ML19206A855
| ML19206A855 | |
| 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 | |
| ML19206A848 | List: |
| References | |
| NUDOCS 7904210544 | |
| Download: ML19206A855 (3) | |
Text
.
ADVISORY COMMITTEE ON REACTOR SAFEGUARDS UNITED STATES ATOMIC ENERGY COMMf CSION WASHINGTCN. D.C.
23343 July 17, t969 Honorable Glenn T. Seaborg Chairman U. S. Atomic Energy Commission Washington, D. C.
20545
Subject:
REPORT ON THREE MILE ISIAND NUCLEAR STATION UNIT 2
Dear Dr. Seaborg:
At its lilth meeting, July 10-12, 1969, the Advisory Committee on Reactor Safeguards reviewed the proposal of che Metropolitan Edison Company ar.d the Jersey Central Power and Light Company to construct Unit 2 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 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 Staf f.
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 aircraf 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 FNt.
Review of Unit 2 has taken into account the similarities of the Three Mile Island units, new features, updating of the research and developmeat 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 Committee believes that resolution of these r.atters should apply equally to th.s reactor.
The estimate of probable maximum flood discharge in the Susquehanna River the site is being tevised upwards by the U. S. Anny Corps of Engineers at and will be larger than had been considered in the design of Unit 1.
The applicant has stated that both units vt11 be protected by measures which would assure a safe, orderly shutdown of the reactors in the event of the maximum fleod.
g)()M1 h
' ' Ib 4
q
~
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 conserva-tively to obviate any adverse effects of repeated proof testing at this high pressure. The Committee believes that such a program, involving measurement of deformations and charough inspection for cracking of the concre:e 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 completen 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 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 accepta-bility of this system and to consider alternative approaches. These studies should include allowance for levels of zircaloy-water reaction which could occur if the effectiveness of the emergency core cooling system were signifi-cantly less than predicted. The Committee believes that this matter can be resolved during construction of the reactor.
The Connittee reiterates its belief that the instrumentation design should be reviewed for common failure modes, taking into account the possibility of systematic, non-random, concurrent failures of redundant devices, not con-sidered in the single-f ailure 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 reactot.
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 rec:mmends that a study be made of the possible consequences of hypothesized f ailures of protective systems during anticipated transients, and of steps
- o be taken if needed. The Committee believes that this matter can be cesolved during construction of the reactor.
62 105 4
g
+
Honorable Glenn T. Seaborg July 17, 1969 The Committee reco;mmends that the applicant study possible means of in-service monitoring 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 implement such means as are found practical and appropriate.
The post-accident cooling system must retain its integrity thrcughout 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 (including stored chemicals), and potentially abrasive slurries. Degeneration of com-ponents 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, tr.= :naterial 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 emphasize the need and importance of quality assurance, in-service inspection and monitoring programs, as well as con-servative safety margins in design.
The Advisory Committee on Reactor Safeguards believes that the+ 4t em s meny 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, I
J.
~3 f
. %D, x
-L.
f Ce acu a. v,
,1<
ad w
.~ Stephen H. Hanauer Chairman References-
'.. Three Mile Island Nuclear Station - Unit 2, Preliminary Safety Analysis Report, Volumes 1-4 (i.mendment No. 6, Oyster Creek Nuclear Station, Unit 2, Docket No. 50-320).
2.
Amendments 7 - 10 to Application fo Licenses.
3.
Mecropolitan Edison company letter dated July 3,1969.
c.,,
u..
. t,r 2