ML20037A664

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Forwards Request for Addl Info Re Consequences of Steam Line Break Accident
ML20037A664
Person / Time
Site: Crystal River 
Issue date: 07/17/1972
From: Knuth D
US ATOMIC ENERGY COMMISSION (AEC)
To: Deyoung R
US ATOMIC ENERGY COMMISSION (AEC)
References
NUDOCS 8003250680
Download: ML20037A664 (5)
v • d • e


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..g-:;.. 2.- ;24;.:*..,.e o ;. s.. ; K.. ~ ~. 9,X... BLuestorate et Liaomeing .a. ,.n s.., w. +, ; w. c; 4 F1421D4 POIER CORP 0t& TION.' CRTSTAL ME Wrma essaatta pLart. '.^ W t DocET No. 50 302; INSTEDWrfATIGE ABD RIACTRIC POWER SISTERS QUESTIONS J' + s, y# ,s a. r ...2 ,,.m. -i The enclosed' list of questions was' prepared by the LIE 14CS Branch for? transmittal to the applieent.- The additional information roquested in those guer.tions is accassary for our assessment of the consequences-of a steen line break' accident..'. .'5 + 7 g - 1 y L-s. ~ V._ 0dgina1 Signed by ~ 'Dondd F. Knuth ("'.', H3-99 Domaid P. Knuth, Assistant Director L;III.CSr03 nfor Reactor Safety - Directorate of - 1,1 censing

Enclosure:

Quastions Ptsut usr.s Crystal Rivc: 3 Licencias stage: Operattag License - m u Dockat Mu=her. 30402 Rasp:ssible. Branch & Proj:ct fr= M rs Pmth, M< Faulkner - Applicant's Raspeesa:3 ate Necessary for Completion 1 ' -ef Ndat Action Planned ce Projects S/3G a Dascr4 ttom of Esspenses Questicus'.. .,;.. %'.Q tse States,e/d Austadag'.Appmame'.s.,%sponse ', ','A #

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.. +* . ~- -,...._"e k ar .. a, ~ w,,,.. ~.s... p... x.. "*:..,,..... .w .. ~.,~.,.e - ~.. w. --*.u- . c=< -v ' ~ ~, n s .....D.'..- NT~#' .,.'. e' '.I s ' -_ . ' ~ - b. ." ; ~ ,. aides of -the stema generator;. V .... - w.n ^>r

(d) thq ste
s generator bladesm mdel, faciuding the break flow model, L/D' racio, discharge coefficiast, emic goality or _

~ a s,. superbent assumed; ~ (a) assumpthr.s on &chater flow histcry during the accident;. (f) the' prassariser level model..

n. :, -

3 Describe the aantron kiactics modal usedi including equati me, k, resetivity fccdback rechanisas, and appropriate impet data,. e.g. eeutrun lifeti=e, dalcysd precursor cenecatratica, nodorst:r and fuel temperaturo ceafficients, boren concentratica Acd verth, cad 3 ~ tripped red uceth. ~ '~ 4 Describa how the recults cf cha thermi-hydraulics and neucranics calculaticcs are v.ced to calculate the D5 r: tie. 5. Provide ar.alyses for the doubla-ended rupten of all stera lines assucing the reactor is initially eyerating at hot stand-by (scro-pcmer), 601 pcwer$ amd '100% ywer condi: tons; assa=a me red is stuck and the cripped rod worth is equal tu 1". phs the pwar defect D ag neck ecaditica. .+ ' y, ~ (. cn ^ ^ .~ ?;, m ' :g: s; n~5'. % b (Nas results,shoub!'fac1kade $s'A [ oft D g g- ..~..?,

a.

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m,a,;p x

$y9~f, (e}' avezaga, seactes coolant ' temperature k,,, ,7 1.., '. m" i ~' ~ - ... ~. m. .s. a. g.;. s.r .. v. 2,; ,. %... e,,. cyk(d)(.entktempsetenseandpeessureslefthsjteamgensretors - 1: 3%. L m ' _~ 7 iz :L _ wwLW. y Q* y a.:: s %r.': + W. 3 : ,. L- ~. ~ u-u u = m n - nna .4 * :, -m. . m... _ m,. L .y=. #3*'-

,.n. g. 9. m...,..

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..n p. n 1 with tha'esclast/ tha Juel; the control rods and the total; ..Y.

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.+ .1 n 7 .....m- - '.- (f) the ENR rette and the appropetats aoswelation used. ~ 4 %, m u .s ..~., e , p, =_ ~ ~.., 6, table 3-4 of the F5&E h4mes that a ama-Jinnasiceal, isothermal ~

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= . emiculaticsiof tha moderator tegerature coefficient is perfyrmed uhink could be man'-censeretive for the staan Uma break accident. Astify the uma of this coefficient and' ascertain the assimus 7 negative coefilcicat possible considering spatial temperature and j f~@d effects. 9 e (N<o o Ov

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