ML20024B294

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Preliminary Generic Plant Comparison
ML20024B294
Person / Time
Site: Crane  Constellation icon.png
Issue date: 07/06/1983
From:
ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY
To:
References
TASK-*, TASK-06, TASK-07, TASK-6, TASK-7, TASK-GB GPU-2425, IEB-79-06, IEB-79-6, NUDOCS 8307080291
Download: ML20024B294 (19)


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7 Enclosure (1) provides generic design differences bet >:een a typical C-E Ve are providing operating plant and SW Three Mile Island Unit 2 (T::1).

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this esterial as background, kncwing of your need to res;cnd to regulatory L

inquiries. Since this infor: ation is generic and was gathered fre. services 1

at Windscr we recc:::end that you check the specific validity with respect 1

to your unit.

{'

Part a of Enc 1csure (1) provides a listing of design features of C-E oper-ating plants which w:uld tend to citigste an event similar to the 1. css of Feedwater Event which occurred at TMI.' Such design features in ccnjunction with improved o:erating and emergency procccurcs will give c:nfidence that p.

an incident similar to t*:e TM! event has a very Icw :rotability of cccurrence

. F at a C-E d:sigr.ac c: crating unit.

C.-E will c:ntint.2 to evale:te ptr.inent

E A

design featurcs and pro ptly rec ::.end any. design ecdifications a'nd/or procedural changes which ennance c;erating plant safety.

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p Part b of Enclosure (1) provides systen descriptions uhich form the basis' r

Part c of Enclosure (1) crovides

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for the cenclusiens stated in Part a.

a C-E cngin:ering evaluatien of the sequence'of events which would transpire t.

if an event simiitr tn the TMI 1.:ss cf Feec: ter incident were to cccur on

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a C-E operating plant. Part c of Enclosure (1) is a generic descriptien of r,

expected C-E plant response to avents.similar to these experienced at TMI.

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Co.bustion Er.ginccrir.g studs retdy to assist you in ressending to I.E. Bul-1ctin 79-0G. If you should require such assistance fres C-E. please contact r

me.

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Sincerely.

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V..:.th c l'anagte, Engineerin-) Services t

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PRE 1.ll'.1:a2.Y GE!:IP.!C PLA*:7 CC:7.*P.!S0?[

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The infer:ation presented in this enc 1csure is based on prelictinary i

!j information cencerning the nuclear incident at Three Mile Island f

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The results stated in this enclosure t

t j

Nucicar Power Plant i:ueber 2.

c C-E engineering judgement of events which transpired t

are based on The information and esterial discussed >and durtrig. the accide'nt.

presented by C-E with respect to the Three Mile Island Nuclear Pcwe P

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Plant d: sign has been obtained frem the Preliminary and Final Sa ety E

Analysis reports and personal knewledge. '

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'GEllCRIC DESIGft FEATURE 5 UllICH GIFFER DETifEEN C-E AND Tit!-2 CO:: TENT

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DESIGIIFi'ATURE s

The combination of a low steam generator watcr

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.a) C-E p' ants have a low steam generator water level RPS trigil '

trip and higher steam dump capacity prevent l

Tc-a loss of fced ater, Ol'J plants trip on high

- significant primary system pressure excursions en The Dati trip signal also lifts '-

ri
ary sysica pressure.

the pressurizer relief valves.

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C-E plants.

b).C-E pl.:nts have 40% steam duinp/ turbine 1[ypass capacity.

ei S5:!. plants have 15% steam dump / turbine liypass capacity.

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'j There is more time availabic to estabilsh c.ccrgent C-2 s*.:am generators have a larger secondary stater inventory

  • feedsater af ter loss of main feedester in the C-E th.tn the lill steam generators.

system and less likelihood of lif ting a primary relief valve.

..t Only 2(1-25% of RCS fluid is needed to cover C-E i

. l C-E core cicvation is relatively lower in the tiSSS.

A significant fraction of the Tiil RCS fluid, once core vs 40-45% of RCS fluid for I;ll,

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Izjout thar. lill core.

C-E st:am generator elevation does not form a cold Icg there is void in the RCS, will collect.in the

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t i Icep scal.

steam generator / cold Icg loop seal.

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The higher elevations result in a better potential C-E steam generator elevation.is higher 'than Tiil steam to initiate and maintain natural circulation.

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generator.

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U-tube steam generator design allows reflux corden C-E U-tube steam generator design vs,TH[once through 7'sation of RCS fluid in hot Icg vs only cold Icg return of condensed RCS fluid for OISG desinn.

straignt tbbe (OTSG) design.:

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.1here is more potential for water retention in the 11111 'stca:n generators have score primary..if de water

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1111'SGs uith conscr;uent unavailability of the state f

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Iinventory than the C-E stcan generators.

to cool the core.

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A surge line lool seal allows uater or (teo-phase i

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  • The TTil design has a loop seal in the pressurf rer fluid retention in the pressuri:cr with t.:e prf r The C-E design.has no loop scal.

system volded. This will produce an indication h

sure.e line.

the reactor vessci is not voided based on a false pressurizer 1cvel Indication. The C-E systea uithout the surseline loop seal is Icss p

su-;ceptibic to this falte indicatien.

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draining the pressurizer uhen the hot icg is'voidec l:i The increased clevttien increases the likelihos c' Tt:c C-E pressurizer is at a higher elc0Nions than the 9.

TlIl pressurizer.

10., The C-E P. PSI shutoff head is lower kli$n {he TMI IIPSI The lower shutoff head will not result in liftir?g if the pressurizer relief or safety vlaves if the liPSI ji

,i shutoff head.

, pumps are lef t on.

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Part C

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GEfiERIC FESPC:;5E CF A9 03 ERAT!!;C C-E PLA!IT TO LOSS OF FEED,:ATER GR A Sii.:~r. PRESSURIIER RELIEF /5AFITY VALVE

),

In the event of a. complete 1:ss of feed. tater, the it ediate resp:nse of a C-E fl555 would be a de:resse in s:ea generater level. The Icss of sub::cled E

feedsater flo.e would result in a r.ocerate reducticn in steam generat:r heat rceoval capability and a sica rise in react:r c: lant system (RCS) :ressure and temperature. '!ithin 15 te c0 seccads, a reac::r trip sign:1 uill be produced by lou steam genera:sr le.el, resulting in a resc cr trip and a turbine trip t

(for a plant with a canually initiated auxiliary feet. tater system). -After trip, 4

r'i.he du.p and bypass (if in the n:real autcratic ::de) will regulate steam gener.*

r.tdor and RCS pressure and tcrperature :: hot s: ncty c:nditions and the power operated relief valve would not open.

The stean generater pressure will stabiliae within a couple of minutes; decay f

heat util be re :ved thro.;n :he 6.20 and by; ass. This c:::itien could exist for 10 to 20 minutes tef:r,e the hea: tr:nsfer ca; ability w:uld start te decrease

~

due to an ex:essively Icu steam genera:ce level, thereby resulting in an increase

[:

in RCS temperature and pressure. If auxiliary feed.:::er is initiated within ~

this time perf 0d, the pr.er ; crated relief valve will not open.

y 2

For an event which d:es open stie ph:er operated relief or safety valve and which is followed by the unanticipated failure of a pressuri:er relief or safety valve v

to rescat, the,RCS ulll rapidly capressuri':e..If the relief valve d:es not' l

rescat, the c:wnstres: isola:ica valve c:uld be ranually closed once the RCS pressure drops below :he relief valve closing set;oint. If the valve is not, or cannot be isola cd, the de:ressurizatien uill c:ntinue. In this sitdationi A

the pressuri:er surge tank rupture diskawill fail, releasing fluid to the.

4 containment and the pressuri:cr bubble may be' lost, resulting in liquid relief 2

thr vsh the valve. The release of significan: RCS fluid to the centainment

=s:M.:sult in a cer.: sir.tcr.: is;la:icn signal.

3 "r.5re the RCS pressure appret.ches the saturation pressure of the hot leg, a

'g 5.5::y injecti:n si:.1 uill ec:ur, aut:ratically s:artir.; the high pressure E fety injection (hPSI) p*.: s.

If the ficw of the MPSI pump is sufficient to increase RCS pressure, there is no incentive for the operator to : Urn off these

(

pumps to prevent overpressuri:stien since the MPSI shu:sff head is below the s,b pruer opera:cd relief valve setr: int.'

If only mini-. : s'afe'euseds are availabic 2

h (i.e., one HPSI purp) due to 1 css of offsite p:wcr anc a failure of a diesel or p

HPSI pu:p to start, the HFSI ileu will be ade;uate to keep the core ccoled. If r

the reactor c:ol:nt pur:s are trip;cd, the core will be adequa cly cooled as long

(

as the ECCS (1cu is : in:ained anc the s:ca: concraters c:ntinue :o re=ve decay i.

heat. The layout of the RCS cnhances core c0oling t'ecause the core is essentially,,7 l-at the lo..cs: cic a:ica ei :he sys:c:a, thereby, rc;uiring icss th:n 25" of the i

RCS inventory to Lecp the core covered.

p

  • Tht: sitter:r.: c:s n:t crply to ":ine Y;.nt: which has high shut:ff head pu M.

(

r l

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=<:.g.:::= r: =~ n r.t.: ~':-. - :- - -

~ ~ ~ -

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i, 2

C.t t.u k 1

' :c.;:

Subsequent coold:un of the PCS e.st be regulated within rates specified by existing procccures, by canual ::ntrol cf auxiliary fecc.sater flow, and dump k

i valve ficd. The coole:wn cust n:t de so ra;id as to also cause a

~

depressuri:ation beloit the saturatien pressure of the hot leg.

h Depressuri:ation cust be c:ntrolled within the a;:propriate existir.g pressure-tc:;erature guidelir.cs by talaneir.; cf the ECCS ficw with the stuck pressurizer :

UJ valve fir.r.t and Other fic.is out of the P.C5. as well as the shrinkage of the RCS \\{-

as it cools d:un; again, to assure that the RCS pressure remains above the N

hot leg saturation pressure.

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