ML20040F634

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Forwards Amend to Technical Assistance Request BWR 1-4 Re PSAR Review.Addl Areas Include Fuel Mechanical Design, Thermal & Hydraulic Design & Safety Design Basis
ML20040F634
Person / Time
Site: 05000000, Limerick
Issue date: 06/15/1970
From: Morris P
US ATOMIC ENERGY COMMISSION (AEC)
To: Case G
US ATOMIC ENERGY COMMISSION (AEC)
Shared Package
ML20040F238 List:
References
FOIA-81-385 NUDOCS 8202090442
Download: ML20040F634 (3)
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.. :e. n... .t- ~ - 3:: 1. ( e r. SM .s ~r... n -..t...,., wf! [ p .i ~.,. .) ~. '5 A, - 2 .~ E. G. Case. Director.. Division of Reactor Standards 5, >. a c ... p AMENDMENT TO 2ECHNICAL ASSISTANCE, ' REQUEST 130. BWR 1-%, I.INERICK,, s. + .,a '( GENERATING STATIGI (ICS), DOCKET HOS. 50-35R & 50-353h..,':. ./ ented ia,.,. n,r,- C;. 1, a,,.. ...., ~ -.,...... + s.. :... Trequest to DRS.1 .m... Request No. BWR 1-4, dated -April ~ 22,1970, pr for Limerick'cenarating' 3 ., u ~.. in the review of the r' r'./:.; for te hafeal assistaneaFurther review of ~ the PSAR has hiighli encefenclosure'2?','.~ ' additional'-areas.? ', . Station. ga for which specialized review by DRS is' requested.- g. - has been prepared as an -- "t to Jtequest' Bo!,13WR 1-4. .@ '-" j..., ' ' - U u..t.3.. %,,. '.-*x. %x '.a :.. '... ; '. n..a-9

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s. --.. x' :. ~ i-q. DRS review based upon -questions.in the ciclosure should be schaduled -so '. ? ) s.J..,. as to permit discussica with the applicant during bhe'second technical' j J s meeting planned for June 30,-1970.. Coordinatinn with DRS. staff members y;:- ' 7' (who most likely will have. action in response.to this request) indicates " ' c --3 M?.N

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A=-ahat.Ao Request No.; BIR 1-4,,. h i'.[ .i'S.h. m. .e u... .y s.;.. ...g <. .a.s.c y .y.,.. - ;-; - f ..~- n:- - ~ - s s Distribution: + Docket-(2) o 3.3 g, 3' Q s i " !c," '.. ' s DR Reading - -, 1, DRL Reading BWR-1 File FSchreoder _ TRWilson ' 5- ' RDeYoung ,'4.'.' DSkovholt - '~ - * < .'-,.c - - RSBoyd BGrimes T. Englehardt, OGC t DRL, BWR Br. Chief s s GELear SMKari n I DRL- .f B.W. R-B..WR-1/,DRL .. BWh.) RL .f out > h. 'RS 'pd Ppris y G r:es:1, '~., $URNAR4C> T7.7.91, ;,,,, u% mr, 6/10/70, - 6/[/70[e 6//g/70.. >6/lj/7,D'f. i , '-{ _. u summsunnsomuame-o .i L .Wrm AEC.818 Ouv.9.!,0 s i., < 4' -U D P ne e:- T C202090442 811019 h~~ + h PDR FOIA + PDR 3 ZITZER81-385

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  • ~ ^3 2. Fuel Mechanical Desig@.u.. u'W..e,n.8. ?,-r.SM%... J7-..V.. u.: n ;;,d. q~;..

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m = ~~;~.y *4. ;. v.. : mn. ~. e. M ~....v s.. m. e,.s z....... a s.. s>c o... =p... v. ......z... 2 ~ Determine'.the.'various ' fuel' enrichments (page '3.2-3) and :asceremin the Of' .M 77-: d, l l .?. c - (f ~ possibility?and effect of improper. fuel >rodJ1a=A4ng within'the'fue1cf:. JjfQ f .7S ~ J :.;.9Q, X ' 3. ' ' Y r.~.:;'.-l ~ jM. t.' assembly.-6 E ' ' '.J'&, v' G pu_ clear _ Design: Review all significant difference from previously approved 3.6 ufacilities. Ascertain adequacy of design and intended operation of nuclear fuel to assure: r Capability for shutdown (sufficient negative reactivity available n. for all possible conditions). .~ . ;,.<: c 2. p. ~.. s. .i

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s, . f .. - ?. '.'. ~. b. Suitability of -gado. linitsa..uxide'.as '. bur,n..ab. le.po. ison and.'.r.epine==amt,i.I.1.e.h.. w . :.r/7. ; y. c x... a ....w., a n.v..i.. for-temporary control. cure =,na m..... g ?. g +. t .e..a.,.;- c. g., u n,w.. ;. ;, v ..., g. y _ y.9.- ..o. Validity.of fuel.ru.pture threshold. (425 cal /ga). ~, '. y ' - ~ c. ..., m. v. ..7 d. Margin on limit for peakJfuel enthalpy..to protect -against rup,, ~ ;,.,. (note 280 cal /gm was selected for.1ES while 220 cal /jpa was sel' cted : :; ' ' e for Millstone). -.. s. E - . ^...w:

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'x. 3.7 Thermal and Hydraulic Design: c Review all.significant differences from / y-77 0 - . g f,, I. ' i'.E..,:..e, '_, previously approved facilities.7. .. (,< <. t ,a '3.7.5.2 Determine and review adequacy and approach to analytical methods used to i calculate thermal and hydraulic characteristics of the plant. Ascertain ~ the codes (computer programs) involved in uniti-dimensional. analysis.of distribution, particularly power mismatch between' adjacent assemblies -

pov, with different power and exposure distributions. Determine technique for correlating local (core) instrument readings with the local power generation rates in the fuel. Obtain'and evaluate current program for -

experimental.and analytical investigation of two phm=> flow.. .i.; g.y'.,q*

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,i -- d .v..... ...<.n .~ Evaluate the selection of 2700*F as the mart =um clad temperature permissible y following a ' design basiA'IDCAf (Note that 23000F had been previously selected ' rf

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of metal-water. reaction.-o.n clad at thin temperature *=H== into account R ly[ ._s,..

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D .recent BER-PLECHT data.. y'h y M.L. ; ' s.,,3., ~ >-= s _y';;%{'),-{;.4-. g ' p .;,2y. ~,. 4 n. ... y... l:,.gf ' '~ ' 1.y 6.5. Safety Evaluation ;.,,,f.p _.c :.. :,. 3,7 4,.., 'Q y. ... u,.- _ -a:.. . E.p C -, .~. yd n: ...e-9 4n Verify'and' evaluate experimental data mut/or calculations which support.the }'s+( ~ ~ ' ~l applicant's selection of 2700*F as the mari== fuel clad' temperature allowable.. Evaluate the method essed to determine the amount of fuel-clad 3 perforation. 9@ G.. L .E u. ..r.a 47 'p. N d.4,.k. . ?e..n.. p ..,'s, 2-s- m... - ~., .a p 2 . -... g. s.w s. ,...s 4.. -v ,,,.s .y.- ',.,; r * -.,,. ~ _R.. Q

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