ML19322C591
| ML19322C591 | |
| Person / Time | |
|---|---|
| Site: | Crane, Davis Besse  |
| Issue date: | 01/14/1977 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Rowden M NRC COMMISSION (OCM) |
| References | |
| TASK-TF, TASK-TMR NUDOCS 8001170915 | |
| Download: ML19322C591 (3) | |
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A **AM, ADVISORY COMMllTEE oN REAcIOR SAFEGUARDS
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January 14, 1977 IIonorable Marcus A. Rowden Chairman U.S. Nuclear Regulatory Comission Washington, DC 20555
Subject:
REPORT ON DAVIS-BESSB NUCLEAR EO,iER STATION, UNIT 1
Dear Mr. Rowden:
At its 201st meeting, January 6-8, 1977, the Advisory Comittee on Reactor Safeguards completed its review of the application by the Toledo Edison Co:rpany and the Cleveland Electric Illuminating Conpany for a license to operate the Davis-Besse Nuclear Power Station, Unit 1.
Meabers of the Comittee visited the plant on May 18, 1976, and a subcomittee meeting was held in Washington, D.C. on December 21', 1976. During its review, g'
the Comittee had the benefit of discussions with representatives and consultants of the Applicant, the Babcock and Wilcox Coapany, the Dechtel Corporation, and the NRC Staff. The Comittee also had the benefit of the documents listed. The Comittee reported on the application for a constr:uction permit for this unit on August 20, 1970.
The Davis-Besse Nuclear Power Station, Unit 1, is located on the south-western shore of Lake Erie about midway between the cities of Toledo and Sandusky, Ohio. The minimum exclusion distance is 2500 ft. The les j
population zone, with a radius of two miles, included about 870 people in the 1970 census. The nearest population centers are Toleda (1970 popula-tion 383,818) and Sandusky (1970 population 32,674), both about 20 miles from the plant.
The nuclear steam supply system earploys a Babcock and Wilcox pressurized water reactor similar in most respects to those first used in the Oconee Nuclear Station. This system differs from the Oconee units and several other similar units in that the steam generator loops are raised about 30 ft above the level in the original plant arrangemnt. Although this change was made to eliminate the need for internal vent valves, four such valves are provided because of their beneficial effect in reducing steam binding folleging a postulated loss-of-coolant accident.
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Ifonorable !! arcus A. Rowden January 14, 1977 The proposed power level for the unit is 2772 IGt, as compared to 2633 MWt proposed at the construction penait stage. This higher power level is the same as that proposed for the Rancho Seco and Three Mile Island, Unit 2 reactors, both of which have been reviewed by the NRC Staff and the Comittee and found acceptable.
The structures and conponents of Davis-Besse, Unit 1, were designed for a Safe Shutdown Earthquake (SSE) acceleration of 0.15g at the foundation level. Because of changes in the regulatory approach to selection of seis-mic design bases, the Comittee believes that an acceleration of 0.20g would be more appropriate for the SSS acceleration at a site such as this in the Central Stable Region. The Applicant presented the results of preliminary calculations concerning the safety margins of the plant for an SSE acceleration of 0.20g. The Comittee recomends that the NRC Staff review this aspect of the design in detail and assure itself that signi,fi--
cant margins exist in all systems required to accomplish safe shutB59n'of the reactor and continued shutdown heat re:roval, in the event of an SSE at this higher level. The Comittee believes that such an evaluation need not delay the start of operation of Davis-Besse, Unit 1.
The Comittee wishes to be kept inforned.
The performance of the Emergency Core Cooling System (ECCS) has been evaluated using a Babcock and Wilcox evaluation model applicable to the raised-loop configuration. The NRC Staff has reviewed these evaluations and has determined that certain assunptions regarding return to nucleate boiling do not conply strictly with the provisions of Appendix K to 10 CFR Part 50. The NRC Staff is also reviewing several other areas relating to ECCS performance. These matters should be resolved in a manner satisfactory to the NRC Staff.
In conjunction with the evaluation and assessment of the impact of routine waste releases from this plant, the Comittee recommends that the NRC Staff provide leadership in encouraging the developnent of improved environmental radiation surveillance capabilities on the part of the State of Ohio and appropriate local regulatory agencies.
The Conmittee notes that post-accident operation of the plant to maintain safe shutdown conditions may be dependent on instrumentation and electrical equipnent within containment which is susceptible to' ingress of steam or water if the hermetic seals are either initially k
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Honorable Marcus A. Rowden January 14, 1977 i
defective or should become defective as a result of da; rage or aging.
The Committee believes that appropriate test and maintenance procedures should be developed to assure continuous long-term seal capability.
The Committee recoxnends that, prior to commercial power operation of Davis-Besse, Unit 1, additional means for evaluating the cause and likely cource of various accidents, including those of very low probability, should be in hand in order to provide inproved bases for timely decisions concerning possible off-site emergency measures. The Committee wishes i
to be kept informed.
The question of whether the design of this plant must be modified in order to comply with the requirc5nents of WASII-1270, " Technical Report on Anticipated Transients Without Scram (ATWS) for Water-coo 1M Reactors,"
remains an outstanding issue pending the NRC Staff completic
? its review of the Babcock and Wilcox generic analyses of Ar.G.
Committee recommnds that the NRC Staff, the Applicant, and the Babcock ra :3 Wilcox Conpany continue to strive for an early resolution of this matter in a manner acceptable to the NRC Staff. The Committee wishes to be kept
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informd..
Davis-Besse, Unit 1, has installed a bypass loop containing two mnually operated valves around the decay heat removal systan suction line iso-lation valves. The normally closed bypass valves would be opened in the event of a spurious closure of one of the decay heat rer. oval system suction line isolation valves during system operation. The Committee recommends that further attention be given to the means e.Tployed for iso-lation of the low pressure residual heat removal system from the primary system while the latter is pressurized, and that reliable means be developed to assure such isolation. This matter should be resolved in a manner sat-isfactory to the NRC Staff. The Committee wishes to be kept informed.
The Committee supports the NRC Staff program for evaluation of fire pro-tection in accordance with Appendix A to Auxiliary and Power Conversion Systems Branch Technical Position 9.5-1, " Guidelines for Fire Protection for Nuclear Power Plants." The Committee recomands that the NRC Staff give high priority to the completion of both owner and staff evaluations and to recommendations for Davis-Besse, Unit 1, and for other plants nearing com-pletion of construction in order to maxhnize the opportunity for improving fire protection while areas are still accessible and changes are more feasible.
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