ML19294C955
| ML19294C955 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 07/18/1980 |
| From: | Plesset M Advisory Committee on Reactor Safeguards |
| To: | Ahearne J NRC COMMISSION (OCM) |
| References | |
| ACRS-SL-0252, ACRS-SL-252, NUDOCS 8009190766 | |
| Download: ML19294C955 (3) | |
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WASHINGTON, D. C. 20355 July 18, 1980 Honorable John F. Ahearne C h ai r.na n U. S. Nuclear Regulatory Commission Wasnington, DC 20555
Subject:
TWU-HUNDRED-FORTY-TiilRD MEETI::G OF TriE ADVISORY COMITTEE 0;4 REACTOR SAFEGUARDS, JJLY 10-12, 1980
Dear Dr. Ahearne:
The Advisory Com.aittee on P,aactor Safeguards held its 243rd meeting in Washington, DC on July 10-12, 1980.
Ite:as considered at this meeting are sum-marized oelow:
The Conmittee cowleted its review of the proposed NRC Reactor Safety Research Program buJyt for Fiscal Year 1982, The Committee developed comments in response to your inquiries regarding reactor coolant pu.np trip requirements and high pressure safety injection termination criteria.
The Co,amittee completed its review of the Sequoyah Nuclear Plant Units 1 and 2 for full power operation.
The Committee responded to Commissionar V. Gilinsky's request for its views regarding the capability of an ice condenser containment to cope with the stea.a from a large-break LOCA, and tne matter of hydrogen control in the Sequoyah Nuclear Plant.
The Committee responded to Commissioner Gilinsky's request regarding the utility of in-core ther.nocouples as an aid in detection of off-normal conditions and for use in accident analysis.
A memorandum was sent to the Acting Executive Director for Operations requesting data on the analysis of a barge collision at full speed with the new intake structure for Sequoyah iluclear Plant Units 1 and 2, including effects of LNG fires and explosions.
Copies of the above reports and letters have been provided to you separately.
The Committee met with maabers of the NRC Staff to discuss recent operating experience, including potential loss of the scram discharge instrument volume at the Brunswick Steam Electric Plant and the Edwin I. Hatch Nuclear 8009190 g
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Honoracle John F. Ahearne July 18,1930 Plant, failure of control rods to scram fully at tne Browns Ferry Nuclear Power Station, response of the primary coolant system to cooldown on natu-ral circulation at the St. Lucie Plant, and the venting of contain:aent to release Krypton-85 at the Three Mile Island Nuclear Station.
The Co;comittee met with you and the other Commissioners to discuss the manner in which the Coramission plans to integrate the current activities regarding consideration of degraded cores in the licensing process, NEPA consideration of Class-9 Accidents, proposed changes in NRC Siting Criteria, and Emergency Planning.
NRC policies regarding the standardization of nuclear facilities, and the role and responsibilities of NRC resident inspectors were also discussed.
Since the last report of ACRS activities, the following subcommittee meetings have been held:
Advanced Reactors, in Washington, DC on June 13, 1980 General Electric Test Reactor, in Sunol, CA on June 16-17, 1983 Me'.al Components, in Washington, DC on June 17-18, 1980 Safeguards and Security, in Washington, DC on June 26, 1980 Waste Management and Fuel Cycle (joint meeting), in Washington, DC on June 20-27, 198J Advanced Reactors, in Los Angeles, CA on June 30, 1980 Reliability and Probabilistic Analysis, in Los Angeles, CA on July 1, 1980 Consideration of Class-9 Accidents, in Los Angeles, CA on July 2,1980 Reactor Safety Research, in Washington, DC on July 8,1983 Advanced Reactors, in Washington, DC on July 9,1980 Fire Protection, in Washington, DC on July 9,1980 Sequoyah Nuclear Plant, in Washington, DC on July 9,1980 Tne following items are tentatively scheduled for consideration during the 244th ACRS Meeting, August 7-9, 1980:
Three Mile Island iluclear Station Unit 1 -- restart and full power operation North Anna Power Station Unit 2 -- discussions of augmented low-power test program results, shift manning, and plant risk assessment studies
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- Honorable John F. Anearne July 18,1980 Quantitative Pisk Criteria -- development of criteria applicable to the nuclear regulatory process Degraded Core Considerations -- application in the licensing process Regulatory Guice 1.97, Instrumentatica for Light-Water-Cooled Nuclear Power Plants to Assess Plant Environs Conditions Durino and Following an Accident Fort Calhoun Nuclear Station -- pouar increase Si ncerely yours, Milton S. Plesset Chai r na n o