ML19317G511
| ML19317G511 | |
| Person / Time | |
|---|---|
| Site: | Crystal River, 05000303  |
| Issue date: | 03/05/1968 |
| From: | Pawlicki S US ATOMIC ENERGY COMMISSION (AEC) |
| To: | |
| Shared Package | |
| ML19317G507 | List: |
| References | |
| NUDOCS 8003180799 | |
| Download: ML19317G511 (3) | |
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7ef March 5, 1963 CRYSTAL RIVERJ,UCLEAR CENERAT,IN/' PLA'IT_
PEACTOR IYTITIALS ~
The reactor internals and core vill be aesigned to nect normal mechanical,' Sydraulic, ~and therni design loada plus opurat Lunal basis earthquake loads (OSE).* Stress limit criteria used will.be as estab-
'11shc.d in Section III of the ASME Doller and Pressure Vessel Code with
- the exception of fuel rod claddin6 which is not covered by the code.
Seismic stresses will be combined in the most conservative way and will
.be. considered-as primary stresses.
-The reactor internals will be designed to withstand also the concurrent -blowdevn and design basis carthquako (DfsE)** loads, as
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indicated in Supplenent No. 1, Section 9.11.
Primary tensile stresses under such' load combination.will not axceed 2/3 of the stressaa correcpond-inp, to:the uniform strain value at operating tas erature.
Since this criterien results in allowabic stresaes and ctraine ic cr than these corresponding to the 20% cf uniforn strsin criterien racently accepted (Diable Canyon plant). we consider thu strena critoriou preposca :o provide en adequate cargin of safety.
Tuc proposed defor=ntion limita of. reactor internals'are shown on pagea 0.11-10 enrough 9.11-12 of Cupplement No. 1.
The allcw ble defcr:atica limits are about 302 -f the estim ted nelow-of-funettm deformations im 9m v ecific cemencats,
and we t'elinve the e linits provida an adec:cate nargin of Jafety. Ihn applicant has - stated (pa:.'e 9.11-2) that the final calculatua preenure differentini time historica within the 1.cactor vessel. u n11 s tae
. final responau and irrrtial loads, will not be av3.11aole until 2acar.
Preliminary estimates of thcan response and inertial loads due ::o LOCA and carthquake are spected to be available beforu nid-1W and tha applicant has agreed to sub=it these to M L as soon aa practical.
Whers simultaneous occurrence of LOCA and the NE is considered, the applicant intends that both excitations will be input to the system si:multaneously.-- Relative starting times will be changed untiA maxieum structural motions, indicating :naximum stresses, are obtained. The
. dynamic loading resulting from the pressure oscillations caused by a loss-of-coolant accident vill not prevent scra:maing of the ecutrol reds.
'A - 035. - The.analler earth ;uake.
- he 7. nc sill be u%iene :o enticuo 1
normal' operation during_an operational basis earthcaake.
- D3E:- Tha.lar2er earthquaka. The Class I (seinic) items 7111 he
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designed to retain their functional capahill:y during 4 SE.
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o-w Cryatal River [
We have coccluded.that' the reactor interaals desi::n criteria are acceptabla as proposad.
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Section III of the ASME Pressure Vessel Code will be used to design the reactor vessel, pressuri:er, pump casings, and the stcan generators.
' The _ vessel and its internals will be constructed so as to permit recoval of the int tra ls durin; plant _ life, i'
The reactor coolant pipin: vill be fabricated an; analyzed in accordarace vith the requirenents of CS.*iS.I. 331.1 Cede for Pre 3sure Pining.
The applicant stated tnat tetal strcoscs reultia:: frem tnermal axpansion,. pressure, r.cchanical and scis aic lowina,s vill ba cons 1,lared in the design of the re:cter coolant :1;ing.
It.a prcosuri.:er surn iinc connection and the hi;;h pressure injection conn +>ctions vill be er:uipped with charmal sleeves to liaiit s:resses f rats ther=al snock.
The rascter ceolant systen,and all other Cla.in I (seismic) we nnical sysP-. - rill b )
'e ;1 ned to W :tud ac:nal e tim ' ua cf w.m i c d.
3 hydraulic cod taarmal ori< tin plus operation: 1 t anis entaquake (03T.) lo tda vithin nemal cada allevable stresses. In a:!dition..: stars in Suu:la-
- cut 1. Clasc I' ayatens :nd components vill ne f esI-I to withstanc :ce Orcurredt SicWa u' czai daai;n baab.;;.rth::uab.
".A) la.us.
F r L:ary
&':d ranc strtna latensitits under uc.. ! cad cc :1
- 1cus vili acc, encaad
- 2/C :( the.:::casca ecercapon. tin,, to :.:e unif orn.: train value at operating tenparature.
Since thi: critarinn Tc311ts-in tne allavabla strains and, ntrescas le mr than chose corrnsponcing to tne M of tmifer:n strain critcrion reccatly accepted (Di2blo Canyon plant), we consider the stress criterien proposed to provide an adequate u r,,tn of safety.
THERMAL S'10CK ON REACTOR VESSEL We are co.tinuing our general review of the' thermal shock effects en reactor' components, induced by operation d the a:nergency core cooling
' system'(ECCS). - Our review : includes the four major water reactor designers, i.e.. ::abcock & Wilccx, Cosbustion Engineering, General Electric, and '
.Westinghouga.
Tuo nedes of-a-potential f ailure are beina considereC: ductile yielding and brittle fracture. The latter is baiag analyzed by some applicants using both the Pellini-Puzak fracture dia:; ram approach and f racture sechanics.
Cur prelialaary evaluation indicates that_the ductile. yielding analysis is generally realistic, and that cuctile
. yielding is not.likely to cause a vessel f ailure, j
1 con the basis of our preliminary evaluation of the brittle f racture j
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analyses submitted by.several applicants, we conclude that the transition
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tenperatur2 appro.*ch is not quite +1 equate to creat crack propagation pro'alens in plates with sectioa thicLncases ;reater thsu 2 in.
The d etailed. reason.9 for this conclucica are vell nunnarlaud in toe report C23L-;GIC-21, ' Technology of 3 teel Pressura *ieusels fur Water-Coolad
- iacicar Reactors " Section 7.6.4 Specifically, while the transition ca:perature diagran providea a good and conservative guide for a f racture-aaf a design using lov-strength staols of about 1 in. section thickness, it is based on caly greas satizates of stress level and does not consider f actors such as leading speed, crack geometry or the plane strain conditica in tha thick vessel places.
Fct this reason we generally require that a f racture =echanics analysis be perf ormed to chev that a
. cract would not propagate through the entire thicknees of the vessel, when subjected to the thermal si ock stresses.
l The information sub=1tted by the applicant in connection with Crystal River spplication is lacluded in 'Supplacent 1. pago 3.11-1.
It rafara to tbc information submitted in :ennection uith the l'hrea Mile Island application (Dockat No. 50-2S9) accordini to hich a crack could
- .ot >ropagate any furtnar than 35 per cent cf tae vall thiciness.
In addition to this infornation, the applicant stnted in Supple-neat 60. 1 :- f:'e Crystal River n plicatien, c5at the f racture mechanics 1 :19 cur r:.1-'under vey if11 be hn d e, m a:S 1 rutlic.ca b7 2:
7 Prof. P. C. Paria.
Thic nothod censi ts ei a 91ana atrain two-dicensional approach utili:in;, superpe@len of pure tansinn nrd ved e f orce, vich correction factors for the finite thicknesc of tae plate.
The critical stress intennity factor to te used in the anclysta vill be 3C Ksi-in.1/2
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We find this value to be conservative.
The fact that the outer layer of tha vessel shell um21d be subjected to a compressive stress is an additional f actor suppcrting a positicu that m' crack should arrest wicheut propagatin.3 thrcugh the vessel.
In general, however, the information submitted so far is insufficient to conclusively show that the vessel will maintain its integrity when cubjected to the thermal shock stresses and we remain seriously concerned about the potential for vessel.f ailure' through crack propagation.
The thermal shock ~ problem is not unique to thr: Crystal River plant, andive intend to continue the general 'reviev until the lutegrity of tha
. reactor vessel is proven by either analytical.or expurimental data.
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