ML19256F538
| ML19256F538 | |
| Person / Time | |
|---|---|
| Issue date: | 06/19/1979 |
| From: | Fraley R Advisory Committee on Reactor Safeguards |
| To: | Gossick L NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| Shared Package | |
| ML19256F505 | List: |
| References | |
| ACRS-1648, NUDOCS 7912190362 | |
| Download: ML19256F538 (2) | |
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APPENDIX XXXIV Requests, A reements and Commitments 9
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION g
ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 8
WASHINGTON D. C. 20555 o
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June 19.1979 MDiORANDUM TO:
Lee V. Gossick Executive Director for Operations ERCH:
R. F. Fraley, Executive Director, ACRS
SUBJECT:
REQUESTS, MREEMENTS AND COMMI'IMDTPS DURI1G 230TH ACRS MEETING - JUNE 14-16, 1979 During the 230th AGS Meeting the Ccanittee reached the following deci-sions and conclusions regarding several matters related to AGS activities as noted below:
Millstone Nuclear Power Station Unit 2 - Power Level Increase The Ccanittee concluded that it would not object to the NRC Staff plan to license the Millstone Station Unit 2 to operate at a power of 2700 MWt.
Sequoyah Nuclear Plant - Oceratino License h e ACRS deferred completion of its review of the OL application for Sequoyah until the NRC Staff has completed its review of the implications of the TMI-2 accident as they relate to the Sequoyah application and has reported its findings and. intentions to the ACRS in a specific or generic manner.
In order to avoid unnecessary and/or unproductive administrative delays in the consideration of this application, the ACRS is willing to consider li-censing actions to permit fuel loadirg and low-power testing at this fa-cility to an extent that will not preclude those modifications to the plant that may be considered necessary by the NRC or the ACRS.
Palo Verde Nuclear Generating Station Units 4 and 5 - Construction Permit he ACRS deferred completion of its review of the CP application for Palo verde Units 4 and 5 until the NRC Staff has completed its review of the implications of the TMI-2 accident as they may affect the design of these plants and has reported its findings and its intentions to the ACRS.
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. Lee V. Gossick Nuclear plants on which the ACRS has already reported on OL applications but the OL has not yet been issued
%e Comittee expects that the NRC Staff will keep it informed regarding the additional requirements arising from consideration of the mI-2 acci-dent to be imposed on those plants for which the ACRS has already reported to the Comission on OL applications but for which the OL has not yet been issued. me information supplied to the ACRS may be provided either case-by-case or in generic form, as appropriate. te ACRS expects to review this information and any actions proposed by the Staff and to report thereon to the Comission as considered appropriate by the Committee or as requested by the comission.
Return to coeration of Three Mile Island Nuclear Station Unit 1 2e Comission expects that the NRC Staff will keep it informed regardity the applicant's proposals and the Staff's intentions regarding the condi-tions that must be satiefied prior to the return to operation of M I Unit 1.
The ACRS expects to review this matter with special attention to the in-It is teractions between Unit 1 and the recovery operation at Unit 2.
our understanding that this matter will be ready for Committee considera-tion during the August (232nd) meeting.
Bailly Nuclear Power Plant - Modified Piling Desian In response to a request f em Dr. Hendrie dated June 8,1979 (see attached) the Comittee plans to consider this matter during its July (231st) meeting.
Zimar Nuclear Station - Accarent false statement made durino ACRS Subcom-mittee meetina te Comittee discussed the matter identified in the memo from James G.
10, 1979 regard-Keppler, Director, RIII to Dudley Thompson, I&E dated April ing apparent erroneous itformation provided to the ACRS during its review Se of the operating license application for the Zimer Nuclear Station.
Comittee requested that it be informed regarding the resolution of this matter.
R. F. Fraley Executive Dir tor
Attachment:
Ltr. frm. J. M. Hendrie, Chairman, NRC, to M. W. Carbon, Chairman, AGS dtd. 6/8/79
/ b S Chilk, SECY R. Mattson, NRR R. Baer, NRR V. Stello, I&E 4
H. Denton, NRR D. Vassallo, NRC A eso
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o UNITED STATES NUCLEAR REGULATORY COMMISSION
{g j e WASHINGTON, D. C. 20066 Y
June 8, 1979 CHAIREAN 1
DISTRIBUTEDTO ACRS MEMBERS Dr. Max Carbon, Chainnan Advisory Comittee on Reactor Safeguards U.S. Nuclear Regulatory Comission Washington, D.C. 20555
Dear Dr. Carbon:
Y The Comission currently has before it petitions relating to the proposal of the Northern Indiana Public Service Company to use shorter pilings than originally contemplated for foundations of the Bailly Generating Station, Nuclear-1. The licensee proposes to drive the pilings into glacial lacustrine deposits underlying the site, rather than into bedrock or glacial till.
The Comission requests the Comittee to identify and address the signifi-cance (if any) of the engineering and safety issues arising from use of the shorter pilings as opposed to the longer pilings.
In particular:
(1) is the use of shorter pilings a significant design change from the standpoint of engineering, and would it require significant alteration of other aspects of the design of the facility; (2) what difference, if any, would there be in the safety of the facility depending on whether longer or shorter pilings are used?
The Comission would appreciate receiving the Comittee's views no later than June 30, 1979. The Comission, after receiving the Comittee's views in writing, may request an oral briefing should that appear desirable.
The Office of General Counsel will provide background information on this matter upon request.
Sincerely, Isep M. Hendrie
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APPENDIX XXXVII Sumary Comparison of Stainless Steel nd c 10 Fuel Rod cladding o
UNITED STATES
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j-I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON D. C. 20555 0,
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June 18, 1979 Honorable Victor Gilinsky Comissioner U. S. Nuclear Regulatory Ccemission Washington, DC 20555
Subject:
SUMMARY
COMPARISCN OF STAINIESS STEEL AND ZIRCALOY EUEL ROD CIADDIIG
Dear Dr. Gilinsky:
h is is in response to your mecerandum of May 8, 1979 concerning the relative merits of stainless steel and zircaloy fuel rod cladding. W e following dis-cusses cladding performance of these two materials under both nonnal operating and accident conditions.
Early claddire developnent programs (1950's to mid-1960's) studied both stain-less steel and zirealoy cladding extensively. Both materials have been used in power reactors.
A consensus was eventually reached that zircaloy is the more desirable of the two materials, and by the mid-1960's, zircaloy was predcminant.
Safety aspects of both types of clad material were reviewed by the ACRS. Today, only two U.S. comercial nuclear power plants (Conn. Yankee - PWR, and Lacrosse -
BWR) use stainless steel clad fuel rods.
Mere are primarily three factors which enter the comparison of the two materials for normal operation:
1.
Stainless steel cladding is susceptible to stress corrosion cracking in a BWR environment during normal operation whereas zircaloy is not. Con-sequently, stainless steel clad rods experienced much higher defect levels in BWRs. For PWRs, stress corrosion cracking of this nature has not been a problee 2.
W ere is a large neutron economy advantage in the use of zircaloy as op-posed to stainless steel because zircaloy has a much lower neutron capture croes-section. 2 1s translates into lower uranium ore and enrichment requirements.
3.
Of the tritium generated in the fuel by fission, much more is chemically trapped by zircaloy than by stainless steel. Stts, use of zircaloy sig-nificantly reduces environmental releases of trititm and problems associated with this nuclide at nuclear power plants.
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t Honorable Victor Gilinsky June 18, 1979 In regard to accident conditions, as indicated in your memorandum, stainless steel has an advantage over zircaloy in releasing only approximately 16% as much heat on being oxidized by steam. However, during this reaction the two alloys will generate comparable amounts of hydrogen. Other proporties that are important in accident sequences are rate of oxidation and melting temper-aturg. The rate of oxidation is reasonably low fgr both materials up to about 1650 F.
For temperatures between 1650 F and 2000 F, stainless steel has a some-what lower rate. In the tgmperature rarge of most importance for accident con-ditions, greater than 2000 F, zircaloy has a slower rata of oxidation although this aspect is complicated bg the higheg heat of reaction. Also, zircaloy has a higher melting point (3360 F vs. 2550 F) than stainless steel.
For an accident such as occurred at TMI-2, there is no reason to believe stain-less steel would have had any performance advantage over zircaloy.
Sincerely,
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Max W. Carbon Chairman ec: Clairman Hendrie Comissioner Kennedy Comissioner Bradford Comissioner Ahearne Office of the Secretary ACRS Members i
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!*r*CRANDCM TO:
Chairman Hendrie Commissioner Kennedy Commissioner Bradford Coc=issioner Ahearne v4 FROM:
Commissioner Gilinsky h
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APPENDIX XXXVIII e sfons of
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e9 lato y Guides UNITED STATES NUCLEAR REGULATORY COMMISSION o
5 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS
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- WASHWGToN, D. C. 20555 June 19, 1979 Mr. Lee V. Gossick Executive Director for Operations U.S. Nuclear Regulatory Comission Washington, DC 20555 SUBJECr: ACRS ACTIN CN PROPOSED REVISIWS OF RKiUIATORY GUIDES
Dear Mr. Gossick:
During its 230th meeting, June 14-16, 1979, the ACRS concurred in the regulatory position of Regulatory Guide 1.9, Revision 2, " Selection, Design, and Qualification of Diesel-Generator Units Used As Standby (Onsite) Electric Power Systems at Nuclear Power Plants."
Sincerely, Max W. Carbon C1 airman ec:
H. Denton, NRR R. Minogue, OSD G. Arlotto, OSD S. J. Chilk, SECY 1624 438 qqo8D @q&
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