ML25196A431
| ML25196A431 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 10/22/1976 |
| From: | Moeller D Advisory Committee on Reactor Safeguards |
| To: | Rowden M NRC/Chairman |
| References | |
| Download: ML25196A431 (1) | |
Text
ADVISORY COMMITTEE ON REACTOR SAFEGUARDS NUCLEAR REGULATORY COMMISSION Honorable Marcus A. Rowden Chairman WASHINGTON, D. C. 20555 October 22, 1976 U.S. Nuclear Regulatory Commission Washington, DC 20555
SUBJECT:
REPORT ON THREE MILE ISLAND NUCLEAR STATION, UNIT 2
Dear Mr. Rowden:
During its 198th rreeting, October 14-16, 1976, the Advisory Committee on Reactor Safeguards completed its review of the application of the Metropolitan Edison Company, Jersey Central Power and Light Company, and Pennsylvania Electric Company (Applicants) for a license to operate Three Mile Island Nuclear Station, Unit 2.
This project was also con-sidered during a Subcommittee meeting held in Harrisburg, Pennsylvania, on September 23 and 24, 1976.
Members of the Committee visited the facility on September 23, 1976.
During its review, the Comnittee had the benefit of discussions with representatives and consultants of the Applicants, General Public Utilities Service Corporation, the Babcock and Wilcox Company (B&W), Burns and Rowe, Inc., and the Nuclear Regula-tory Commission (NRC) Staff. The Committee also had available the documents listed below.
The Committee reported on the application for a construction permit for Unit l on January 17 and April 12, 1968, and for an operating license for Unit l on August 14, 1973.
The Comnittee reported on the application for a construction permit for Unit 2 on July 17, 1969.
The Three Mile Island Nuclear Station, Units 1 and 2, is located on Three Mile Island near the eastern shore of the Susquehanna River, about 12 miles southeast of Harrisburg, Pennsylvania. About 2380 people liye within a two-mile radius of the site (the low population zone).
The minimum exclusion distance is 2000 feet.
The nearest population center is Harrisburg (1970 population 68,000).
Several changes have been made to bring the Babcock and Wilcox Emergency Core Cooling System (ECCS) evaluation model into conformance with the requirements of 10 CFR 50.46, and Appendix K to Part 50.
Analyses of a spectrum of break sizes appropriate to Three Mile Island, Unit 2 have been completed using the approved B&W generic evaluation model.
The 1649
Honorable Marcus October 22~ 1976 results of the analyses for the reactor coolant pump discharge break, believed to be the "worst" break, show maximum allowable linear heat generation rates as a function of elevation in the reactor care ranging from 15.5 to 18.0 kilowatts per foot. Corresponding calculated post-accident peak clad temperatures range from 20020F to 2146°F.
The NRC Staff has identified additional information that it will require to complete its review and the Applicants' submittal is expected by the end of 1976. The Applicants propose to use both in-core and ex-core instrumentation to assure accuracy of measurement of core power distri-butions. The Committee believes that the proposed monitoring methods may be acceptable, but that an augmented startup program should be employed, and that satisfactory experience at 100% steady state power and during transients at less than full power should be obtained. This experience should be reviewed and evaluated by the NRC Staff prior to operating at up to full power in a load following mode.
The Committee wishes to be kept informed.
A question has arisen concerning asymmetric loads on the reactor vessel and its internal structures for certain postulated loss-of-coolant accidents in pressurized water reactors. The Staff has required the Applicants to supply further information in order to complete its assess-ment of this matter. This issue should be resolved in a manner satis-factory to the NRC Staff.
The question of whether Unit 2 requires design modifications in order to comply with WASH-1270, "Technical Report on Anticipated Transients Without Scram for Water-Cooled Power Reactors 11, remains an outstanding issue pending the NRC Staff's completion of its review of B&W generic analyses of anticipated transients without scram.
The Committee recom-mends that the NRC Staff, the Applicants and B&W continue to strive for an early resolution of this matter in a manner acceptable to the NRC Staff. The Committee wishes to be kept informed.
Emergency plans have been developed to allow plant shutdown and mainte-nance of safe shutdown in the event of a maximum probable flood.
Such a postulated flood would top the levee surrounding the plant by several feet. Included in the plan is the fastening of water tight steel panels in doorways and other openings of safety related structures. The Com-mittee believes that the details of this plan, particularly relating to re-entry into the station during the post-flood period, need to be more clearly delineated.
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Honorable Marcus October 22, 1976 The Committee supports the NRC Staff's program for evaluation of fire protection in accordance with Branch Technical Position APCSB 9.5-1, Appendix A, "Guidelines for Fire Protection for Nuclear Power Plants".
The Corrmittee recommends that the NRC Staff give high priority to the completion of both owner and Staff evaluations and to recommendations for Three Mile Island Unit 2 and other plants nearing completion of construction in order to maximize the opportunity for improving fire protection while areas are still accessible and changes are more feasible.
The Committee notes that long-term post-accident operation of the plant to maintain safe shutdown conditions may be dependent on instrumentation and electrical equipment within containment which is susceptable to ingress of steam or water if the hermetic seals are either initially defective or should become defective as a result of damage or aging.
The Corrmittee believes that appropriate test procedures to confinn continuous long-term seal capability should be developed.
The Committee recommends that further review be made of the battery supplied DC power system to assure that non-essential loads do not interfere with its safety function.
The Committee recommends that further review be made to assure no unacceptable effects such as release of hydrogen into the plant can occur from the failure of a hydrogen charging line. The Corrmittee also recommends that studies be made to assure that failure of an instrument line cannot cause plant control-lability problems of significance to public safety.
The management organization proposed by the Applicants to delineate the safety related responsibilities of the off-site and on-site personnel of the Three Mile Island Station left open questions as to how these responsibilities are to be discharged during normal working hours and during evening, night, and weekend shifts. This matter should be re-solved to the satisfaction of the NRC Staff.
The NRC Staff is still reviewing various issues related to accidents leading to loss of fluid in the steam generator secondary side, such as steam line breaks.
The Committee wishes to be kept informed of the resolution of these issues.
The Committee reconmends that, prior to commercial power operation of Three Mile Island Unit 2, additional means for evaluating the cause and 1 ikely course of various accidents, including those of very low 1651
Honorable Marcus October 22, 1976 probability, should be in hand in order to provide improved bases for timely decisions concerning possible off-site emergency measures.
The Committee wishes to be kept informed.
The Corrmittee believes that the Applicants and the NRC Staff should further review the Three Mile Island Nuclear Station for measures that could significantly reduce the possibility and consequences of sabotage, and that such measures should be implemented where practical.
Other generic problems relating to large water reactors are discussed in the Committee I s report entitled 11Status of Generic Items Relating to Light Water Reactors: Report No. 411, dated April 16, 1976. Those problems relevant to the Three Mile Island Station should be dealt with appropriately by the NRC Staff and the Applicants as solutions are found.
The relevant items are: II - 1, 2, 3, 4, 5, 6, 7, 9, 11; !IA - 1, 4, 5, 6, 7, 8; !IC - 1, 2, 3, 4, 5, 6, 7.
The Advisory Committee on Reactor Safeguards believes that, if due regard is given to the items mentioned above, and subject to satisfactory completion of construction and pre-operational testing, there is reason-able assurance that Three Mile Island Nuclear Station, Unit 2 can be operated at power levels up to 2772 MWt without undue risk to the health and safety of the public.
References Sincerely yours, Dade W. Moeller Chairman
- l. Three Mile Island Nuclear Station, Unit 2 Final Safety Analysis Report (April, 1974) with Amendments 1 through 44.
- 2.
Safety Evaluation Report (NUREG-0107) related to operation of Three Mile Island Nuclear Station, Unit 2, dated September, 1976.
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