ML19220A238
| ML19220A238 | |
| Person / Time | |
|---|---|
| Site: | Perkins  |
| Issue date: | 05/31/1978 |
| From: | NRC OFFICE OF STANDARDS DEVELOPMENT |
| To: | |
| Shared Package | |
| ML19220A237 | List: |
| References | |
| REF-GTECI-A-31, REF-GTECI-DC, TASK-A-31, TASK-OR REGGD-01.139, REGGD-1.139, NUDOCS 7902090396 | |
| Download: ML19220A238 (7) | |
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U.S. NUCLEAR REGULATORY COMMISSION May 1978 (M,.s ) REGL L COPaV GUDE E s%,.... e OFFICE OF STANDARDS DEVELOPMENT REGULATORY GUIDE 1.139 GUIDANCE FOR RESIDUAL HEAT REMOVAL
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USNRC REGULATORY GUIDES
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A.
INTRCD"CTICN General Desic;n Criterion (GCC) 19 " Control Raum," of Appendix A, " General Design Criteria f or Nuc lear Power Plants," to IU CF R Part 50, L icens ing of Prcduction and Ut ilizat ion F ac ili tie s,'
requires that it be possible to ta(e actions fron the control room to maintain the pn.er plant in a safe condition Juring ncrmal oper3 tit ' or in the case of an accident. GCC 34, "Residaal Heat Remo al, ' requi rc that a system to recose r esidual heat t'e p ro s i ded.
GDC 34 d"firec t h, system's safety functi..i as tr.e transfer of f ission product decay heat and other residual heat frem the reactor core after the reactor is shut dw n so that acceptable design limits of the fuel and the reactor coolant pressure boundary are not exceeded. Furthermore, GCC 34 requires that the system safety function c19 be accomplisheJ assuming the avail ability of only cnsite or offsite poaer, coincident with a single failure This guide describes a method acceptable to the NRC staff for complyiry wift tre Co mission's regulations with regird to the removal of decay he3t and sensible heat af ter a reactar shutdown.
B.
D:5CUSSICN The saf e shutdc., of a nucle ar pc er pla"t follcwing an accident not related to a loss-of-ccolant accident (LCCA) h is been typically interpreted as a hot, t a n dtiy.
C0nsem.ently, considerable emphasis has toen pl3ced on the hot standby conc; tion of 1 po.or plant in case k
of an accident or abnormal cccurrence. A similar degro of eg hasis has been placed on long-
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term cooling,.hich is typically achieved L:y the resioull heat remnsal (RhR) system.
The RHR system stirts to c;m ra t o hen the re!rt ur coolant pressure and tegerature are sub-3t3ntially lower than their Nt stind, condition values It is tre intent of this guide to place the s am degree of eghasis en tv entire range of re3cter coolant temperatures and
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p essures, including the range cetw*n hot n4 and R"R ei eration conditions Tr,e impcrtance of reliable ystems that temme dec3y heit f rom the reactor coolant system j
(RCS) while the latter is at or near nere11 eperating ten er itures is indicated by the results of WASH-14M. " Reactor Safety Studj' (R55).
he capabi,ity of a typical P.R p lan t and a typical EnR pIant to r&n. e decaj heat folloning a plant trip was (511uated in the RSS on a pr
- Dilistic b3 sis The e aluaticr ir :lude d both these eserts in which the reactor protection i
.j system (RP5) failed (Anticip3ted Trarsients aitNut Scram) anJ events in which th.; RPS func-g t iened as desig'ed.
For these t, pes of events it was considered acceptable to maintain the reactor at or near normal cE erating te'rgeratur e and pressure f or a long tim
-2 Ho*e.er, in the event of a plant trip even with a successful operation of the RP5, systems
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or equipent failures that led to the inability to r e o se d"c ay heat resulted in a higher prcb.1-j bility of a core melt than that prejitted fcr 13rge LCCA for both I'nRs and BnRs Ccnse-i 1
quently, a significant safety benefit will Le gained by upgrading thase systems anJ eqaipment i
needed to maintain the RCS at the hot-st3ndly Londition f or exterded periods or those needed to
-i cool and depressurize the fd 5 so th at ti> RHR sj stem c ln be operated.
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f urther mr e, even thcugh it may generally be censidereo safe to m.
ain a reactor in a 10t stindby condition fcr a lcnq time, experience shows that theic na$e been events that required enntual cooldown anJ l ong-ter m cooling until the RCS was cold enough to perf orm in<Jection and repairs.
It is therefore obvieus that the ability to transfer heat from the r > actor to the environment after a shutdo.n is an impot? ant safety function for both PWRs and B a'R s.
Consecuently, it is essential that a pc.er plant hne the capability to 40 trum not-standDy to cold shutda.n conditions (when this is determined to be the satest c ourse of iction) under any accident conditions.
These accider.t conditions can conceivably include a safe shutdhn ?arthquake (55f) and an extended loss of offsite power that mz have resulted from that SSE.
In that case, a'l cemponents and equipment that are not seismic Category I and ill systems or parts of systems that depend solely on of f site power sources for their operatirn would be assumed inoperable.
Under these circumstances a plant safe shutdown (inmluding cooldo.n) within a reasonable time requires sy stems designed to safety grade standards and operablo the control room.
"hever, limited operator actions outside the control room may t<
petmitted if suitab'y justitieJ fw processes are nect ;sary to achieve a cold shutdown in a pewer plant: (1) the inser-tion of the control ro in with or without t oration to the cold shutda.n concentr3 tion, (2) heat rejection to the surroundings, (3) depressurization, and (4) long-term cooling.
These crocesses are discussed t elow.
1.
Euration of the RCS to tne required cald shutdown concentration provides an addi-tional reactivity control vasure to ensure that tne reactor will not beceme critica, during ind a r ter t he RCS cool ing.
a.
F or pressurized water reactors (P.Rs), the tioration of the RCS is used in addi-tien te tu insertion of the control rods Bor it ion is achiewd by the chemical and volume nntrol system (CwCS).
It is i m;m r
- it that tnis safety function can t;e achieved in all ic c i de t conditions, including an 5sf ar,J an extended loss of offsite power.
In case of a loss
)f offsite po er, the only means at mixing the injected t;orcn solution with the reactor coolant i s natural convection ci rculat ion.
b.
For bullinq iter reactors (EWRs), t r.e t; oration of the RCS is ac'ieved by the
,tindDv liquid control system (SLCS).
However, that system is activated cnly if the cantrol od> f ail to shut down the reactor.
2 Heat rejection to the surroundings is the only w'ly to avoid a Core melt under normal or accident shutdown conditions.
for Va4s, he lt rejectic achieved by the main steim system and either the s
n o r:r.a l or the ausiliary feedwater system.
In c ri s e of an SSE cn'y seismic Category 1 covonents ond equipment are as sumeJ operable. Daring a loss of ottsite power, the a u x i l i a ry feedwater system provides cooling witer to the steam generators W tt;ut ffs te po.er,
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reactor <coling depends solely on n1tural convection circulation induced Dj tre ctoling effect of heat transfer to the steam generators.
N1tural circulation may le1J to slow an? uxqual cooling. Sinw cooling w:ll result in longer cocidown times, whi(n, in turn, requite 3 1.i r g e ?
rlean feedwater insentory. Unequ31 cualing m y leai to hat spots and high wessel stresses. It is essential th1t ade qJite CColing te provideJ ulder ttese circumstances to keep the integrity of the reactor cool 2 c pressure bound 3ry.ind maint iin..be re actor care in a coal ele form.
b.
For EWRs, heat rejection is achiesed by tho niin steam system and eitrer (1) the normal feed-ater systent in conjunction with tre nain condenser er (2) tr e reactor core isolation cooling (RCIC) system in conjunction with tho cordersate storaa, tink, the residual ne3t removal heat exchangers in tre steam condensirq mo t and the ;. re s sure s t;.p res s ion poo l.
Further heat rejection is achie,-1 by the E"R s,Etem after the RCS his t;e o n sufficiently dep.essurized.
3.
For all tu 'ent desi@s a4 FJ Eass c ourization of the.,
is a prerey Jisite to the casration of the RM s,
tem in the ic y t+rn tou!ing ricde; therefore, it is irrportant that s,., ter s o r c m o nen t w ired to deprossariz, t% RCS tv esigr.ed to withstand d
3ewere postui tied accident ccoditic 3nd be ble to,ortu'" thei r int
- ded tunt ic m.
a.
For t 'a R s. depressurizatin of the RCS cil te achie.ed bj the pressurizer in
- enjuoc tion with cv ar more ut these ccs.:rsrt, (1) the main pre, eiter spray, (2) the luxi liary p r es su-izer spr1j, or (3) the pro rize relief salwes r
st b.
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ossur;zitico of tN RC3 is Khie.ed b depina steim to either tre y
T11in ccrdenser or the pressure su;pte,, en ; col.
5 ten cordeo3] tic, during tho M aste m aperation in th-,te u carders irl w j > will t 91p deg w iur ize t he RC3.
4.
For Loth Pa R s and Eai>
followi, th> reactor dutton a r -:
- oth the initill rd ths intermediate cooldcan pericds Junq-term c )li% is
.e 3 s v, to pro i nt he at vc z ulatic., in v
tre RC Ihis tu ctirn i s u t oc l l j ac ccep l i s* e~ t < tro - G s y s t +-1 In all cu r ent plant vs ip the FH >,, ten h is a l o-or design pres >ure th3n tho n
'n nast of tho e designs it is 1ccited 11ryoly nut s i d*
tt> ca tainw nt.
Hwower, in s;r p l int Josigns, the RH Sfstem is lecited inside the t ant tirment C.
E @ L A T Cii y FO5ITICN l
1.
flACTICN4 The,ystemi etessiry to take the reactor trto norml r; erat im ct odit ions '. o cold i
shutdc.n, includ 7 tre B R system, s h ;1d,
s f y t* e f u"c t ice ll quid ince,re-W ted tvloa a.
The desiG,,hould be Such th 3t the it
!. t u r san ta t-Ik o n f rnn ' O m 11 cporating cGGdi-y I
tiens to cold somtdcan usirq only saf etrgr Me
,ste-c that sati>fy C neral Dosign Criteri1 1 I
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These systems should h<ve suitable redund3ncy in components and features anc suitable i nte rc onnection, I *ak detectiun ard containment, and isolation capabilities to enture that, for ensite electric po.rr system operation (assuming of f site power is not available) and f or of f site electric p3wer system operatien (assuming om i t s power is not available), the sy31 m safety r unction c an be accomplished assuming a 3 ingle f ailure.
In demonstrating that the system can terform its function assuming 3 3 ingle failure, limited operator action cutside the control rcom would be acceptable if suitabiy justified.
c.
The systems should te c ap ible of bringing the reactor to a c( ld shutduwn condition within 36 hcurs fnllowing shutdaw1 with only offsite power or cnsite power available, assuming the most limiting single f ailure.
RHR SYSTEM 15CL AT ICN a
isolation of the suction side of tN PHR system should te provided t-y at least two power oierated valses in series, with v41se positiens indicated in the control room.
Alarms in the contro; room should be prc'.ided to alert tre eperator if either valve is open when tne RCS pras.ure erceeds the RHR syste* desiun p re s <, u r r.
Th-v a l.- e s snauld h.ive irJependent diverse interlocks to prevent the saltes from t eing ope'N unless the RCS pressure is belcw the RMR system design pressure.
failure of a power s1sply snould not cause arv saise to change posi-tian.
Independent diverse protective tre is ce s stoold be presided to close ans open v a i '. e in the event of an increase in the O pressure ahuv> th e IN sy stem des ign p res sure.
b.
One of the following sho;1d te p i n s i i.' > d on th" d i sc h 3 rc,
,10 of the WR system to isolate it frcm the RC3.
(1) The v 119es, po s it ion i Hicators ilums. and ir terlock s described in i:em a.
(2) Cne or more check valves in se'ies with a nortill i closed po er-oporated valve with its position indicated in the control room.
If the EHR syst5m dischargo lire is used for 6
an LCC$ f orction, the power c;er ated v al v e #cuid te aerei upun receipt of a s3fety-injection siqqal nce the re 3c te r c c ol, int pr ess ure has dec r e 3 sed t.e low the ECL3 design pressu re.
(3) T hree cbc;n wil es in,eries or
'4)
Twa check <al,, in (eries prov id M there are sesign provisinns to permit periodic testing of the ch tk val.es fo-leak ti# tness and tt+ testing is terformed at le it t an allj.
3.
R h 8 'av5 TEM FRt MUM RE Lit F a.
To protect tN RHR s/ stem ig tirst accid.nt il ove r p r ess ur il:1t ion when it is in era-s tico (not 1solated from the GCS), pressure relief in the N system should t4 p'ovided with reliev ing cap 3 city in acccrdance wi th the NE Poiler and Pres ure.es sel Cade.
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limiti,g pressure transient during the plant c;arating ccndition when the RHR system is not isolated f rom the RC5 sh0uld be censidered when selecting the pressure reliesing c3p3 city of the RHR > stem f or example, during shutdcwn cooling in a PnR.ith na steam butble in the pressuri-zer, inad,ertent <peration of an additional chtrging pump should t>e considered in selecting the design bases.
Fluid discharged through t.b e R"R system pressure relief values shculd te collected and contained sa tn 3t a relief valve that is stock in the open pasitien will nct:
a.
Result ir floodirq of aw saf ety-related eyaipment.
b.
Reduce tre cap 3bility et tre ECC5 Lelew that reeded to mitig3te the consequences of a pastulated LCCA.
c.
Result in a non-isolatatle situation in which the wtter provided to the RCS to maintain the core in a saf e condition i s <' ichirge d cutside the cantainment.
If interleck> 3ro presided te amtmatically c lose the isolation valses aten the RCS pres-sur. exceeds the
""R design pressure, adeculte relief capacity shauld be pro'.ided during the t ir th it the,al e5 are clos irg.
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- MP PROTECIICN The design rd cperati ~ procedures of im: -Hs system snould inclade provisions tn prewent f ama ;e to tre Ph? system pump d e ta oser% 3 ting, ca.it3tiur or loss of adequite pump suction head 3.
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for tb - c system. th(
cylation,,..e c; cr 3bi l i t, W interloc= circuits shculd f.e 9,1 m f 1 to permit cn-line testira ren ca ratir, in tre hR mMe.
System testing should meet the reqai rement, al IEEE Standard ! B and tre r occ-m 13t cm of Regulato < Guide i.22 Th precpe 3tional and initial st trtup test progr am should be in ccnfo'mance with Regule e
tury Guide i.E-The p ro g r 3 T.
tcr pressurized w 3tt r re3mters shculd "clude tests oitr y p a r t. i n g 3nalysis t<
c:r f i rm ( a) that afeguate minirg of Larated.3ter
- eJ prior to ur d ci r,q 2
taaldcan c3n te Ichieved u ~. d e r natur3i c i rcul 3 t i un conditicrs and par-!t estimaticn of the tim > can te Khieved within tr> limits s;A cit ied in th emeryecy w_riting procedures Cov irison with tne pertorr.ance af resicasly tested plants of s iir i l a r *esign maj w sub-g stituted f er the tests ff
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AutILIAR) FEECWATER SUPPLY The seismic Category I water supply for the auxiliary feedwater system for a PWR should have sufficient i nv en to ry to perm t operation a htt standby conditions for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> i
f ollowed by cooldawn to the conditions perm tting operation of the RHR system. The inventory needed for cooldown should te based on the longest cooldown time needed with either only onsite or only of f site pcwer a.ailable with an assumed sirgle f ailure.
7.
CPERATICNAL PROCEDURES The oper.itional procedures for bringing the plant f rom normal operating power to cold shutdown should be in Conform 3nce with Regulatary Guide 1.33.
For pressurize d w3*er reactors, the operational procedures should irclude specific procedures and information required for cooldcan under natural circulation conditions.
D.
IMPLEMf NTATICN The purpose cf this section is to provida information to applicants regarding the NRC staff's plans for implementing this regulatory guide.
Except in those c3ses in which the applicant prcposes an. acceptable alternative method for complying with the spemified portions of the Ccmmission's requiations, the method descrit,ed herein will be used in the evaluation nf sut mitt ils in conrection with applications for construction pormits for all pl'nts (stand 3rd ind custam), manufacturing licenses, and preliminary design approvals dacketed on or after January 1, 1978. All applications docketed tie f 0re J 3nu d ry l, 1978, will t;e reviewed aQ1 inst this guide on a case"by case basis.
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