=inary Safety Analysis Recert, Volu=es 1-4 (A=e=d=ent No.,6, Oyster Creek !uclear Station, Unit 2 Dochet No. 50-320).
2.
Amend =ents 7 - 10 to'Applicarin=-for I.icenses.
2.
fictropolitan Edison Co=nany letrer dated July 2,1969.
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