ML20024B294
| ML20024B294 | |
| Person / Time | |
|---|---|
| Site: | Crane |
| 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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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 '-
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- 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.:
- U-tube steam generator also allows more uater invc tory.
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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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- 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.
(
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i, 2
C.t t.u k 1
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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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