ML19221A836
| ML19221A836 | |
| Person / Time | |
|---|---|
| Issue date: | 11/02/1970 |
| From: | NRC OFFICE OF STANDARDS DEVELOPMENT |
| To: | |
| References | |
| REGGD-01.002, REGGD-1.002, NUDOCS 7907100050 | |
| Download: ML19221A836 (2) | |
Text
ild To i Reprmteii 12/1,70)
SAFETY GUIDE 2 THERMAL SHOCK TO REACTOR PRESSURE VESSELS A. Introduction gram. Since reactor vessel materials are ini-Proposed General Design Criterion 35 speci_
tially ductile and their fracture toughness prop-ties de<ign and oparating conditions necessary erties are not significantly changed upon irra-diation durmg the initic f> years of operation, to assure that the reactor coolant pressure boundary will behave in a nonbrittle manner.
the potential for reactor pressure vessel fail-ure as a result of co!d water injection is con-To provide piotection against loss of coolant sidered to be acceptably small during this
.a ridents, present designs provide for the in-jet tion of large quantities of cold S.a gency period. Sutlicient data should be available from t oolan t into the t eactor coolant system. The the HSST Program to permit a final judgment within this 4 e.ir period on the acceptabihty effect on the reactor pressure vessel of this cold water in jecto n iru.ern because the reac-
"f the pr nes ted behavior of vessel mater: 0 to; sessel is sub lecte J to greater irradiation throughout it, 3ers ve lifetime.
In the esent that the resu:ts of the HSST than othei compma ats of the reactor coolant piessoie boundary and. thus, has a greater po.
Program m u hri rewarch mdicate that the tential for becoming brittle. A suitable program potential for grou th of defet ts in rad:atio.
brittled reactor prewuie s ere: material re-which may be used to implement General Design d m es the availah'e margm of safety agamst Ciiter ion 35 to assure that the reactor pressure vessel will behave in a nonbrittle manner under brittle fracture to an unacceptable level, an loss of coolant accident conditions is described aweptabk enginming mlution to the problem
- ould be apphed-for example, thermal anneal-in this guide.
ing of the reactor vessel material. Naval Re-II. Discussion search I.aboratory data indicate that annealing of a PWR ves,el at its design temperature The injection of cold water by the emergency (650cF) for a perio4.of 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> should pro-core cooling system into a hot reactor pressure duce a r-covery m fracture toughness proper-vessel after a loss of coolant accident raises the ties and reduce the transition temperature shift possibility that a vessel embrittled by irradia-due to irradiation by 30 to a.0 percent (i.e., a tion and having a small internal defect could 100"F >hift in transition temperature would be fail suddenly as a result of the large thermal reduced to 70 after anneahng). Annealing gradient imposed and the resulting high BWR vescels at design temparatures and for stresses. Analyses by the reactor vendors indi-equivalent time periods would, if needed, pri,-
cate that cold water injected into a hot reactor vide an equivalent degree of recovery. Based on pressure vessel toward the end of the vessel's the calculadon of potential irradiation eff ects in service life could cause incipient defects of the presentiv designed P%.Rs and B%.Rs. this de-maximum size expected to grow; however, the gree of recosery of material toughness proper-maximum crack depth is predicted to be no ties combined with the potential for repeating more than 30 to 60 percent of vessel wall thick-the annealing rro(ess if required, appears to ness The vessel is not expected to fail under be adequate to permit contmued plant operation these conditions. The maximum crack depth with the same reactor pressuie sessel through-expected cannot be firmly established since the
" "I D'" "l Iif"I' * *'
ses.sel material fracture toughness properties assumed in the analyses have not yet been com-C Regulatory Pmition pletely confirrred.
To assute that the reactoi pressure vessel The additional data needed to tesolve the will behave m a nonbr;ttle manner under loss u ncei t:o nties m the frat ture toughness prap-of coolant conditions. the following program erties of t eactor vessel mat'etial are e,pected 3hould be follow ed.
@N-to be provided by the Heavy Section Steel Tech-1.
Data collocuon and research w ork on nology ( HSST) research and development pro-the propent s
ica ressure ves-2.1 790710006C
,ci mater ial *tiould be t untinue<1 in or-ently approved core r reacto; pres-der to permit verincation that expected sure vessel designs ce proposed.
m.ttenal properties assure nonbrittle 3 Should it tu conclude.1 that the margin i.eh.a mr of the reactor vessel through of safety against rractor pressure ves-out its hfetime under postulated acci-sel brittle failure due to emergency dent conditiona. It is expected that this core cooling system operation at any determination can be made within 6 time during vessel life is une ceptable, an engineering solution, such as an-i suring the 5-year period necessary to nea!ing. could be applied to assure ade-deselop the needed data, the potential quate recovery of the fracture tough-reactor pressure vessel thermal shock ness properties of the vessel material.
problem which may result from emer-In the meantime, applicants shoulo out-gency core cooling system operation line available engineering solutions and ned not be reviewed in individual show that their designs do not preclude cases unless significant changes in pres-the uw of such solutions.
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