ZJl-120 1972
J.HMWM
ACCIDENT AND SAFETY ANALYSIS OF THE KS-150 REACTOR
SKODA WORKS Niriiir Powtr CoMtrvdiofl DtpartMtnt, Information Contra . PLŽEM - CZECHOSLOVAKIA We regret that some of the pages in the microfiche copy of this report may not be up to the proper legibility standards, even though the best possible copy was used for preparing the master fiche. £ 1ÍQ
"far. liniové.:-
: С I D E II T л i: D ti A P В T Y ANALYSIS OF TUK К G - i. 5 O REACTOR
O!. ,л.и V/ORKS
.fiai' Power CŮHÍJÍ/Í .ir.txori Гера^аито, Information Centre Notation 3 1. Introduction: 4 Й» Tfe* state-of-art and the development trends in the safety of power reactors 8 2 л feclear reactor as a source of potential danger 9 2.2 A brief review of nuclear reactor safety 11 •'o? Development of safety philosophy 14 Z Л Зй?е<^у devices of nuclear reactors 18 3e Thff A~l nuclear power statioa 24 ЗЛ The main data sheet 24 3.2 Power statiac maia equipment and its thermal schematic chart 26 3e3 The control system 39 3 4 The system of safety 43 4. The method of the XS~150 reactor safety analysis 50 4»! Comparison of the KS=>150 reactor with other gas cooled reactors» Requirements laid on realtor safety 50
4-B2 The method of analysis 54 4.3 Extreme initiating events 57 i Л Mathematical models 58 5. The KS-15G reactor basic dynamical characteristics 61 5.1 The effect of intrinsic feedbacks 62 5»2 The effect of the control system 63 5 Л Summary 65 6. Analyeia of the KS-15C reactor accident events 75 6»! Uncontrolled withdrawal of control rods out of the reactor core 75 6ЛЛ Summary 70 6 Л Breakdown af cr* Q,C two circulating blowers 8 6*2.1 Breakdown of ыче circulating blower 88 *> ~ .. .«.U-JWU ui' two circulating •-. ...«..•were 90 -,. ... F-v* lu&?, ;,-.?•& of v^&xiltá ' 92
-.-í-- ;::'. ?Í;'.."• -.'.,-.?•iíii/Vvcc. sapply of the power
. :.:' ::.*•'..-:•• L.^r~3-&r>.pt-„c.i:-- supply followed by -.-. -t-jv -v л-:к>л\? rv ".h«t srřety system 114. • - " --.:>: г. .irptícr supply with subeequeat .,.;•-.• .'^'':.> ď*.fr.'"..y system 116 , .. • '•,;ш*:;,- 118 - •;'\_/^:у г-' ' .* .-, ; tw*3v* saic. pipings of the A-l •'* •- si? ,-Í рг^.жагу ^irsuit on the nominal pi^vsr l*s"s-*:*. 129 -olo:l P
ÁÍ ABsmc?
15515 first part tf the report reviews briefly the sta~ te-sf--*rt and the development trends of safety of power r«ar-tc*»s in the world. Ťh9 main part deals with the КБ-150 reactor cafety analysis. A method of reactor safety evaluation based c» tii* pretf^aťe. af the extreme initiating events is propoeed and &jjplii?a1?ede different accident state» are analyzed in detail,ftfil th e probability of their occurence daring reac- trr life-time л
i Jj m 0Tr C' C~ ) *»...,.,*»*.—* tr,f. tc,T£lp4|T.at.«xr-A eo«fi.i--i.«iit cí .-.r4'íJ.U".-í.!. reSi •.t.ivTt^ 'i'0« Г У~> .е...... t>.* teaipsr&^ii-í»
'-ty (X) l O ,*,,»,,, *x-".'všiví-. ,r tne fuel psitS та- * • xamam i-amperature related to lt-S i niv'i.«.д. V3:. Ue •^*m (£) l°C) .*...»»-.»*. mignii.'*•!# of txe temperature signál cn the thermocouple Ж.Т.) í.J*s." ) **,*. w с с.,., . r*a tor pcwer GCC) CkgaSo^"1) »,»»„,..,., j^t* ^.f я-е flow through the react:г GTj('0 Ckg»s»"" .) •»...... « quantity of i?;as delivered by th* Mower J£ "С, Сз) ...... ,,.«...,». ''•'зв .interval of ibaxiinum taming ** t%f t,V'.v?.-' veg'-'et-.rAg vanes Л С5 Is,' •• ..•*,..,. ťlau» internal during which th* slide vaiva an the blower bypass i a opened ^(grad) •<''i:viti*.l. in она в а О and final angle f\grad) „,.,... • r.-f tu.,;ni ug if Ыь blower vanes 3 Ú) /m/ • * e efxe t.i'.-« position of the con trol rod ,1.u .realtor core Barred symbols (for i net an -.e
0th*p symbols ar* d-efju^d :iзл čutali an fcVi t*itt. л, INTRODUCTION
.bcťsty of nuclear devices is inseparably associated •"-f ~;\e civilian development of nuclear energy ut.iI:Lza- ",,:-. из is tie time of its birth and development '?»!"« :.ii trir. a very great fear of atomic nucleus destructive :';_•'Jí, the nuclear po'ver engineering has been forced by :.'• >.,:,.;. opini-m to accept such a "ilg.. '.ie^r^e of safety •••.'.ix.'.:"i r..as no parallel so far. It can be se^r^ in terns • í i/'i^ie":. oal data gathered from several uundreds of research»
Vvivintst and power reactcrs tha"" the results reached ar« ^xtraordiparily good. In the course of reactor operation representing some 3 000 operating years, only 7 cases of cl*ath di« to radiation have been recorded. Hone of luortei cares «aid above has occured on a nuclear powei station. ?řas sue зев* <Ав far more important if seen together with "'•£:« fast that nuclear power stations, in spite of that l-*~ ><ýur an:1 money consuaing devices, have become an econonu- «ally asctptabie source of power. It she aid be added that tin psse Utilities of their further development are far from being exhausted. 'PowardLs the end of 1967 there have been as much as 67
auclear power stations in the worldf representing an ove~ Tfll power exceeding 10 000 Me, Another 42 nuclear pov;er gt/afci-глз with planned operation by the end of 1970» are un~ d«r construction. Taken together, there are 109 nuclear po- w»r stations to be in operation in 1970, Their cumulative -p-wer is noře than 52 000 Ив /1/» Proportions of reactor V'j^s will be clear frosa the following ca'fclei
Ц- Table
Number of reactors ii. Type of power reactor 1967 1970
Gas • graphite 22 26 В íling water reactor 16 35 Prsssurized water reactor 15 26 Heavy water reaecor 7 12 Past realtor 3 5 Others 4 5
Total 67 109
A great majority of these nuclear power statione are s.lťiated in remote areas far from densely populated centres. In addition, the nuclear power stations proper are equipped with intricate and expensive safety devices which, along with special operational measures, are to pre vent the sta tion and its surroundings from dangerous consequences ir. ca se 'f an unexpected serious failure of tne reactor. Concer ning cost it may be said that in the present designs of nuclear powar stations of high importance is the proportion : f money consumed due to safety aspects. ÍPhis situation may Ь9 explainable by the fact that safety is based oi the ana lysis of the "maximum" theoretical accident of túe reactor, :; a:-oo;int being taken of the probability of the origin of initiating failure and its consequences.
As far as perspectives concerns it ought to be stated
• 5 - that an extensive building of nuclear power stations will bt s-ibrject to the possibility of locating them independent ly of the density od population. It is also necessary to no- lit the evident fact that the results of nuclear safety геье- «:-?ih along with the favourable results experienced in the •yxvs* of operation of reactors are in sharp contradiction with the irrational demands as to radiation safety of nu clear power stations^ This is also the reason why in the ;ast two or three years quite new, progressive views based • a uant.itative analysis of causes of reactor failures and their influence en living conditions in the surrounding uveas appear. These are developed with the aim of ena'.iar^ ••«a.-itcr design with a hazard given in advance which is ac ceptable from both the public and economical point of view» The report deals with the analysis of certain selected a< !;ident situations on the Czechoslovak A-l nuclear power station* The KS-150 reactor of this power station is a heavy water one, cooled by compressed carbone dioxide. Na tural metallic uranium is used as a fuel. The reactor is encased in a steel pressure vessele This type of heavy wa ter reactor is neitner mastered лог its detailed design is worked out in the whole world. As for the most part of A-l nuclear power station components the operationally veri fied characteristic of reliability do not exist, the eva luation of safety cf the station as a unit based on the probability - quantity approach is unacceptable. Therefore the author has followed an essentially qualitative method of evaluation through analyzing consequences of possiblef aven if very improbable extreme failures of power station •siqaipmentj and looking far the ways to limit their conse quences i. Judging from the data available it may be said that the probabilities of occurence of said extreme acci dents are also evaluated, in terms of order at least»
6 - The report consists of its fundamental part and the supplements. The former contains the method and results of solution, the five independent supplements present the mat hematical formulation of the problem, some experimental da ta, and computer programs» The author thanks to lir. Z. Kobeda, Mr. 0. Valter, and Mr. R. Kalivoda /all from Nuclear Power Plants Division, SKODA, Plzeň/ who contributed verv much in the setting-up of the oalculrting programs presented in supplements No. 5, 4, and 5.
/ <** 2.. THE STATE-OF-ARI AND THE DEVELOPMENT TRENDS 1И
THE SAFETY OF POWER REACTORS
Tha litankind has bean made "acquainted" with the nucie*i 3i?*rgy ror the first time towards the end of World Wai II through the use of nuclear weapons. Owing to the apokaiiptic affects of Hiroshima's and Nagasaki's bombardment the term "nuclear" has become equalized for the world public with '"mortally dangerous" and this has resulted in opposition against all nuclear. Due to this fact the designers of the first nuclear devices have had to take very strict safety measures having no parallel up to that time in any other Ъгахг-Ь. of engineering, thus diminishing the probability of a *-IK±L>JLS accident hazard to a very small value. The nuclear safety philosophy in the first two decades of civilian nuc lear power development has been based on the assumption of a hypothetical great accident of nuclear device. Combined with situating in a great distance from densely populated areas these safety devices have been assumed to prevent from •its;, consequences on living conditions in the surroundings f the site. Considered from economical point cf view it is evident that said safety factors, both the intricate and money 'onsoming technique, and the isolating distance, have a nega tiva influence* It may be also stated that the twenty years long utilization of nuclear devices has resulted in an extra ordinary operational experience. Ho case of an endanger of the population has been recorded. It is true that one cannot speak of an absolute safety as to she future, but, on the other haifcl, it becomes possible to evaluate the degree of probability of an accident and its consequences. The new KHthod, based on a|* probabilistic, quantitative approach to ťie acdd « 8 - public hazard. Эае following paragraphs give a brief r 2,1. Nuelear reactor аз & S.JI' * of potential danger Hx'.lear reactor is sometimes incorrectly called the ""•• .-rit;v; lied" »t,:,m,_... t-.-zOz iv. ?'• >• -; s*nw that the star- ,.i v' чг> angaardeo. геа-st,г should lead to a nuclear *•:..' ři'-*:>. f ll^win >y te.urperftt;.-.re and pressure waves and V.7 ':b.'9 effect ot •••.•ad.i.&t,.l:>Ti within large reach. It has t-**;i pr.vred bjitsi '• '-he /retically /2V 3t 4, 5/ and in prac tice too /6f 79 8„ 9, 10/ that the dynamics of a "star- t-iiig'* unguarded .rea^/r is t-; slow compared with the de tonation of a bomt ani is rather more similar to the ex plosion of a standard steam Ъviler. It should be stressed Ma-sit the fissionable m serial in the reactor core is ve- :-y "diluted" (fa"^r } 05 up to 10^; í^hitói, along with the presence of large quantity c.f structural materials «u,i '.'••• riant results in an intrinsic mechanism limiting t-.i* т.на?Г1.!+ ".-de ani the rats ,.«f increase of an unguarded resits/rr power (propagation of thermal affects, increase p the temperature of neutron gas, boiling of coolant in the reactor cora9 bubble forming due to fission products, the Doppler effect)» The potential reactor danger consists mainly in the ťig* amount of highly active products of fission, accumu- I%';:in;i in the rue." in the course ef reactor operation* + ) i, et starting of a reactor on prompt neutrons without -A,y -ii't;. »n of 9жЬ*гь*1 safety devices (for instance neu- *:•••<'.;• a dbs-.rbers acting in thf realtor core), 9 : According to /11./ г.'-'л-- ;'-V":H -.ví ^ňjjmd .-vviviny -.JÍ : i-, j.--"-^ ••'r-3 '. j a raav.t..;;: A;i >;-: .,; rvťT-'- ^'é power c-perv.i •.=,•.. th* p*r?L--.d f.f s>. • .- 100 da.?=. : я s.-m* Г tiO^/SSft'+r,. f» •.-:•-•..- 1 1(-:{+;-лЧ1д#| -.азе á. '-^"xa иг.- t • 8 MO^ff .' л £ *•-.- e« .<-*. "V:-.v ' , 12/» Cto* dAy aťt-ir- :'«."•.-.' ••"•>!:.,-"." a .--..vn -';? a-^.j."i?ty j., .HT'o;.mc I Ш;:-./МИМт* From the rVrsg-; lag aat* It и* у v-y . - veá that a 400 М7/в •»-• * a • t - г г*рг*я-.*:-Л1г ^ sour--"** .•.-" •--.-.;.. r4.- Q "0 -ha Ir.t3ris.it-y «.-f w!i7 on is soma 10' tv- ^0"" Cu, wái-vi -..,,•••• xlaa'vaiy <:.. rrespob-ib •;:• Ю"" ..-.p r- -•••' ^ •-ras of Ь.Р 2íc., Á1"*'"~ar the 124^ "„.lig :f гз'а-'.t.r -. r> ".Uo«s ;•. ;•: •?а~1*».а-*« c.ť fission products, Ťhrt-зь ъге cbaractv-r iz-^d r.r • р.аазав iaitial a. ta.v.it";yy Lei^-l гк;*,, аг.1 otologícdl ::•-.г,• .-.• x /13/• FrjJu the bir-log/ca.! po:.£G or •-..«* ^L* most 1;-.1-:>.- of ail fisslob pU'li--3 l.í i-5 ^^dilj ás-íomed the 'U=č..;-:v jsc-t;-p* 131 /Л/t, Thm sffs.'-t of c-ther fission pr-oau^tt; .-.:: .^adi,oasit.iv« materials т.^> "be impressed hv m«acd cf ar; appropriate relativ* "weight". So, if in сазе с*' •?. rařot^r a-:-.*i.d*i:t w* ar* ьао.--:.•.:;,-•. «aough to prevent rhe iodixe isotope 1,31 i'row Its prv; .- lion into the r-sactcr surrcurding areas it ee«m^ :-U:' & A 400 m* re^v.r «HIAWR <.i аЪо-.л 10 Ou ,-ť Loa-,-- 1 "?1. Supposing <*j±tžcv ..^afrle. cliaist* čoudit: . :»ES and a LSSN ; A "•-•: rdi rug t;1 /15/ wh^i t-.h* fae'l irradiation in tin» :-• t.jr is longer thax. 2"'-:-0 ••- 300 о.:-уз¥ vh* meat dsusger^ua a.r-* the solid ••stať* pr',u^-;t.6 ^f ř:as.ica чшзсКг th* «sěsuaptir'* that -.vib.cal d:-ď*» <-.f t ... ^rjyr.sid г . -.-.a is 300 r#mt b-r-..; 25 х?вшг and thi.-.*: from ^o.a fušl Г*1««АО#Ь 25% of доЛОм* -•;,•-. 1% vf 4'/í..vl-č!,-ř/-!'. rles.lu, ur-avts/ v.? distance from a 3 .-w populated area some 1-2 km, a ~*- ; j.v5 vea.-tor accident w.-ald ;aise still permissible xelea >s •* about ± .''«?«. Л brief rsview of nuclear reactor safety Thsre are three different aspects from which reactor safety may be considered. These are as follows: - safety in the sense cf operation capacity of the device .' har'a ••terjzad for instance by the period of designed p owe.? avai lab 11 i t y/ , - safety of personnel, • 3afety of population in the neighbourhood. The operation of nuclear devices in the last 25 years лаз given excellent results of safety for all three view points mentioned above. Very extensive and detailed summary of nuclear acci dents in the period 1944 - 1964 has been made by Schulz in /I?/* In this boi,;3t 901 cases from 15 countries are recor- del» beginning with usual transport accidents where radi-» aťion has bean involved up to undesirable starting of an experimental realtor., Trie book includes the damage caused l-j the testing of nuclear weapons. Review of mortal cases presents the following tab Lis 11 - Table 2 Type of nuclear device Mortal cases or the field of its use Owing to With no re.-* i , radiation at i on iffiiCj. c&v! Critical assemblies, inclu- 4it% military cases 3 С only military cases) Experimental reactors Bart&cla accelerators Gbeaietry, engineering, me- dicitoe 8 Transport, including bomber crashes 5 СЪоаЪег ereshe5 alone; Surrounding» of nuclear weapons teste 4 Nucltar power stations, marl** reactors - Total 24 15 (Military cases alone) (9) The average number of employee» on nuclear devices in the whole world for the last two decadea la approximately Pro» this follows that i» nuclear industry some 0.6 aortal caae a year due to radiation corresponds to Mr employees. The working accident rata in nuclear industry is as much aa by two orders lower than that in any other branch of "classical" engineering /16/» - 12 - In addition to Schulz's data there are USAEC data collected by Back early in 1966 /19/ according to which there have been altogether 31 accidents with 7 mortal c.*.ses during the last twenty years of operation of 306 US reactors and other devices with a total of more than 1 700 operating years. No accident has been reported to endanger the population. Besides, no serious accident of a nuclear power station or marine reactor has been r-t :-ivTded. The extraordinarily good balance of safety of all rea.M'••?.? types is not unfavourably influenced by the higher rate of nuclear power stations being put into ope ration in the last years. Table 3 presents a review of operating years of nuclear reactors for unit power excee ding 10 MWe, tabulated towards the end of 1967» All reac- : -:«rs are in operation with a high degree of designed po wer availability (from 75 to 92%), with no more serious a<- -idtnt events* Table 3 JTumber of operating years of power reactors with unit po wer У .10 MWe as of the end of 1967. Light water Heavy waters Gas cooled reactors reactors reactors Other types pvrax/ BWR2/ SGHWR CAfflW Magnox AGR HTR 36 40 0 6 152 5 2 75 U:970) 1877; (225) (4897) 130) (53; (1823) 7 6 6 159 (266 7) (319) (4980) 3U 3 (978< - 13 - {./-.рртлкга'.-ntiily 100 opsi-?,*" it)g 7^r»rs '.; v i 2./' Ejfv:addne; TQ*ÍJ?.*---V£ -nibh. n-vA^^r svps ••-j:4...'. ng., i.-'tu:.!-.-- d.^. ;тЯзг '-.-.•ther tjp-is^» У Pj^ssurce ^4ss4x HV/R ari; iu-;I:;i*d u:ui^r "other ty^o'V. 4/ Fimbsrs cut cf parenU'esiA й^Л-й th* timber <,f op*- rat.Uig yaars, t-hos* ia parenthesis denote th* t-;!t-f,3 i.st output cf the resotvrs.. 2o3o De-7S^:-pment of safety phi Iusophy The development of any hranc-h of ч. ..:. .ч»г1п£ is asso ciated with certain danger for the public It may he -aid " "h*t аз ">Л -; 4S the d.is."ovř.c-y fř 11*! чаве of nuríear energy the first what has bean se^n on я new hr-?.n'h ot engineering .Ълтэ :-'Hsxx /'ts U36.f'.i.i ^pe.'s arid o-\ly ru the course of its 1//5:.'Д:г at:- on in a largs о xt *•-/•• th=í dar^,*r ,,f serious damage ^.pi-sars^ !Ея5 eji.p«sri«ir.-3i *:.•• ••IÍHJU.Í-ifc«d ы f-.r snows that fcha P".-.V:.: •". :'s n.-t int*restei t :• - ir/.L-'-h in "hš lama-ge caused by th* Litiiizatlon cf "5l?.3Sl,-:a.:" eng.i Д**П.*Й (.p-.-llatinn cď r th-, .4.1..v and 4vat4v-s '"v lands'ap* d*vastaT;..; . n, «tc,»Ju On the -.;,"..tr-^y й the development of ťiyil.lan л v.. I-sar pever utili- : 7..:,--. ha-» he*n а - потрав.L*d су а strong initial r;:os.itlor- - -h-í piihl'. . .. iLi8 híiii í^íjn thy result *;f the .fact that ;•.'. ".ir e.r.g' ц^«г1"д has demonstrated in the vary beginning t .Its deva". ^proent a real possibility cf endanger of the + - 'У- should i:^ r,,.ted that r/ae radioactive exhalations of :r?i..f -..к p-w-v stati.^is г»."-,", ню."* dangerous to popula*- •*".l::.. arid landscape in tba.'r vicinity thy,;;, present naelear П т/ч" í?te+•.. :'J3 /20/,. 14 ь%л stance of the maritind. их2±щ .its da7-H.Iopukat„ toe Flinty pra.losophy h?.s h-.ч to b# ^onfrar.ted with a wide «aotitmal opposition agairiřt n.>. ,1,алг p:>wer евд,aeoring,, and this is the reason why to* philosophy has ha^ to accept, with no experience, the cri- t:-í.rl•;:-"••. '.-f the extraordinarily low peniussible danger of a ...,•..,. li^r r*»st--•''">-r- &.;• i'i^'^* '-• ^is^^.ier^.-ís. The safety philo- --••':j" 7 }---.:5 bs*:_. Ъа&чп. v-*-, Í, ťh.s -rati -..-ňl experiment express*-! t.;.:-j:' A\gi. ti.*; зсяъ5лл>;ш •глс.гЪ'ч я. -:'\*r:t", MCA, character!-• 1 ^*\ ;i_i th-í f.. !.] cw-iug ai^stii*' /21 s 2;?, 23/: •= e-zery re?, -tcr шау ехрегдепсе a serious accident whici;. "м.:-.у bs cUf:rn*d as the moet credible, « -fy-iv. jf H '.is a\-'vi'i'jat as very improbable, the reactor •i-«sisp: m-3+ i3TipA,y ~!1 possible engineered safeguards to reduce its 'ОП£^'5,ичп:ьз, - «t.-,+ im^^i'"^ '-:.f rad • ?.t/,: r:.r> effects of the accident is made iad*r the assumption that fell safety devices fail, - +:\* v ^-••a.r-.+_ г*«>;Оз ?r. t'-ii. г-<'г*';.%я .-f r-adlcactive ma- tibials int» rea-:tf,r suri>'isdirigs, - •'••л-* radii*' «vti.vl^y :^issB u^peroássible damage in the de- fir^i eir" ilar i^giv^s .yro;:.nd the reactor site. In these regions only I'^.v density of population is per- It may be svi thiit this philosophy has put into fore- £•:-• :.••/ -1л ••o".5ej .r<-\ is i-f a :v*a-'--tr>r accident, with no •ac- c.vj-.t- of tJ:e pro^a1- •.: i^v of its origin and the reliability . ? t\-.< saf-^y &•;•>' -'IB lv.it «Mad» It has been successfully из-»* -w^.-c. *-.Ъ.*; f .i r-s* ,лч*^^•-•.->. •PH^-.t.-.rs and the first gene- rat i'/'s cf p ^r r^š.-t, r« -•* i,:-.v arj^ medium power*, In these , ^*s fh^ ;r^..i^s- f •-.'*:; ^••lat.ile fiasir.vra products from the -..--, •*• ,?• v:;'.:".x'rig w•••' •• я.-.'" ,1-чна to as тарвгта! sslole irra- с/л/i ••._ '••? tu* p ^/'Л"1.'.* or. b-t;/-)-,.d a giv^ri radius* 15 • ,v<: ^.я*;*.•.-.-•.5 wif-. =: pc-w; ;,f ^:- •;: ~ 000 MWts >лт?в а ij.:•*-•**• i'vfl.U4n:"'-s --'V. v'-..:: 3.1 tis?*".ior?.-. ТЫ ÍVIOÁ pblLosopby Кай < '•' >ГЗЛ*Г f • 7^ f -..•'! t.-•'•.=: -Í.S^TV > ^j ly 1С Sil p';'a^3 Ъу ť'i< : ••-•• l4'"g4 a* i^i!v3 Í or»s:i-•- ;f а ч-Лн1.п of events9 aver;v oť " - 'I'.'f :.">.•.!-ле; •"* s •.;•>,'•.''. <*Л '- га ах i ar не ."--.-..иЫе; --ч-"."^ Д-J'^.^" •*". у, are -^s *"->;'+..!•-. i". 7 ^<>1:.;^7 •'n ^Ь.<5 pro":^''' ".. ч"7 ' r fa" : *ч* Í/Ш p-: ,>-,:;•;*-•• •-••;:<:• * . í-rr*^ ;..soi Л i-vA p-'*•'-' A.H'V- "t ".'•;•<-..-'.s: >':..r r f f :. 3 мз f Eк v• . • -:., ' .- .. :-!•,•••.''• --tEió •/:• •>•• "* Ж";~ 3*. -.V; M-'/A ph: "• %-; vh./ '••:•.. ••.'' - \--. ' •* ч ; iiUti+-»т. i уч ^.чМ •• m-.tiiou ~.f V-П'-! dbgx*••*•:":' рлЪ'1л,-•• «rvi ч :o v uniemi .1а^й, .дд 'ů* - у Г 3 а йч'Ьаг ро'Л'л.т statton,. ft.;-, í • .joml ^! '1г,"1.к on "'•'••«' s*.ř*ifcy asps-^s of ^ffií-i^nt 4'i)V pv^-sr» statics of prosen" t:;pas i-v^x"* soina cp«ra- • v--'.', ^^ríii-'» i.3 •здгча'чу •-.••"••j'i 1 able) ar.d tL.^ appro- r.ius; parspíf-tlv^ uř tha 'tл<. 1 U.Lag of ácaíaai v^bí'. íast . "-• г-ч.я''+;-/Тй íwt-izb t.>i* ••ii.-íp^^i-vi*! втфвг1а:й;^ -jbtslnt^ а? v. '".•'• ayuft'oii'c KV р|'-й«*1Ъ.А# &*ťious ; 1*/>д Ь.:^,; I в яг.. ir>. *'«>,#'.л- ,уваге to .important mud ••"'. и^ ii s«/<*+^' pru 1 o soph??.. i6 Ia tne United States now in use is a т>Л^-П**1 DBA (Design Basic Accident, /24/) philosophy,, wbj -h is vnr^ similar to BU (Basis Unfall Konsept;, -wv-vlopa.i in the German Federal Republic /25/. Similarly as in the case of MCA both the above mentioned philosophies art* based on a basic accident defined as the highest possible dan ger caused by energy aad radioactivity r«l^ase from the reactor under consideration under the assumption that no safety measures are taken to reduce its consequences. Now follows a detailed safety analysis luring whi>h the chains '..-f the fl-vxrs* of the accident are v-onstnv-ted, including the probabilities of individual partial events and the probabilities of failing of individual safety devices and measures. As a result a probability of the basic accident is obtained evaluating quantitatively the effect c.f safety devices and determining the magni tude of danger associated with a "large" theoretical accident and its consequences. In the reactor design it is possible to find analytically "weak points" in the design or, on the other hand, to find where is an unnecessarily high accumulation of safety devices. The most progressive reactor safety approach based on a quantitative evaluation of radiation danger has been formulated by Parmer /14/. The danger is defined as a function of accident initiation probability and the extent of its consequences. The Parmer's safety analysis needs that all 'sredibl* accidents, including those which а.т?з not credible according to foregoing philosophies, be taken into account and, every of them be supplemented by the probability of its occurence. Then the accident chains leading to the release of radioactive materials must be followed, and for ail the respective damage extent mast be determined. If the probability of accident occurence is expressed as the reciprocal vaia* of the ti me period of £'*a<"to.r failure-^re-? operation CB), and the ~ Г? damage as the amount of iodi&e 131 released due to the accident (A)t а В - A diagram, representing the "accident spectrum" cf the object under investigation, may be con structed (fig. i). The object fulfils the safety codes prescribed if th° total accident spectrum lies below the Vrundary curve of the permissible danger (fig. 2). Parmer's safety philosophy is objective in principle and enables good comparison of nuclear danger with, that in other spheres of activity. Said philosophy is obstacled .from its full utilization owing to the shortage of expe rience in the dstermination of failures of nuclear compo nents- To bridge the gap it is necessary to collect and e,"i] uate systematically all kinds of failures experienced .a those nuclear power stations which are in operation /26/. 2.A- Safety devices of nuclear reactors Safety devices of present nuclear reactors have been enveloped essentially in cor.ocrdance with the requirements formulated in the "Maximum Credible Accident" philosophy. This means that various safety devices and safety measures in the project have been designed without any consideration to the accident initiation probability, lEhe same goes for economy . } The proportion cf capital charges into nuclear safety of present designs amounts approximately to 20J6 of the to ts,! cost of the nuclear power plant. The main contributions are as follows! a) Emergency systemst as for instance cooling, filtration, stand-Ъу drives etc., which for a 500 MWe PWR amounts to some 10 2/fcffe* •>.: Oi'dtainment. For a ?00 W.H nuclear power station it amounts to 4 - 11% of total installation cost depending on the ty la D*:'z;.r':. :. •; í "b.-3 Л.лЩ~1л. аз т:':.ы p.v ^aac-4: у С an evident v ;••^•,^г'8>л"гч р^ЪаЪЛ.- y an., ".G.S-I *»T':''UÍ: О::* bermfui cons3~ Ч'Л-т ?5 -1П-Ъ1Т»О ť 11 vi:'й ta* sa:!S4j vu^.ineering чг all ths sa*^"* sy^tv-;ji.ó v? эп^жпевта^Е -end •.-r-s.ř.^i.aatioual - ш*ч-айиг:*£ ť rad'"-, a-r.sldeat conseqaenc-aa, 2i4-„l,, Accident prevention It acts in the direction of reducing the proba'Mlity of initiation and propagation of faults and failures which could eventually l»ad to a reactor accident. The first p*ri cď the pr^Tínt•.;, :-i* i» the extraordinary quality of псу<вт atet?.;-;a ďw ч^ This principle of quality should be re garded as *r, integral part of the project, material se- 14-itt^, design and man'Afa-it'are of individual components, Ъ vill^iig :f c-bje-:=ts ar^d ^ra-oparational verification of *Ул - ilpm*^» The project must also contain measures to Ъ* Ьахш t,, avoid tins faults of operátorss to train and examine them, t;- perform periodical inspection of the ^'...ipment^ *t; „ The second part of the prevention are the MiTiísLJ^ZÉÍSSe indicating the faults and failures and ftithar warning the ©peratcrs to intervene or automatically c; Sxt4»* In ccmparison with a conventional power plant it in by 1 - 2 $/]cWe higher. X; 0 -.at ad-lit ions tc the: t.r-rmsrtasaion lines due to remote sites of n4o\?ar pcw-r stations According to /27 and 28/, f:-r 3 500 - '--00 We, 230 k\ ?ir.a Md 50 km distan ce fr'oa txj"9 ventre ^f power consumption; these additions шау ". ','> in the region 3-6 ф/к^ч. - 19 * ; ex Terming the safety measures up tc i".he r-.^ctor bhafc-cLo,u 4 T:*e safety systems mas ' be j^f л:. у ••" orsneoted aad, in ÍV-ÍI- + :.\-:\, every important sapety funí .ion must: be se.ared through several хМэрчпавгЛ c.fevices* 20 т HIGH Ю' 2 PROBABlL:f -?С- OF FAlLURc a- с о 5 10 LOW L',1 -ОАУ.А^Е - — HIGH i 10 10 •in' A 1-131(C'R:E: Ft*. 1: EXAMPLE OF AN ACCIDENT SPECTRUM OF A NUCLEAR POWER STATION / •ir НЮН HAZARD (л ^• .-HAZARD BOUNDARY LiNE V) -j. SHALL HAZAkr or 1C S Ю 1С 1-131 (CURIE) Па?? EXAMPLC nF -.WZAkQ EVALUATION i'ft '•-. - • ii • ~t ,я ••чп ^•' Xing; л •I r ie.THitv" •. bit т- '. t-t-ъ; tr.i;.' , !--i„ ~f:. -• < •'..•r.:..^. - sh.oa- —- _; г. *" :_ * ^ £ 'V').- ^ О ">V ..,.:.-; -:i . • .! uom-~ '•V'.'-ib-: •'••< "V ŤÍ 3--• S'i , O.ť-íSAur-S .'JS • :. •' .'Л г. "Г-.*-' I" •-'Т'.-'.'.г- ', ví',, i-"" V.,-..r- v o^t»s fár i ч -- :>. . - .f prima ••• ;,rr \i..3>i >гб '-. • . ••••:•••; •„ 1?Í;-V .•-.-•. •. VHU.; .-••ns ha, -•лО-а:. г<»<~" :>'а '\- '•. •..i"'J ::Ua рг^йя;-7-- '. u ." H i %vS лл,.;. -ар^ "Я 0 « SB 0 co ct и* p. •čí к ct ct o» »•* p ct o i» et t* % tt g ? O Ф a 0 5 & ct o •i P к* n Ф & o Ф co % •-* tt •tí ct 0 o 3 » e ct H> Ф o o1 o 8 čt l-b kJ 4 0 »tí i et 4 i n AC t <14 J "'., ':Й£ & = •; il'JPIEAK IV;,> S!Vi!iL'íí ;..:•.' Г .•> . ::.,.' -Г L-.•; .-г P^T;: n ;?ith a gas •'ř?- _:, •••.•-•: . • ..*•-„ i/ . .::!<• r-;ó:t../ •ruiclsfil Iv a -п.- -. .- ' • .:..-,: / --ТА •..: .Л-з i.-г'1. i:i the wo:.-iu. - •••• .. v&r., . -iVv.'- - c/r^r^d &v:ii -.-.the? í Г5рег l *.,'.*.. ' ^"/O •>- ;:.--;v -:i :-^s - W.='.::Í S-.;2UH sperxtio •'. TO *- •.•>>• .-.'.v. "\.-:-r tu ""MÍ ós=ential devices d í.íf-i'lv jvyett.-^ :-r tne power station. ůLtiň*: /"55/ .. V-1 •"* •- '. >J i: 'A ' -> J •< * * =S Л v * * ц fr -A ř * t Ť • * » » - Л * % • * * * в « ft ^/W& i-'-.'» "5 "I* '•-' - •-• '"'..-'•''.• •';'.'., £• i 41Д. i .'? \AiWfe .« p (. с . - ,-; с * >. * . • • * * • * * » . » • « IT/ u''-> ?-' '. ...» i'y. •-••-•:••"•••.• .:...•, .'.;'..Í::U'.'-' v „ i'eact":.r cor* (offi.) *..•• 3 660 iY"* " '. ' .• I",' »_,ř"* AÍ '»; ' /i".' V-&3& * ,s « v * * ; 'v " <> H; » « » « r .«,'""•"• ' V-'....' ./'* - л - - . J '. '<-* x *: ^ i ^ L." X „* л 4, -i >. * •-- f % ví * л * * * * *» • O /2 Jíl- *.>.;- tt f,.-.;< -:!:--xr?Í£ ) i: "be realtor o *,»***»»*. »• 148 Г.'.'.'- . *•-. i *' *? — 1 !'"V --Ч»" '•'* - ^ '•-' VXiUTi ^ / * « i •-• W 'й •* и>'и .; íí'/..' ''-• ', '. !-•.*•-. ?• C\ v- -: *--f£ ' >'\JJK>. i ''••*, К ^ J f* '• "• л * •'• * * * * # * * • * • Oj? o;-.'-./i.-.-;. t.r-, ^.-- ., jc-adivcí .'-f the realtor oí.re ClsW/i) 10,9 №'•*"• -oyí-'л.': cf г ;i..č:ía*.U'2a t-'iV-o - .. Í »,••.-.,,»-•..** * »„, * • AI alloy Ех>:5»гп'дГ. cHa.íi^tor -< . eiandiia tube wall H5 x 3*5 trC'.A.'-Si.v.?. •«,íC*.' ,' ,i , • t- л -, ., i •.,. ... л i- r « * Í i> •. .-, & .- . «i*.». XcO X y»y 24 Material of. fuel «\a:.iec.t «.t-.v•:.;..-•,•.•-• " • '••* ,*..*..-.. Z-r .MIÍCVÍ External diameter x envelop * ub-s v...-vU 114 x 1 t&ivkrisss (mm) ..*,., »».,,, *»„. * .,•...»• * ':л2 x 1 JL? -JLJ i tv»***w* »•?*-; .. 4 » » i * ^ * i i- -*. * > •' % j* * л * * « * w * •» лЛ1^ •« «A- ^ 5 *JbA О Fv.-*1 ' -i.r- m-'^r /1 „.«.,...» »• , , ~ - * j ...... » i»Ig-B Trtál ói:arLÍii ciisr^e in Ves v.-,,:/.-:: Cooling circuit: С. Rate cf coo last flow thrf^gh the res-t.-r- (kg/s) 1 400 Coolant temperature t react- г ид i -л { J ) ».,».„,.,. 112 react-:-1- олгЛ*;+-í i;} ».,**,»,,• 4-35 Coc-lant pressure, reactor u/Ut (кс/...;:Г } «,•»••.* 55 ť reactor outlet; vkv/;;iB ~,5 „..aí-, 55 Lí-«9 of e-í3í,lant ргеззаге in tbe ::ч-г-:^ • •'•• ÍV?/.^} 10 C^apr^ssií^i :ď tj?e ovci-Árt in tn>s •. .;.:.;г1н"и«2; 2 < blowers СЦрА'Л i' • *•»»**•»••••»*-••-•••«* •-»••• 12.5 - 25 Secondary circa":ts М-ЗАХ. ргзз;чцгз ^t-iř-ти, $?.<*&*<. ;.:'i> (kvV '~ .•' *• *«• ^ .* * .»••• tamp^^atu.:'??» ( '0) »****•»** i •*.«,.. mi,; Low pressure stoat., pyeasurs C*-p/<\m }••»•*.•.••••4,«. teaperature (°C) ..., ,..•..* IS* 3*2.. Pďwey station m.dlri equipment and its thermal й-íhemetic r.hart 7T-:,n p:;4fřir ítfttion ÍS dsSigiVj Í 4S i;. .':? •'-o^d-jr.ieatins; W.lJ;'.y ;зе E3-150 nuclear reactor witii throa turbina blocks. The аь...>: coapci-Tiats of this nuclear powftr «tation are as fellows: - nza-i KS-150 b'.avy wa',*r9 ---mpresawi care-one dioxide carbon dioxide cir culation. There are two blower,--, in every block; ~ t.hre* AK-50 c-ondensating turbines* 3«2Л. Principal thermal schematic chart of the power statlcn (fig* 3) Th.* A-l nuclear power etatioa has two closed circuits. Th* primary carbon dioxide cooling circuit consists of six i:;tl«per:«dent branches r.^nnected to one roactor* There is one «ť*í«n fis^n.-srator, one *-i rculat.tftg blower» two main gas pl- pir^Rj, .and ,:1.« auxiliary equipment in every branch. Two pri- .«•?'..-"У V»raa';he»s fi-пв оглз block. Every blv^k еГ the secondary wat — 26 - ťifi*-rt tw,« ?ста.*пьс\-г* ча-Л *ti:5J.Xim\r *~ Upbeat. Тле -Voline: Kas, oran?!-^:--*^ in иЬг h.J.-wers to a pres- 2 5iir« :--f у.Ъ-i.it 66 кру'лс , is >v:i through twsive pipings Ctwo iz. «г?*™? Ъг&лсд; „.-.-.t;, -л.ч r--.-*.: t.•;••:? i&lei с!..аюЪаг. In the .•.t.':iV.-K-ři of its pasr.v. £,' j,.!r:.i:,'--'i^";. ^..e .tr^a^-tor active -"í-re the ЙЧ.-З hí!.?.ts f'.т VÍU 112'"0 t- 435 v, its pressure on the outlet *•-' *ач гза-jt^r cutlet *?^&Ъът е-: ^ «ppr-xJ mátaly 55 lep/ / :c:% Ihrcafrh the twelve pipings the gas is ^applied to a.:.;*. ;-*е*э& generators 1л whi^-h it cr-.ls +" ' ti--s tamparatura :f 5C*' G and goes on to r*ech the suction of the six drcu- Is'/ing blowers* Т'дч 3ft-.:-:;sla?y '-irc^it consists of mean and low pressu re .3tsi-sm i'-i:?-.- ; its, Th^ m-^лл pressure steam goes from the superheater's - f s-^am gensraters into the mean pressure part r.-f ths t-rbins» mixing ^fter its last stage with the Iw pressure steaia to fs-ii the low pressure part of the tor- Mi-a. After degasifying i.i the vacuum, degasifier the conden sate is fed into the condensate heater wbioh forms the first stage of steam generator low pi"í.?aure part and then goes on ii..ti> the feeding tank» Рг-л said t:iiiik water is pumped into hcth. circuits of the steam generators. 3.2»2. The reactor It is a heterogenous heavy water carbon dioxide cooled rea^tcr encased in a steel pressure vessel and using natural uranium metal as a fuel Ifig* 4), The rear tor cor-e is located in a cylindrical calandria raade :ď aluminium all*y. The gas cooling the reactor enters th* reaefcw inlet chamber í-hroigh twelve nozzles in the upper p*rt of ьы vessel, from which it flows downward through the furO„ oaean&ls iat.> the thermally isolated outlet chamber. Into - 27 - v:ť_3 HXMĚ:••»:' :t* -.-ti#\vtt f;-'" -\v nci^líiň iBcI&tí?.-? or tie inne. s:."iVit"s o;; '.V- чгн.И wí ••-:- у, f.hir-ws.llad «ubs reacLirt- ?•: :^ J.".5..-4S ^.J: :. f.-4i•-*s ths rvia-tcv cere ie surroundaA •--. s гп~г-:з.--Л e^;?.č>i* ~\ t\.e ICVMT ar.,1 the upper part of VL* с::лзя,и?^ v%ss^l there Is a Molo^loal shieLl consisting ••• ii'-s.plrLc-a ar.d ir^a layers, in radial direction there is «. witňr biologies! áhisld. - 28 - l_l i : V si 1 7TJ o ч 5 S l ;; I i -t i, r? -я '. - control .~r :• d - ирр«:л bi •' • .' 3., ÍM'.' ~i - reac"or £• — =>• ^ "• "Sílírírá^ S"le~. d p - i'es:'.;.' ^-•':•;-* 8 - hot ga* .'"•-'ятг.-е," 10 - heavy ^feter xrúei- fneZ ^hanre:. I.- - water biolť.gj.•'•э :. shield 1? - <1C'Г'л.. П£ gS»3 \r-\.*.* 14 - centra, t:;.---lce";.'."-ft1 bc:.eid "-> = fuel elem*:,* 18 - hot gas outlet- IV - support i .^g oase 20 - heavy wa+-,*r out, Let, í : Гэ 17 Jg. 19 Н,0/' ífe DtO f .6f , iCL FÍG.4 K^ •';, />cw •;•, line в^-v.-i T'-.- V - *^' V чз лс?:'-?':.-'-."" ' г ví.-ř% lists .;! J»"b.Hpt:-i ňlaeisiits of . ,,,., <>+-• •••., ; -. + • • i-. v i- •>; t.yT-'- J; S "*: \bt ii. t:,\-. -„ .:;. t:h^ 159 s^ d; r. . tub? through VÍ.-'••;'. r?'••••-.v3 id-: '-."'- ... .. „•;• =.,'-. <:•.- ..^r;, .". 9 'dibes агч lo- c --"":-.., :''";..-•.•' \. ••>.&-••"••.< v ' - -d .-=.' k - ' 2,_ : .:• t tne medium f lo- ,íi\T^; i i; ";'••-• "Л.-- V-^-Г fi.rr- ;v '••" ^ i "' V ~ — -"V-'v. The i;irC!.tla~ r rfj :.,- ., ,f -;.- -st *=ir;. = .v--- :i •••--• .r-. ;,-.. tue е;Л oi'atd_ig section ~-. f T-. Tiw •*u-»~3t;^tíd fi^-d^T v/hji; iiriii: oi .,'.-•;: :-:1эМ rig blowers i'l.e pr.iaar-y circuit -;.гт? driven by 3 1№7 -yrichrormous mo- *я* Tíuy del: v:ir 1 000 *: :rur ^ith М..з .-.varpressare of С.--*.,-*:г..'.;. • •••' ;.'•• •;-.- . ,v •--' ..... i^i-s Tv; Kieam of tur- :д f^ač «w;-,'. •;-. л.:.-.- '. ' L-- >:,'». /dc w. •-+;'.-.'• - Tivis should ••,•:..:..< HI: .••iC'-toaucj.'!. . ,:^.-.;•.<;. „.г. d..-. :-=m:^ fr'-oi 100 to 66%. р ?! ШЯ ! А-А А- «=| -А СО! hfí 14., & ^ в-в .-Ы* *.9Д f»4|-5; ffS • 150 Ri'ACTOR FVEL L'L EH EN T ¥ » :' '-^/Ь^Ь-л =^ v 1; 2Й> "£^,'ví Ti/.>:' i. •: : .• Í.-;;•.' -т* :~E Г'.Ш/ЛЕ с' 1\. ifi/cnt -« -p /= /Í-* •-.' ••/:,iT.Tř v- /,- i..Я í: •J'A'-'CAATOk ',4. -••# L —JY KÁTSR INTC THF - ^ / P KtV- ' :"ř; V/TF/f FS"/-' 7H.F ":f • ifi Fia o: '-PVm КИГ-ЧАТК CHA-.-J OF THE STEAM GENERATOR - •-: .с р--у ~> .. £:.-• 71; . -э.'.'..... ÍV *ч -Г. <.*>-, О 1 :7Ю .•..-i :-;\ U . 3,2,7. В-- u.f^-; -.. --.čil -i >Jj •z • Hff-'i-'^---,--- " • ' -" ' -i. ./ WW .• ~ ^. Д '- i.. .*-. .. . . 1..ÍÍ-. V'' í. -,'ii 90° С :.и с-J .vied xr. -vv.' l T1"* Г'*Л Г*у^ S D-4-~-i£ f >, j TK* " r " ÍÍJV 7-jg -2;-, .^ ý*:. ..- bO E.trc-> 1 ••- а с и 4; • ••'".. ; O i," C.-lid* Lí»?. С O.. » <:,) Ж»0--:Г;д1 Jr-VC'0 :'Derated rap ř...... » t>-' ,t л .•* . ..- ..--,-• Л * .- . .-*. ' *«• •- •*. * •> •** •*.••*- Г •:,% ла£: Uo\.sr*;.:^: •<..-.: -••• ч;- " :Л-i..-^i >-'f about XC# F.T0, 7a SCHEMATIC; CHAR? CF Till. hkLii I-T fAKT CF THE FPIbiARY гчрсиЛГ. .rigíir. part FICi, 7t. SCIIuM&TIO CiinJST ОУ THI3 kAHJ CCL„ РАЙТ С? THE PRIMARY CIRCUI, :;;ght part A •-=* ri«.''ju; hit • <3./.' 3 - pr«i--B'aV'i v»oi*l ax.:. D =- rlgv:. ръг*. riumbers at the pipings express tne ler4rth of their straight™ forward par?;5 the values given have fce*r measured on the Ms бКО design; all u^d«v5igr.ať?d ra'I^i •,.-- th^ ?00v^:.a.. } p;. pi rig ore r = 780j г all и„Ла?„^г ',аЬ^с1 ra v) "-". ••'*".* -.••-, •.:.. a ••> . r,, I-, з: ap.,.lti'3 J5i3.1, í--i o ir; .vir of neutron desslt-v control stav. ".-*fs th-. -=3,.t-^ power in relation to a given valvH. Tii-ř :-ф,.'т andsr cnn^r?! .1,3 charit erized by the feran-^fai' ví гча.Л-.'Г r.e.i.ti:v,:j kinetics» CXvLng to the depen dence j С t-Va amplification e-^.f.. ;•: ;.згЛ с i' Vánetics trans fer *r> the reactor power Laval i,he controller of neutron Д-.пв ;.'*">' ?.я male a» adaptive, thus enabling to perform fcin •• :xv*T". 1 in the range up to 3% rf с he nominal value, tb.9 sensitivity b^ing 1%, There are two such circuits in ths re а г •;•, -ne jf them serving as a stand-by automati cally taktia-cver the controlling inaction in case of faílur--. i T-ČLJ ation on the ;*-;lr:;-.i.,t- wbjch is in operation. 39 Fif.8. PRINCIPAL. SHEříATIC .;wofv POwLk L. О''' I > 7 í £ J Г„ J. J. 11 AS J 8 I 1 Rt-ЖТОР 2 ST £ AM GENERATOR 'J TUK80 - GENERA TOR 4 CIRCULATING BLOWER 5 DEVICE FOR NEUTRON DENSITY MEASUREMENT IN WE REACTOR 6 NEUTRON DENSITY AUTOMATIC CON "ROLLER ... У CONTROL RODS ^f- SWCLANT OUTLET TEMPERATURE f*\ MEASURING DEVICE \J y TEMPERATURE CONTROLLER 10 COOLANT FLOWMETER •11 BLOCK ОТ POWER CORRECTION ACCORD;?!? TO FLOW RATE (3KP) •':' COOLANT FLOW RATE AL"• ОМАTiC CONTROL LER 1-1 SFR.'CMOJOR ", в:..-},:.r REGULATING VANES IT ••;••)*:/• LER OF THE PRESSURE . '.'. H.T S/£AM o í:.''>i stabilizes tb<* :atlet temperature in relation to г £.r-r:-'3*it ^(.nBtmnt value* It consists of a thermocouple mee- s.>.r.i.3g thR t:^s temperature in the cutlet pipings, a tempe T я' re r/ortr-lleT* BHTS and a neutron density л» rntrol cir c• .-±1 -\ The temperature controller is superior tc the cent» roller of nei;tr;?n density and its outlet- is a pre-set value f :r neutron density« 3u3*$4 Ci.v -:At f•-...r the correction of neutron density ai^-rding to the rate of coolant flow Í< rre-t:ing in desirable direction the reactor power in the case of a sudden and considerable decrease of the coding gas flow rate. It consists of a measuring device measuring the gas flow rate, and the BKP - 04 block the • oatlet of which is connected to the inlet of neutron den sity .vontrc-ller (comparing member). If the rate of co-ling gas ďvsreases,, tfcs BKP - 04,block transmits a correcting signal directly proportional to the rate of gas flow? if tLs rate cf flew stabilizes, then the correcting signal -.pprosr-h*is Z9T--- with a time~constant of about 120 s« The increase of the gas flow rate as well as its small fluctuations ere not responded by the BKP - 04 block /39/. Njvt^s Тле BKP =• 04 block lias been developed in the Huclear Pewar Plants Mviei^r of the SKODA Concern in 1968 and in-:-orp; ':-at*.l i^t.; th-э design cf the A-»1 nuclear power sta- t.b-.a (.ontrci system» That is why the present work does ii'-t descrlb* an earlier ASM block (for safety power decrease) m^nti*дч! ir. the introducing part of the Supplement 3, written in 1967« - 41 - ::,;«** Cir.tivL ^ " , т ^';л :о-.-1.1.п^ gas flow rate stsMl.:.;-•-- "•-• ' ÍV3 cf gas f 1 'w ti^*-. igh the reactor ^/í^rv t? + *i я;'-"-4-: M .us» Tr.-i.-3 at"? t#-- vlraaits of тч<"-.- л'-^т • •..•.*-::•••.'.. i.-.. • •••-- l.l«>-k '-i_*. f- ^ ,.::x iE flin power si-•:•"'•::.'••^.)f г -v-.-/v v ь-л-1 íÍTuuits f-'dnprisini; an BR - III ^,;s tím rafř •••:%.?••• ,:Her gc^ernsd by з eervcaotcr* Thy ч••.r-\::-a-.-J--:'T з--4'"- :*••? ^gulatiag -?a£es lb the circulating - •*-.. •:.'-. г &'•!.•/,•;% "•• • * mangirag the rat- vf gas flow» Seal .v."-..,i f_f t.':.-j r-f- of цаз flow behiuu tii< r-actcr is шваза- : ÍVC by a iiivHsariij< -levicea -.:-J~ f?r the aeaai-pressure steam з1аЪ'^;?вз t"r,4 p;---'Ss :?•'. of the mean-pressure steam before the f'^rbl.^ *» •:' -v"l".-; to a pre~set constant value. It cr-nsiets f-f а й^гд^вт '.---nt rolling th^ steam pressu re befcre the tir^in*? t -. KPI III controller of steam f pressure, and a ЛУГПГ <:r 'he control of the rate of ?-,oHng tias, There is о\н sa-h circuit in every block» The utput quantity от ^.\я pressure controller of the meavi- -pressure steam is *he given value for the rate of the tcllcg gaa fltw aroing on, to the input of both caatrvllwrs of the gas flow rate in the respective block. 3*3-6,. Centre! circuit fui* the control of pressure cf Ic.v--oressi;a,ft steam consists of a hydrs'ilii- - mechanical automatic contrcHsr of gtaam pressure aoting on controlling val- vas bti'-'Vt» the Low-pressar*? part of the turbine to mini mize t?.i*i f;.'5-;tap,tione ©f the prasea-re of the low-pressu- - 4a re steam in tbs a**** gn^-eat-iT (tane ais? minimizing tha ga« * 4ap%rat-.re on t.h^ s^nam generator safcT.etO and t-o a&- k-: p 3í Ъ'ч for all tfa* steam Ьс ba consumed by tne turbi ne Tn cráfT bt**- th-=» erat.?;! process woula not b* t?í- а л 1 Mga пи ->тп1 -а1 Гзч.^Ъаск between the circuit of She :'. -чг-ргвзвлгч $:;*1*ш »iot,tr :.I and the turbine revolutions r:-itt:?..-.l 'o'r-.att is mad** 3.3»7« Ci^íuit for turbine revolutions control а+?«Ъл1л'.-из th« numb-ir of T^v^lutťyas in relation to •-x g• vf;:: va"..'ir and лз stirl'v:? t that .i.sed in conventional pc*ar plants„ 3-3«8» In addition to the cir«*its mentioned above the re ara ether on^s on the A-l nuclear power station, for instant the water lev*i contrcl in the steam generator 1гдшав voltage control of electrical generators, et-г. The se circuits are similar to those used in conventional po wer stations and are not shown in fig* в* 3„4* The system of safety The A~i nuslear power station is equipped with a sa fety system for a qaiok and reliable shutdown of tha realtor and its emergency pooling in case that symptoms cf a dang^rrus a-í-ííd^nt are indicated. It consists of a number of sensors ar.d apparatuses serving as indicators ftf an ac-iident, a system of realtor ahat-desn in case of an accident» and a system ri reactor emergency cooling» - 43 3..4Л. Reactor shut—d --л. ?д* Bystem of reader aco.id*i-,t shut- - lioii :f station self-;:t,s.mpt.k:r; a^p;/'у due *•>* a) s»itching~off of the 220 kV switch i;-; the Kriř.rvany tijťitchjard, b> switr,hrr,g-^tf of rhs 220 kV switch in the A-l nuclear pswat- station, •:.} smt jhing-cf£ at th* rnr^ switches cf 220 kV gene- л-At „г-transformer blocks, '.!.; Bias Jit ane^us disintegration > t the 220 and 110 kV system, - decrease of the rate of the cooling gas flow below a pro-set value, the reactor power remaining unchanged (f'.>r simultaneous breakdown of two circulating blo wers), decrease of cooling gas pressure in the primary circuit below a pre-set value, • iacrease of steaa generator outlet gas temeperature о/яг a pre-set value, - increase of reactor outlet gas temperature over a pre- -set valae, - .fsuluey of both automatic controllers of reactor neu tírán density, ••• .ns.lf-.jnction of reactor control 5.ad safety system. "*" "f *T* The reactor may Ъе abu.&~-dcwn alít Ь- thri. operator by pressing the safety button. Таз aetive m^x^mva of the system foz rea^t^r shut -down a~s f-'.xD safety г:1s saep-»a-iad on electromagnets , í: c/er the reader • т ^), ani wV."-•? ompenseting rods* The rea*t~>:-: is shat-d:>wc Ъу fre* fall of all the safe ty r..;ds into the геаг-t^T' core (the fall *4ikas 0,9 s» }f and Ъу twelv» compensating rrds whi~h зге plugged in at a higher sp«Ňi than at normal rperat.l--»» (;.5 ;m/s.). J;i;3t after th-ч тна- t--r shutdown Ll\e power station S9if-»-:'-onsumpticn (;/л ••'.:;ding thi circulating blowers) is sappli*i by H v-.'^aa:- w. °r. falling frequency from the ronnxng-oit t .1гЪ•.-.„^n-irat: i s. IIÍ orier to moderate the thermal sho;k saused by slower decrease of the rate of gas flow thr-r-ugh the rea?-trr sompared to the reactor po wer decrease t the automation of the emergency cooling system ensures a qxU-Jc partial closing of the regulating vanes of th* blowers (by some two s„) and opening of the hydrauiical valve in the blower bypasa (approximate ly by 0.6 s.). 3*4.2. The self-consumption supply of the power station Schematic chart of the self-"consumption supply of the station is shown on fig. 9. The electric power of 3 тс 50 MW from the G, , G~. ani G, generators is transmit an т ted by the T, . T2» '^ з block transformers with ewitches on the 220 kV side, connected by a eimpln line with the switchyard in Sered. The eali - consuiaption is supplied by means of the T-1( ТОЙ and T2 transformers connected via the disconnecting switches to the с lampy of the G,, G~, and G» generators. Every transformer supplies two switch boards of self-consumption R6, designated as la, lb, 2a, 2\ 3.».., T;-» Езчт,у iilat-ri.bur.--r directly f*Gás -iiie synchron-- ПОЧЙ m'..'t..'?r '.-? ••:.rí-3a'Lí»T;ins blowers, Mhsr drivers, and se~ л vada:-"? ᱫfrib'-.t; л?. As * £:.':с:з^ *? jaad-b,? .for the R6 dis^^.b*..-:-!'-^ í-r-í tbe Te^iCV-lag: transformers T 18 and T "" •-••:xi^ -l^i -V- th-í л,0 kV L::.n« Trnava - Nové Měst?» the T г.. .Y»*g;Cat.i.ng г^зг -^ч-- ^ГЯИ^-ÍI t- the Madunic* hydroel-. • - - :r-:l.- p.-*-.' station fsrvl'ig as «. s^/md stand-by. Available :'aua%d.L.v.t-3l.7 is tťs first sta.nď-^y, the second one after startle the generater in Madur. ~e (3? s»)» In case of the ]<«з r.f ti-..- :--H:.t'«..uusaBipt.^ori s jpply due to a failure cf '•a* of th-jr T :.i, T 12, arid T 13 transformers, the stand- -Л-уая T "в, м-:--? Т 19 Ч-.-Ч .i.raaindiately available. 3.4,3. TV- art matj.o- f emergency cooling Th-; ьаъг§эг.оу cooling of the reactor after every a tion of safety rods in nominal operation implies the fallowing operations: - chcice '.'f thH scarce to provide the emergency cooling with the necessary power. This scarce will be either thy 110 kV lin* or, if it is not available, the hydro- •siH-^tric pcwsr station in Maiunlce, - swit-íhiríír í--ver of thy a--t(, generators of the secured supply t-..• Ъч í'3-i from the a! suraulators, - easaring th« хаю -ovt of the turbo-oompressors and tiu?bo--seijer&ti-4-s, - dis^utmectiziž the electrical appliances not necessary in vte res •*•>,: em^rgenr.y со-ling (through the use of a subordinate aat:ma^'c atard-by feeding CA2R) and the antmati-tz --t z*.r;-.~swi--.:.hing cff.), - etns:^risg tbr; seif'-ccnsrunption supply In C'-ncordance witb '-. 6г'-... p'.jgram of the emergency cooling (cgain - 46 in combinative with tn?. automat • зс«::Л->.у feeding =*r)d the automation :-f zerc-sw\ ~ ^ILW\ off.;.. *,44*4„ The em-^ger :y o:^.n^ program a) In the first stage - up t<; '-j? я aťbei the action ~f safety rc-is - the self-consumption is supplied by th* t. > b^-generatore which ^re running- -:-i~ , t} In th* s^- M"l stase fr^m 35 з to 2 minutes tha f- L- I^w:Lng parte are swi+":-.h^i -.v^r .n the stand-by sour ces s - on^ n^s.vy wat4r pump, - f-Xld С "_2Г Of рГ.'Ша'Гу .'ir^.Uitj -=• tw> ntann-preůoara p-uDs of steam generator, - two low-pressure р-чшрв of steam generator, - оде low-pressure feeding pomp... 1л .•,pe_'»ii--:íi are .inverse mot;rgenerators supplying the distributers of the secared feeding. c) In the rhird stage "beginning with the third minute, the following additional parts are also supplied from the stand-by ьоитг-и « one •: •.:>!.'.ng Р'лщр Л rhe heavy wat^c •icc-.r.-t, - one coojing water pump, •- one fan of prr.mary •"-iT*o.a.i.t сogling rowers, - some exhausting systems> - ona compressor "»f th«j gas - air irdxtara. d) LL the f ui-*-:b -t-sg-. - after 15 cvinutss - connecting ;-.f l*i39 гж.',"^ъ.::Л -»;:')':.^ллэ.з manually. - 4? Uj ^~ V Г; T a; a 5: H -^rэ(ЕН—- © -rv-j- «o S •o 5 • J o - • r> H to ^4v*t I* 3. s~ 4~. I Q ar — j-v.•••^}\Qj^^-^| s; The accident impulse closes also the fast working ga te of the mean- and low-pressure steam before the turbine. The steam generated in steam generators is then led via a reduction station into the condenaator. - 49 - 4. THE METHOD OF TII3 S3- -15Q REACTOR SAFETY ANAI/'^IS An prevail safety analysis of a realtor requires to kmw i*"s vfTiaai a. properties,, reliability characterise: г •:f :Ъъ íiiálvyvx*.: parts and apparatuses, character of fг- :n products release ani propagation, etc.. With the KS--i5-: г-ча t>т this re j -^rema.ot s s fj...filled only partially for Л'А •+• *..•; tj.=r •í-h-4T,a--.te"r,i3t;..'-.s repaired are known at present. Bi4 3.-.i^s,, a •''oBuparison w.:th ether *ypes f gas cooled reac« • лв ..$>-•*- "-lat the requirements concerning its safety ar« -ifv stj?:;.it wi.th the rea ter KS-150» It is evident that the •--« ia'-ts .Lufu-en.-ie the method of safety analysis adopted* 4-„3» Сопграгаз'П -f fo.e KS-150 reactor with other gas -cried rea -tors. Re Th*? probability of oc uren-a of any accident event is given in.the first pla-je by the reliability of its indiYi- d.ial parts and fan^ti^aa; components of the nuclear power etatlm under 1. avest.lgat.ton. That is the reason why it is Impossible ^т sp«*V aby at individual safety characteristics '• f differed vne. "-br-v *ypes and that is why no standard twists te compare different reactor types. But if in spite i}t this there j.s a strive to divide thermal reactors into type gr:aps with more or 1взя common safety characteris tics, then it will be suitable to divide them according to their :;oc?lantv i„ в* into reactors cooled by water. gasw organ!" material* etc» The reason of this division is the fa/"it that th« most dangerous accidents may Ъв cau sed by a failure of reactor cooling* Despite the impossibility to specify in detail even - 50 - tž. the fr-amsw у:Тл f W-. .;••, .__• '•'.•••: з^:Ч^7 • ;L.ii"ct.iv: s'- —•.- ••'«v',''.* г г -. Л gas •. -1:. .•"'-.•. .••=?. •••; -,-• -'.i,-:."LV U ť" •-- ^ *', di''"*A"4-í ťr*e HlA :. •; -*.-;' \j • *?> •:' V.;- г.ч.н.,1.. "5 ^i'1 - 1 - those hav.-'ne: а. •И^чт'зч <^f.M -.i S4fe''jy - iih ,яе w th ř. i4* . .'лЬ .-• xif .1-. л a on r--i?t--t.ir safety, - Sř.t..' nf4 ••4" • • 'v ••-(••• '^ . tv.'^'is. я." " 4'«...v.g >\f гъ* ^--v Xntb?n*.". pa^Sy diff-.-u". • v-;>.-, ' ;. " >-• ^'i^a-f •• "!•••>.i -f '^ ""•-.v-.ť,r after .its r sil>.--'.'. v.r.s arvi evtr. ОЛПЛ>.- •=* r-;-."-» " -^s f "ea - As a r-.;>n8ftq.jeZ'."e of -2 relatively low power density in the rea-v-tk-r cor* (.whi^h is especially true in the case of gx-aphit* rea-to-'A »' the •-'••arse of transient temperature phenomena is s„*<;wer. — oled reactors are not sensible to the cold f-?;>1i&t penetrated int.? the reactor (the problem of an *,í,'..viiant di« t ii-ld ccclant", serious fcr instance in lighi water reactors, does not exist). - 52 Reactor type Т?«г*ате£э2 Dimension Magnox AG* BL-4 •V.;.-Hag gas с -•p t ap to pressure kp /ош 60 i со s не 25 ; some 40 Spft'i^ic fuell .•^eraa.:- l-:a'iiagj kW/kg 10-12 16.5 ?-< ; average --/ал.*'-*.! iasti'lffivJU i Э SlgK. 4-50 800 600 500 Material of jatainless stainless fael -ana 1 иткпох I steel steel Mg-i-.s UO, UO. aramům Pu-Jl jaetal metal !Tj.m-i constant s 'of fuel Í 8,7 b.4 4.3 'Г:1,.ша constant ; ;o.f moderator I 1 600 900 / 100/ meI^i ng corrosion corrosion '••JP fu - 53 ., T g- + я _, •.-..г-.*--1 ^ _•?•* ..«sip;.ut it vňll be suitable ~. уцр-ч.-^ з m- m~ ;-ч аа^г--"- f rb-* KS~1;">0 reactor with •~и.Ж-- •;•-.--.••''•'•• •- ' t':rt Мт.а;.'-: •-^ . «lift. *%"»'-? SL--4 tyoe (fcl- * Ti- " - . -, лЬ. •<*•- r/aat .:.f *\'i »h* typ-.e .--.-aside red t>v= КЙ-111 ."f- " • ':•-.= *hd hlgr.^st coolant ргев^хгт,, *~he his^ íiiíj *•'-, •.;,-• '. ; -.-•;;. >.ьг тг tb.s fi'-:I, the shortest tiara п--...ок -•-.•••": - ^ -•'" :-•-'" •'••' eri-i ЧЛ^ smallest ''reserve" .in the ^ ,-:. ч\ - ' 2 ~"1п Safety pvlLt Of Via», it Ъб "••ffnSfc --.. ..i'--j-.n^s on j-rfliability of the A~l. c ». ,.. „• ."> - t\-:: , mp-^-i.ts *ivi on the» system of re-v. tot' saf-j•.';•• -л! "-::4-!••-•: 1 IBUUV 'be mor-з stringent. *T*CÍ4 ',JA.~í i-'- • u- .* -л \ \ Л ,/S - С Á, .i -a-:;.-.* lu.: :• v .„_ы; ^'v- ocnseqiencss on any reactor rmy "v* '•'-. ~-v >; ~ l tv ^ 'i • ^su'il. of events, the succession ?•* -»h:'."- s -;. У:- ••;: f я-.,. 1Э, At the very beginning there 1-s HI. '::,it...-t' •-•.я *-•••;эгД --v.a^i by the failure of a part, or Ъ/ a т,..от 4S-. : tíi-ч -píítat.r, í.r by an external disturbing H'jtioia. A;-; -. 'jvsb-r^zc* of the irjitiating event, excursions ••>•** T^a'.-tiv-tty, p-*er5 .::.!=mt pressure, fuel temperature, -•;.r- ?:f •jb.:;: -*-;^-.;;•• г papejme t^rs develop, It is true that the -HT-:.*rft-. гл ш-.у V; llnu-st by t-Ъв intrinsic mechanism of the x-aa.t T ("i.-v- iii-št a.n•-•*» Ъv f-he negative temperature coefficient of т>^н.. vj.^i^Ч v *..n -p-ra+ViT.. If she exc-irs.ion is not sufficiently úmit-íi,, цг. ,£••, -i ;.-,-+ - tne reactor follows, which may ra- o.:i..t i.r tar- eh.it': ,;.^.-.v--Wil oi realtor oparation or in an еплаэд*-? or '!'.-< р^:?з .••v-ii.j.. lis adiltion, xi the containment •:>:> i r^-. - ' •:;•--•--T-? -f;?ifецдаМ systems fail t;, reduce the ! ; :-:-лЗ-!.г. -.JSV ^- -v raq '.;,••/• HI vi-?at (.systems of containment ';.'.'•'. '.'••. '"'•'- ••'"'' +'::-з acc-j .Uiit atay iaad to a radiation »i. , -L - 5^ Fitr. Ю. SCHEMATIC CHART OF THE COURSE OF REACTOR ACCIDENT EVENT AND THE DEFINITION OF THE REGION OF ANALYSIS В EXCURSION EXCURSION LIMITATION OF INITIATING LIMITATION BY LIMITATION BY THE EXCURSION EVENT THE REACTOR THE SYSTEM „MANUALLY" INTRINSIC OF SAFETY MECHANISM J POSTULATE I ANALYSIS I ANALYSIS I ANALYSIS OF Л ESTIMATION OF THE Л OPTIMIZATION OF POSŠIDILI TIES OCCURENCE THE SYSTEM PROBABILITY OF SAFETY .CONTAINMENT AND DEVICES TO DEDUCE /INFLUENCE ON \ ISOLATING REACTOR REACTOR ACCIDENT \ThE SURROUNDINGS j<]- DISTANCE ACCIDENT \ / CONSEQUENCES ч _ **• f INFLUENCE ON \ i-1 f REACTOR L | THE PEŘSONELL Ю" -Oi c«HUTi DOWN V у / Let os have a look on fig. 10. The upper line of events "An up to NDn demarcates the region of accident prevention. The occurence and the course of events from "A" to **СИ depends mainly on the quality of the design, the reliability of the components, and the proper func tion of the system of safety. The course of the "D" event- as well as the initiation and course of the event "Iй is dependent on the qualification of the persona 11, the stan dard of the operational inspections, etc. Simply speakingř the accident prevention depends on tha quality of both the engineering details and the object operation. The choice of the method of the KS-150 reactor ana lysis is further influenced by the following special cir cumstances: - Concerning both conception and parameters the IB-150 reactor type is the only one in the world, from this follows that taking-over of foreign experience is very limited кthe reactor is not even verified on a smaller power unit). - The 1-1 nuclear power station consists mainly from ope rationally unverified parts or components, which means that the quantitative characteristics of reliability ne cessary for a systematic analysis based on probability are not available. A probability-based approach may be made in practice only with certain schematic charts of the control and safety system, supply systems, etc., composed of current electrical components with known re liability characteristics. In addition, usual ways of reliability increase С stand-by, etc.) are used here. - The characteristics of the fission products release from the fuel element in case of an extreme state are not known. This makes impossible a quantitative analysis of - 56 - í>sdag t.í thes* '?irc.Lis*tar.-:.4s tet* author h-нз adopt"< - алз. ««e^'ifciallLy $**Ы.М5:1те лз&ргоаг.г.. ti- -7.1s analysis of «*^i ia&t "taM*^ ,,alíi'. f.-n-Mu the stnndp •'.>Л vf *-чс1$*- р*?ат-*;ьЪ£•*.-* ежЪза* vao. t?pi---&l Ъ;у the fa!?.' sw-ítg m^+far-dic - pr«e«'íxT« v*»*« «lee. fig. 10/: - 2д* »i*r#ms Initiating 9T*ats sra postulated ííbvie^ej. w wilri к. "s*?y I*w9 bat а тпя.1 probability of оесц?%язв i. vtmv :? tb.it r^aH-cr LLft-tla*), - Uh* transient phenomena i* the nacl*«-.r р;*гег station *аазз«1 by said extreme initiating events агз analyzes, tin r-ea-*teí safety system being verified and optimaliz s i fs©*. ti? е^ч'з.гр i:'", ••if the schematic $аатф and the sat» ^•iag if t&a *tať«*tlc controllers, The possibility cf rear"-t;e sura*:* system delay er its failure is also ana lysa*. ~ Тлч possibilities of шап&з.! liaitatioa of ascident exeur- sioas are analysed (in the first place from the viewpoint ©f time). - On tb.3 basis of the results of the analysis suitable changes of arrangement with the aim ef increasing reac tor safety are proposed and dismissed, - To e&*-- в i«aatltative idea of the possibility ef a serious accident iaaitiatiU», the order of the probability ©f oc- ourenss ©f eattreme initiating events is evaluated. 4*' . Sxt.?«me initiating event8 The postulated extreme initiating events are selected rz th* basis of a real probability ef thuir ooaartnee du- 7)1щ *'b.* k*+\ suisie&r po»wer station life-»tiae and of the сош- р«г!з.'Ъ wt*h the safety analysis of fere-ign reactors. This - 5? - • «^.'...a^iis way the iallure of the pressure "sesel du* to bri**•."„•-* fraktura is ast postulated (s&e also Chapter "'• Ti« f-llowlng initiating events er<* postulated; - Failure of tha neutron density automatic i^ntrolier, consequence of which is an uncontrolled withdrawal et b^th the control rods at the maximum soaed from the _••_ re if the reactor under operation. - Failure of one or two circulating Ы cw-nrs» the otó^i' .-cm -.?*aainilag in operation. - Xi.:v. *. •• " \..«.'iV"' station salf -censumpti k;^ supply foil ewe: •y-y i. fast decrease of the rate of coolant flow through tae reactor. - Ruptu?-» of one of the twelve main gas pipings of the primary circuit. 4*4* Mathematical models For the A-l power station accident states analysis three mathematical modele ere developed, one being for the КЗ 150 reactor dynamics, and the other two are special mo dels for the dynamics of the power station primary circuit. In the course of their setting a special attention has been devoted in the first plaee to the precise description úf dynamical modele of heat transfer and coolant flow pro cesses* System of basic eo.uatieas used for the unsteady heat transfer aad gas flowing is derived in Supplement 1* Thy relationships derived are valid for the extent of gas flow rate in the primary cireuit from 100 to some 66% of th? nominal value, and for temperatures op to 650° C. For analysis made in the framewerk ef this report these regions a*?** sufficient (the gravitational member in the momentum - 58 - equation» ae well as the amission component in the equ ation of energy, are neglected). Supplement 2 presents с comparison of the relationships derived with the courses of heat transfer measured on the 1/5 model of a part of the primary piping. The results are in very good concor dance» Supplements 3 and 5 present the fallowing mathemati cal models serving as calculating programs; 4.4..1. The A 1 HAZARD program (Supplement 3) The KS-I50 realtor mathematical model with internal temperature feedbacks and with control and safety system. The reactor core is represented by a point neutron source and by three-dimensional source of heat. The system of control comprises the circuit of gas outlet temperature control, the circuit of power correction (based on the ra se of gas flow), and the circuit of neutron density con trol. The safety system consists of the block of accident warning (acting on exceeding power, outlet gas temperature, and reactor period), and the active members, i. e. four sa fety and twelve compensating reds. The input disturbing parameters of the model are the changes of reactivity, the rate of gas flow, and the inlet gas temperature. The A-1 HAZARD program can be used for the analysis of safety in the range from 100 to approximately 10% of the nominal rate of gas flo /. the changes of gas pres sure being snail* Programming language is the BLLIOT- -AUTOKOD. - 59 - 4„4i2i The A-4.. SHOOK p-ог^ча (Supplement 40 A mathematical modal of the A-l power station pri- лаэгу sir-'jait is t ;• звг'гч in the study of transient phe- югяа*за da* tr% a failure of ai3? -ulating blowers« The reac~ tor is modelled with the coaling c.iretails of the side thermal shi-?ii3, ths ealaniria supporting system, and of the lower biological shield,, The c 4»4;3» The A-l AEPORT program (Supplement 5) An isothermal model of the A-l power station prima ry i.ir>u.:-t for tive investigation of the transient pheno mena of gae pressure owing to a raptare of the primary piping-, ТЬ-í r?a i-or is modelled with the coding circuits ůf the si-i* thermal sM elding9 of the space below calan- dria» ajuu -f the lower biological shield, the reactor co re being assumed to be a caraulative hydraulic resistance. The primary Tl?-,uit model consists of two loops, the first including the piping raptured, the second one comprising all the piping branches undamaged. Both the loops are made in tw variants,, the first one in concordance with the A-l nx::\.А%£ power station d^sig.n9 the second one with the thr:-t*-;Uag elements built into the piping. The program is solved on the APyd analogue computer. 60 5* THE KH'50 Rb'AC-yOR ^STT 1ЖШТ.СДЛ, CHAKACÍESIIISTTHS TJV ^х-'дл-я!.:'?^ ,л;г т^а" :.• pt-u-aiuetqrs -vi* t;4 an init;"-- tirwc ыт*- if.'.n-. аиу Г»А •••*чг\:\:э1 thr-usdb the iatrixuií.' л *^VM1-'- -TÍ^ •• т w\ '•- jn-..-».-!-} -•>' Ь.з control aad safety -• - тл--:-г* '?i'.*? sp*s4 ť* "-^-d Я*--. pnv:t >f на -rrcirsian a.~ Жч . я.€- tb.-í :'j>.í"',a-í:;;í* \f iN*dba'It vf.f *-.ts dep^i considerably .\7». *í.-i<í tjp-. -f the ws-'t.-r, The KS-»I^O гч^.-.^лр^ or the A-A. дл>,1*яг р15»?)1 sř-^irn, áre nc^ad for the following sp-»>1íi ^•fi'VSE having чъ. .l*ať"i.'is.a-=.4 02: th* dynamise of the ofe;J* ť - '**'í.;-.vc..7-*r..;/ s_ar.I.V f \-v.\ irařat^r-y in th; realtor, high va~ 1:чъ ?Z Ъ*&Ъ ^va-as**? -v-.-^f I ^- - 6.1 21#v9 the temperature diiT^j orof or i;h-\ ;ař.ctj.v reEtHinir cuichang**.» Th.4 role i«f this coatvol syst*^ _n reducla^ řč<*iďiK:t «xcvirsiťrAS «í tlv* KB-1C'---- .ti-^iivi.s.-j. baa rot Ъв^л ;• , •1 *\ in datail s* <Ч'?«. 5*1* "Ev* vffoct Cͱ' iDtrinsjLr f^.* i^a-Y-- Sa.^-lť nror-?ítif>aal characteristics cť t\-. KS-150 ru«c- r;••>.!• •í-'it>. -lis--;-. :_:-,•?i-tM ctfitroi a::-1 яч'чЪу circuits fcr ťssllu :vs dus t-- а с1-\з.<%* of rss^ti^ ;v «-.vi .Г th-э rata ...f -зос-Пщ řfVts fl-:.•**9 US tícr:, £/$.,; = 0.0068, and V. /• л • ..,• O.0034 are смзай*»- ťoij th-ч first ošil bf»ía§ YHIÍÍ f->r f:'=»3h fu»*l (natural lira- ni-ш), ti-•i F^coiid опэ for fu-íl uniformly Irradiated (avera ge f i**i irradiation attainable as ths maximum in tfcw KS-l^O r-ía«í^:;r '.v?,r"r a.Vranc-*! ?u*l fcleseats is supposed to Ъе seme 4 400 MWd/tj f'v^ :ь,г<г* detailed Information sea Supplement 3)* ТЫ m.líi^t.г *-аг:.р*г--iter* oaťficiest of reactivity хз лзюапиД +..í *••* -T-Ojj : - 2.10^ С(НД' /40/, and, according to /41/ř tfcí? ful teaiporatvír* •--'íffloienf. si' reactivity - i ,^1з ГОу - .- 3*5 . 10~5 V*1* СЫск calculation for a íWíl witfc. an average irradiation of about 4 400 lliilfd/t has giv*2i a little lower та!«лз, TC^ ^ --• 2 • 10~5 CCT1 /ref.42/. F#.^« 11 pre>33rto the reactor transient characteristics in a sase of a failure due ti- a chan&a of reactivity r*. = - + чО '•' ú-й. a i) jniinal p-'.war la val. If there is но effect of intrluHl'', Се «.г.--дека гЬч ъи,:-.и*:вИ'.;П of the averaged gas teai- V4Vfitixv: ••Ú. fun! charmeltí 5.--itlet ." „,„.. 'v-^ald rise with a lUC Hit B-»evri .v? i5 - 20'' 0/s, and tl.4 *А,-:**л rf the excursion of ura&iun plia ллх1пкя! t?.mv~vzt <*!•'•< t ^ few „._,_, would lie in t\& гащ* from 24 - 5-2'"' O/з, Tha tampttrat-ira feedback cf *'!:.* mod*rater fc^s, roughly spiking,, m;< effect r-ii the tran- ale at, рЬ«я'.ога^пг. Ccncernliug the temperature feedback of the 62 fuel, it is relatively W-AV aid permit3 excursions of fue* pin temperatuva in ^TOSSS of 30° 0 (far A anifomly irra diated realtor core)» Similar sharaoteristics for a 10% power level are ahem r.n fig-* 12* ±z> :4>mparison with the eitinal power sir. temperature axi^irsio-ns of uranium pins and of the gas or.» the f lei channels cutlet weald rise at a speed six times lower. Fí.g. 13 Bhows sjjailar transient characteristics of t.»ie KS-150 realtor for a failure due to a 10% change of the rate of cooling gas flow at a nominal power level» The •sxetirsicns of the outlet eras temperature ú t..„ have an mean initial speeot of about 9' C/s, while those for uranium pin temper-atar»* / tn some 16° C/s. Owing to the temperaturs feedback the maximum excursion of uranium temperature will be decreased from 37 to some 17° C. Similar characteristics for a 10% power level are shown on fig* 14. Compared with the nominal power level they have a speed approximately 3ix times lower, 5;,2„ The effect of the control system Figures 15 through 18 present the control characteris tic, of the reactor explaining the role of individual cir cuits of the control system in reducing the excursions of reactor power and the temperature of both uranium and the cooling gas caused by changes of the reactivity or of the gas flow rate.(The characteristics have been calculated by the A~l HAZARD program ). Fig* 15 shows the influence of neutron density control circuit on the excursions caused by the change of reactivi- ty •> ~ 10 • Comparing with analogous courses without the effect of the control (fig* 11) explains that due to the action cf neutron contrc?. circuit the щатг!тнтп excursion of gas temperature diminishes approximately three times and Cemperafc-ire of uranium pins approximately tare tines (for - 63 - lower pcwar levels the transient excursions are conside rably smaller). Reactor control characteristic during the failure cau sed by a change of rate of gas flow is drawn on fig. 16. It may be derived from the foregoing results that the circuit of neutron density control is capable of reducing quickly and efficiently enough the excursions of power o- wing to sudden and relatively considerable reactivity chan- ges кйк к 10 y corresponds to 8% of the reactivity bound in both the control ь.Лв of the KS-150 reactor). Fig. 17 and 18 show how the circuits of gas tempera ture control and power correction (based on the rate of gas flow) effect the excursions of temperature caused by a sudden change of the gas flow rate. The automatic controller of temperature is presented here as the PI - controller (the BRT type automatic con troller with power selector operating in a pulsating re gime - i. e. in small input failures - may be substituted approximately by a PI type automatic controller /43/) ths transfer of which has a shape ЯСИ,- (r)J ^ _ 'Kr.' Kj'T (1) 4r* oi L j. 'г; г туг :• f ?r • \- r j The block of power correction based on the rate of coolant flow is presented аз the PD member with a transfer in the shape (it is one of the above given variants): - 64 - It becomes evident from figures 1? and 18 that the temperature control circuit alone limits very weakly the temperature excursions of uranium pins and of the gas caused by a sudden change gas flow rate* It is explainer'' by the /act that this circuit is destine to c;>ntrui thó slower temperature changes encountered in tta i-urse of operational changes of the nuclear power station i cad, Tfc. effect of intrinsic feedbacks is eliminated by ths fast- -vorking automatic controller of neutron density winch ac™ iptably stabilizes the reactor power on a level given by the automatic controller of gas temperature iY-,mpare cor-- '•-3 1 and 2 on fig„ 17 and curve 1 on fig. 16). The cir- ••>.-t r,f reactor power correction based on the rate of gas f.>, -w the output of which is connected to the i^.put ;-f natron density automatic controller, effectively limits the temperature excursione caused by sudden changes of the rate of gas flow (curve 3» fig о 17)» Fig» 1& shows similar ť.íontrol characteristics for a 33% reactor power level. The process of control has a slower course and its stabilisa tion takes a longer time, thus indicating the possibility of unstable control for smaller power levels (a more de tailed information concerning the conditions of unstable w:r"K of the A-l nuclear power station system of control may be found, for instance, in /44/). 5e3«> Summary 5-ЗЛ* The changes of temperature in KS-150 reactor active core at a nominal power level, caused by a change of reactivity or of the rate of cooling gas flow have unusually fast course (the time constant of uranium pin temperature change is approximately 1»,3 -2 s«9 and that of gas temperature change on - 65 - 1 the outlet of fuel channels is some 3 a*)« 5„3.2a The intrinsic feedbacks of the KS-150 reactor ary w*ak and uncapable of reducing effectively the ex cursions ox power and temperature. 5*3#3# The circuit of reactor neutron density control Is capable to reduce effectively the power excursion,' caused by a sudden increase of reactivity of about + 1(Г3. 5-.З.Л» Th^. ir-jit r,f gas temperature control on the reac tor outlet is not capable to reduce effectively the -.i-.'arsions of temperature due to sudden chan ges cf the rata of gas flow. Such excursions may be limited substantially by the circuit of power •:;;rreetion based on the rate of gas flow, acting directly on the iriput of neutron density automatic controller. 5*3.5. To reduce effectively the power and temperature excursions in case of sudden changes of reactivi ty or of the rate of gas floy manually by the ope rator is not applicable in practice, because the speed cf these phenomena is too high* - 66 Fýtt-- KS -150 REACTOR TRANKENT CHARACTERISTICS. FAILURE DUE TO A REACTIVITY CHANGE ék4Ó*0N THE NOMINAL POWER LEVEL (WITH CONTROL ViO SAFETY CIRCUITS OiSCUNNECTEO) CURVE I/»/ *.M «•W .- 1 0,0054 0 0 2 0,0054 0 -I. «0* 3 0,001» 0 0 4 0.005$ 0 -*.»* «»«w/V 5 0,0054 -iV -f.rf* * 0J0O55 4iV i»4 'W * CURVE re I* K.M 1 1 0,0054 0 2 0,006» 0 c 3 0,0054 -*.»• ~?.t: 4 0,006» •3,5. Ю* •2.К *U~/r>M í Ю 12 К « 7-'*, -? fW 0.Э5 o,sc s- 1 1 >.'?.•'W£ Iff/ reu [ť] t 0,005* 0 * i .4 0,0054 0 ; - í. ir. i 0,0060 0 0 4 0,0066 0 -2ЛС 2 0,0054 -г .ю -*•* 1 1 3 0,0054 -3,5. Ю* r/-,; ZI.SJ 6 в Ю 12 r: : FHJM. KS-150 REACTOR TR;.NSIE»T CHAPA FR S ГГ FAILURE DUS Tu А ЮХ CHANGE С? THE GAS FLOW* R... '" OS TÍJC IQX POWER LEVEi (ч/пн CONTROL AND L;.4»-£"TY ZKC.HTS L?':,CTV.VI "••"•' 0 2 < s e Ю 12 4 H Jfl *[*] 0(0) 0,55 0,30 • iW N(0) 1 •'! CURVE: X». KuM ; QCK)5< 1 0,006* ° -« 0,0054 -2. Ю1 г - г. ÍC j 0,006» -J,5. Ю* -*5.Ю4 *w K,~.<*>M ^«; По rš KS-150 REАС Г 0* CONTROL CHARACTERISTICS. FA/LURt: DUE TO A R&.ACT!\"TY ChAV-SF 6k -10*ON THE. NCWNAi. AND THE IC'i POWER LC\>EL •'•;•".•- ri+. •rppzwjRE СО*!>'Л I;.:.Í,ŮNN£CT£D} í* í-r' *M ю 12 п-Л; ;.1C N(O) 1 CURVE Nnam 10V тс„№ t 1.0 3,0066 0 0 2 1,0 0,0068 -3,5. Ю* -2.Ю* 3 0,1 0,006» 0 0 W r t КЗ 150 REACTOR CQ^T^OL CHAMCiČRC?': IN СЛ5Е OF A FfilLUR£ DUE T 4 )0*4 CHANGE OF THE С-.Л5 FLOW ИАк: 'Л' Л'*Г \!J4№L AUD W "4. POWER LEVEL (with the temperature control circuit d's^onnertodj 8 Ю 10 \ 0,95 0,90 ( CURVE \JL °L X6; TCj'C"] тсм [x-<] t^nOrr, S k 1,0 Ot 0068 -3,5. W' -2. )0' 5 2 0.1 0,0068 -b,5. Ю' '2.10'* Fiq 1? HS-150 REACTOP -ZOHTRDL Z:ÍAKAC!'ZRiShi:: /v C--S-z Qf DUE ТО А ЮК CHANGE OF THE GЛ7 Я.:?*' Ms7£ .•:.. .-y..£ . POWER LEVEL Ю 20 Ю ÍO '.o H1U ví OJ 0,95 090 Fifi8-. KS-150 REACTOR CONTROL CHARACTERISTICS IN THE CASE OF А 10 X CHANGE OF THE GAS FLOW RATE ON THE 33% POWER LEVEL 0 20 40 SO ow •- vM ~3ic) Q90 vw ',o q» Q90 Aim WM X) K> •• -10 1 WXK OF ажитюм CURVE par* l* "'Щ M M *,N :М ьЫ K.H v/V Т.Ы 1 0,0054 -3fi.ň* 0 V 0,137 0 - — - г 0,0054 Sfiló' 0 V 0,137 0 V> sp $fl ш«ы *»«>"*• . WS»** " t .ft 6. ANALYSIS OF THB KS-150 REACTOK ACCIDENT EVENTS Uadsr investigation are the accident states caused Ъу the following events? - failure of the neutron density automatic controller ая consequence of whir^h both the control rode are withdraw- at the utmost speed out of the reactor core, with no control of this processj - operational breakdown of one or two main circulating blo wers $ - loss of self-consumption supply of the power station; - rupture of one of the twelve main pipings of the prima ry circuit. The pniwer and temperature excursions of the reactor are investigated with special emphasis to the excursions of fuel elements maximum temperature* The effect of the system of control and safety to reduce the dangerous acci dent excursions is analyzed as well. According to the de sign specification the safety system should shut-down the reactor provided its power exceeds the given value by 17%» when the reactor period Tpggtž 20 в., or if the cooling gas temperatare behind the reactor exceeds the pre-set va lue by 20° C. Short excursions of fuel element temperatu re should not rise more than + 50° С over the steady va lue* 6#1. Uncontrolled withdrawal of control rods out of the reactor cere The coarse of events in this accident state is assumed according to fig. 19* The initiating event is an unidenti- - 75 - ?:.<7).t fe. : агб ;f the neutron density control circuit lea dens? t--> an uncontrolled movement of both the control i/olc *t th3 utw-at eps«d either downward or upward up *o their tormina."- p--s::t;':oa. Their starting position s(o) is ass isi^d t;- l.ť? t^ the middle of the reactor core, where thM- movement :..<= fch* most effective (at the "beginning). A d^wnwa-i'Ly directed uncontrolled movement of the •.v>£t •?.:•! r. is 1ч not dangerous, for it results in a decrease f г3.-4•.-•••: v- p..*wer; see fig, 19, sejj^uence 1-3» The withdra wn". <;f :>•- 1 rods out of the reactor core (1 - 2) leads t;:> an is raase :f both reactor power and temperature of a^aniats and со ling gas» The reactor safety system should shut-down the reat r in case of an excursion Щр reactor power _ 76 - lure of the active members of reactor safety system should <-.>зог (assuming *he automatic systems 3f šm*rg9nc.y cooli u.;- are in operation), the sequence of events would be (1 - £ - 4 - ?). Figures 20 through 24 present the results of ав ana; sis e.f the H^idHnt event under investigation, as ca2. .--u •- ted Ъу the A«l HAZABD progs-am (KOR). Pig. 20 shows the courses of rea'-tcv parameters in * case nt an uncontrolled withdrawal c-f. control r-*ds out of the rsastsr i*.ore9 the power level baing assumed to be no minal* Number i denotes fresh fuel, number 2 represents the f ля] onif .irmly irradiated» It is clear that reactor power weald вх?-з*& i+г nominal level by some 50 - 100% with cor- responding exceeding of fael pin temperature by more than 200° C, if there w:>uld be no action of the safety system, Carres 1* Id OR fig* 20 show the influence of safe ty system set in action by a signal indicating that the reactor p^wer exceeds its nominal level by 5 - 20%, Similar ly, c;.rvAs 2a - 2d are valid for a delay of the power evi dent signal лТ £QQ - ^ s. The dependency of the maximum ex- coding of fael pin temperature on the magnitude of acci- dental power excursion*^ and on the delay of signal ^*ACC se1; on t:i tiie ea^e*7 syatea is shown on fig. 21. It be^mes evMer-.t that, if the permissible exceeding of fa-el pin temper at-are by Д t« » 50° С is assumed, then tin d*»3ignei adjustment of the safety system on £ N.„Q 17% is a^-fptahl* even, in the case of accident signal de lay of аЪт-it 0*5 a^if any. Carvss l,e and 2e *--n fig. 20 demonstrate how ineffecti ve is the reartcT shit^iown if based on the exceeding of thy :<;'-.r°,tig gas temperature (due to a relatively great dis- t.v&s-a h 77 - well as da^to th« ti.*zrn;.: jupl*i thermal i^ai*4 v* 4 Fig. 22 ^omparss tha ^"Icienoy „vf accident s-gn>.u- ling ia-iacai by ^-art;? p:"*e-- and рзпМ on the nominal o-rva r LíVví.» T:.i -«v-s-efc .p-i.cl:-^ г-a la* on period meter :••?•' у it lies in tria vicinit," c£ Ciis boundary oř the accid^n" "•'':*.!.i4 T „„ ; 20 s., Ť.ver. if th-з minimam of the period a-.- - i-v.-.l ••;?.!.'•;.-. :1 Й чхг.зтап^ХаНу l~wer„ Compared to a signal ia-A-iv-. ':•;••• ."'aast.:- power, reactor shut-down by a signal i.;'om reactcr period wcili result in an exceeding of fas I pin taiapsratura approximately two times lower, A s•;;•-"'..: .л, -7-*r. :f vary improbable accident oase woul^ :c. ;._ ÍŤ' •".'.'::•'•*?; "-:• • \:<. il r.;.l^ л wi^iErawv. <-. Fis:.. 24 presents the reactor p&rameters during the amort i l;-a 'b->t if :s,.t would be cooled by 10% rate of gaB flpw;, the temperature of a part; of the fuel would ex- •«e*.i ^УУ Q U--xx'7<ň 2). Active safety system based on reac- - ?e ~ tor period would stop the realtor approximately by 1 s ¥ while safety system based on gas temperature is practical ly {ineffective. 6#1.1. Summary 6.1.1.1. Failure of reactor safety system during an uncor.' rolled withdrawal of control rods out of the гад: tor core at the nominal powe? level would cause serious accident owing to the melting of the magor part of fuel elements. 6.1*1.2. In this case the reactor may be efficiently shut -down by safety system set into action by a signal resulting from power exceeding. Safety system set ting to А ЯдсС " *^ doe6 not Permi,b *Ъя ^el Pix*s to be overheated more than by 50° С {A tg ию^0° С„ and the accident signal delay on the signalling li nes is not allowed to be higher than 0»5 s.). Safe ty system based on reactor period may prove to be ineffective at the nominal power level. Safety sys tem based on gas temperature is not effective owing to. the temperature signal delay due to looafrio» and thermal inertia of the thermocouple/. 6.1.1.3. Failure of reactor safety system during an uncontrol led reactor starting at a very low power layel (of about 0*1% of its nominal value) would not lead, to the overheating of the fuel provide*!, the reaotor is cooled by full rate of gas flow. 6.1.1.4-. At low power levels under the accident situation considered the most efficient safety system is that based on reactor period. Safety system based on gas temperature is practically ineffective. -79- б«1Л«5* Brief evaluation of consequences of events accor ding to fig. 19. fable 5 The еаагм of a vest a ConaeejieBoes (fig. 19) (1-2-4-6) Melting of the moat part of the (1-2-4-7) fuel in the reactor core Beactor is efficiently shut-down; the maximum overkaatiag of fael (1-2-5-8) elemente for Д *ACC = 17% is A tU„, пах < -*50v ° С Efficient shut-down ef the reactor, mainlj on low power levels; (1-2-5-9) overheating of fuel elements for TACC . 20 a. is А ^#ввх < 20° С Late shut-down offent reacto r - (1 - 2 - 5 - 10) melting of a part of the fuel la the reactor core (1 - 5) Reactor power decreases - 80 - "SE COURSE OF tVENlS AS ALSLMLUЛ/С7 ;N "St C/.SE OF~A FAIL UŘE OF THE NEUTRON DENSITY CONTROL CiRCUJT WITH SUBSEQUENT UNCONTROLLED MOVEMENT OF CONTROL RODS СМОЛ/-7 .S В LOW ~ I: '•'••7 (6) ГА;С.:® OF DlSCONUCT.i. r~ REACT OP 1 3AFF.TY SY'-JE -' U) \ CiRCULXr AC BLOW ERA. (T) DISCONNECTED UNCONTROLLED WITHDRAWAL OF CONTROL PONS /?FACTOR '•>•:'.'T- OUT OF THE DOWN BY A C;r.- REACTOR CORE (2) NAL iNL.O-ri;. -• (8) EXCEEDING 1 UNCONTROLLED PLUGGING CF CONTROl PCnS \(Ю: It-: TO THE REACTOR :oh~ !fj; Fif 20 THE COURSES OF KS-150 REACTOR PARAMETERS DUE TO AN UNCONTROLLED WITHDRAWAL Oh THE CONTRa. RODS OUT OF THE REACTOR CORE (FOR NOMINAL POWER LB/EL) Occident signals induced by po»er |M temperature TW *[*] ————7 CURVE SAFETY SYSTEM TC fČ'] l*t u AŇ^M ***N *vw N 1 — — • - ±100 la 5 20 0 3 lb Ю 20 0 '.Э 0 0068 -3,5. K1 1c » 20 o. 31 Id 20 20 0 tí • 1e — 20 Í5* 1 0 г — — — >300 i 2a 5 20 ' to a 2b Ю 20 '**«Л* 59 0,005* -2.Ю1 2c a 20 ifl ж 34 20 20 ifi r? 2» — 20 0 22? etrWffl .; */V f,a.21- UNCONTROLLED WITHDRAWAL OF CONTROL RODS.DEPENDENCE OF MAXIMUM OVERHEATING OF FUEL CANS ON THE ADJuS^'NT л NACC AND ON THE DELAY A tACC OF THE SIGNAL SENT ON WE SAFE ТУ SYSTEM (NOMINAL POWER LEVEL) FOR FRESH FUEL 1 lTp£*0,0OS8; TCU- 3,5.10' X" ) FOR FUEL UNIFORMLY IRRADIATED Zp =0,DO54, ТС„=-2.10'ТГ)-i Ч-гЛ 100 Г At»mJCj 50 0 15 (1?) ! с j^ '"ni ••• с ~ -?50 ••ÍÍT/ÍÍ"-/?^ г4г>^ '£'Гл?5 Obr ! or r0 .-i/V FD 7r;ř.iiRAi'/- . IF ^0N7 :L Pfi' • \-,ER Oi? 9У к::/. ' Or: 75 2 / / /" ?:/" 75 1 1 n ... Tr 5.4 "Г/ SYST-.: "" 7 V\ i vwv *^ #*.;& li \ ^ --TI — w- 77 F í Q0C68 •3.510'* 0 . \ . \ i 20 ч -i ~ \ 35 2 O.D05b -2Ю ' -——• 0 !1 20 1 x M 4- •сс.;:г'>т PER,o.i VALUE .20. •is кет MEASURE'-. il/ 4> v/ ACTUAL VALUE OF JE АС! Г* <.1L I '• / < '•/ rt:R:OD VALUE ON THE PEP • f,nfc9í .-' / • — METER OUTPUT ,; / / / : ГГ'7 F,f23 COURSES OF KS450 REACTOR PARAMETERS DUE TO ITS UNCONTROLLED STARTING WITH CIRCULATING SLOftERS SIMULTANEOUSLY DISCONNECTED {NOMINAL POWER LEVEL) £/3i*0.0054 j TC/-2.10'fr" С.Ы -ŠÍO}] 0.25- о L—=— x _.. .. i 1 . TM 0 2 10 to tr : i ТГл1 4 1С p.в - > 0.6 CA-J G (Z) :- to) O.Z \ \ С L Ttá .l(T)['C] TM „••• ичсаытшИг. -iaí-г Lú^A^có '*'<-• L I'VE.. ••-.r'iK •Í^L^C? -->, z: 'O* Л- - 5ГГ) q75 t / / 0,5 20 30 *0 ••.У<*Г)\ K» I '.o ! тл; C.M7 : i Ы(0) CUfíVL ' G( *M 600 | / -r>c7 5fr.:- JÍ» I ir> 1. *[*] 6 .,2* Breakdown cf one <>v twa fciraalaMnsrblower s Assumed с*ш?з# of events due to а br^^kdown of one «м- two circulating blowers is shown on fig. 25* The initia ting even4:- is a failure of a circulating blower (1 - 2) azs-% in the .asч of breakdown if two circulating blowers either a similtane/íis fail are of two circulating blowers (1 - 3) or a failure of self-consumption sectional distri buter HV.1 a breakdown c*f the reeyn-j-hrcniz-ation after swli - siring—.гглг on the stand-by distributer 0'- 3% 0 :^:erning the breakdown of one circulating blov/er the A-1 n ::::;!* *r pcwer station design does not consider the ren ter safety davirws to a>tf the excursion being supposed to be reduce! by the system of control (1-2-4). The ana lyses tak*e ii'.tr» аг-scant bcth the control system failure (". - 2 • 5 - 8)s, if any, and the failure of safety devi ces (1-2-5-9). When breakdown of two circulating blower occurs, the design envisages reactor should be shut-down by safety de vices (1-3-6). Analyzed are also the cases of possible failare cf safety devices (1 - 3 - 7), reduction of the ex- cs.rsirse by the system of control (1 - 3 - 7 - 10), and the control system failure (1-3-7-11)» Figures 26 through 38 present the courses of KS-150 reactcr parameters f*r breakdown of one or two circulating blowers as calculated by the A-1 HAZARD program (KOB). An instantaneous change of the rate of gas flow through the realtor,, proportional to the number of circulating blowers "und^r breakdown" is considered. The analyse is made for uniformly irradiated fsi*X ( ^/b^ * 0;0054; TCg $ - 2,10"*^ ar-.! fcr an optimal adjustment of gas temperature ^csitr.v-, cir.vait according to /45/* A special consideration - 8? - is given to find the BKP optimal parameters (optimal in the first place from the -viewpoint of the efficiency of the reduction of temperature excursions during an acci dent). 6»2olM Breakdown of ou* circulating blower Pig» 26 ehnws the courses of reactor parameters ex cursions due to a breakdown of one of the sir circulating blowers uadar operation» Carves 1 and 2 are the reactor transient characteristics corresponding to a sudden decrea se of the rate of gas flow by some 16,7% (with no influence tf the control system). Carve 3 represents the limitation 88 The effect of power selector speed, on reactor tampe- _"-£t.ivr« *x,r-i?elň.z.a i1*, "as* of & ЪгвакЛ^лп of one circula- -•.л.ъз bl.-w3.'.- (w*.th 3KP dia%-иав:^*-!) is ph-лт on fig 27 *'.- m*.7 'r* з*-?г. th»t- *ua in','.rfiaee of the p^wer- selector sy*',? ^r'oi 1 t-. l.yfo/в. woild net influent- th* magnitu''? •? Я1£.:.1ж<-£ exc^rs^one. The change of power sel*'-.t .»r speer :ле p:-4ít.l.-*ny яз effect if the BKP is connected (fig,P> F.%» 29 sh>i3 th* effect of speed of dying away rf i** • -vr.^it signal frítm BKP (characterized by the adjust- -4iA p«Tfia^t*7' Т«) IY?! the temperature excursions courses ;. ~;>. л.*ве -..f a vrs^bifwn cf one circulating blower,, The J .•-St temperature excursion is not dependent on TJJC Only *;?e m^srr-t idfl of excursions in the further part Of tran- -• -^:.и. zr-• •••*se is infljsno.ed. Ths fffect cf BKP -an the reduction of initiating tern- T~-.i: ^ «•! •'•:.."•=. i -се for temperature control circuit discon r--?tiíí С*, g. the ВИР failure) may be seen on fig. 30. It •'.v"S£'i.3 ^v'• : is that BKP with tender time constant of dying г**? i'T^ л 120 Si) weald provide sufficient time for the *;.*?•&№&si ty shut-down the reactor manually (approximately • ring I ml По.) етеп in the case of failure of the tempera- ••ur9 induced safety system. I.ig, y. presents the courses of reactor parameters in -Le лsř* -i n breakdown of one circulating blower at 66% power la?.' (treskíísLwr. c,t eve of the four circulating blowers un- •,••4? -^ratios),. The change of rate of flow is higher (25%) •\ Asts\.\f +>•*» t^asp^rat ;\r* control circuit has little efficien- ••.?„ -vil vh* «i^'.Uiofl*! íiirouit correcting the power according ;. •... *-iw T'.*^ ?f firw redw9в the maximum temperature excur- з1*щ« relatively m;-':ve than at the nominal power level. Similar courses for breakdown of one circulating Ъ1о- w-ír cr 33% pc'-*- - 89 - oixivers i2>.d=rtr apw.vat.Jon) чг* s-лп on fig, 32- The optimal а-.-'.сай^авп- >f *Ьз temperature c-cntrcl ei-i"-;.r; * valid for th* /onia*? p w«r l^v=»"> i.e not acceptable in this :ae*» and tt-; ••;onrr-:-lli.iig pro^ss, in conformity with /44/, d^s not s*a~ Mil?*.. TV-.8* BKF .т-culd radios the first excursion v*ry t?.i *'',VK ,y4 , V-xt ~.«t^r on conditions suitab?. з for tbi tsiapir-- ra~ JTŠ 'nd'^d action of safety davi.-e woald develop. Tc:-. • "-f.i-i* th* -hang* of SET automatic controller adjustment .?:• th*-. 33% power I«vel is necessary if the reactor is t»? ••^.tin-ie ate operation after the breakdown of the circula ting blcwer. é-.?v2. 8im'"lt»;j^. a breakdown of two circulating blowers Fig-TíB 33 fcírr-;'.-sftw3 8 prasat ths courses of reactor ~?гш-.-:;:•;•-. в ir, t.r.ř; v-as-í c-f fúmultanecis breakdown cf two ^Й,- 33 ••.•'"щрагче the --лигзйв of excursions of reactor pa.7aa^7-f.f s v:t*r *h* action of safety device set into ope rát'.:.)': b-' the sle-r^l. indued bg circulating blowers failu- ;-*- (VÁT--* c?\ further with the influence of the control sy?«t4!E Оглггзя 4 and 3)» and without the influence of the system .;Г •;:atr:>l and safety С curves 1 and 2), If in the •>-4is« :ď ?. breakdown of two circulating blowers on the no- Смфаге! with, the nominal power the steady maximum tem- р*гнт.;с*о cf fa*?l pins on the 33% power level is lower by 26*' С This чгр "tains why from the safety point of view the x:.*il * ::2o>-.rat лге excursion л tTT „o, - 63° С tsee the tab- и« шал .= -а ?,5г- 3£) is equivalent to che excursion of Л t„ „. » --. 37;' 0 f r nomine. level. .-п'.чми. p'.Triiv :.ч•"«-". :;.Ъ« гчл. л ." s*ř-»t.? de^ce *->uld not act -;--,*• л v-;.-"•."" ,*y Steffi w;th MP would r-sdu-** Svibetímt^y.ily th- я--, i•. .жктз. *r. • ,v??í :>г. -if *wape.?fttare of fuel elawmts, se th*.* tcs «-ч.?<ч '#jal.i а-.-it pr^tabiy represent a danger «f damag* .. ; J? д'ft^~ n:r-ab^r >í í'i.~d «lamente. T:-;. *-b^> f "'"Mowing tts irč1::i*sziz-* of the control systttt; '•:? ths ^чф;..'•"- •.::•. • f f~e.\ pir. temperature excursions ie .i:> •.•'i*^.J2:s*'-%-i ''.~his r.n *.-«?!i3 the саве with, as íu-tlea of the <5:\fs"'"у iy" aj ,i!ur\*.*it F.:.e, 3+ sircws rev *;h* p?wer Selector higher speed in- ř'-'„T.w,'K-- *--Ъ.д щоЛта,ш * ejtip^re.t -u^ excursion of fuel element;:'. la-rssi-. ;,f br.4 ?-:.-^c f.r-л ". ť 3*3% s. has no practical sp.::-"-?;:. i* f'/-* thft -*i:i •* * -n. f th-з temperature excursion. m;i- Ъг^.-'.Е.а; f +-vi-i aw.gr-; t ;Д^ -f the temperature feed- н ••i.'-'i: • r. "\з :Г.З*Й f :Ьч !Пт»л:1жии tamperatare sxeursiaas f f'ial :, з demons'":'':ated on f'ig. 35* The favcorabl* influence f tha 3i5;at:.7-i f*»3.lba>-.Jc is practically fully eliminated by :.v.a f\s* ^^ti^r. '-.-f th* natron density automatic control ler \n t:.t *.w-,~ '."%, - ",:»- »„';, ••;-.s.i;t • liae suitable f*>r the aotion of sa- : -**7 ;4T, ..4. v-, ';•> .;.-? >,.y t-arap^rature would dev*Iap by some 35 ss 91 - '-:£ф> 20° О* Ths ^.perfttl.tig pe;'s.jiuaej woald be provided •rJLťh. some 30 8. to shut-down the rsa-vt :r лчпичНу эуеь :* t£,e temperature baaed safety system ^raid fail. Pig* 38 presents the courses of reactor param«*-ftrs l . *^e breakdown of two circulating blowers on th« 66% pr.w- ; «..9 7el (breakdown sf tw.t of foar operating circulating r*.-• ?^rs.'. W.„tb:,it i-h.9 a-'tíi>íi :-f *he safety system initial?-" bj tib blowers failed^ the ^ontrc.l system wi^h ВЫ? w::ald т ; 7.ЛС9 substantially the maximum temperature excursions it 3-;A1 slsa^nts,, b*it despite this the permissible short-tins- .-•- Г4?'Ъ^'-Л1ПЙ af the cans of ar ani tun pins by some 37° С wci* 1 ,.,«. + > Sin'ilt.aneO'IB breakdown cf two circulating blowers -•:«• ts.g 56% p-*wer level compared with the 100% level is there • f^re ms,re expressive,; having as a consequence the danger P ;->3..»ibls damage of a great number of fuel elements. 6.2.3. Evaluation of results The «acrees of reactcr parameters (figures 26 through 33) have been calcalated under the assumption that in tha case of a breakdown of one or two circulating blowers the ršťs лГ gas flow through the reactor decreases instantane- : -злу ял1 stabilizes on a new level corresponding to the пши- b*i' nt ;ÍT^alating blowers undar operation. In fact, this Against the nominal power the steady maximum fuel tem- -•-.ratarf» on the 66% power level is by some 10% lewer, This •tplairis why en the 66% power level the permissible shórt- • "Ime excursion of fuel elements temperature equals 60° C, - 92 - ргогчз-ч ^аз two stages* During of about 1 a. cf the first stage t.ha aaaber of revolutions decreases rapidly, the 3wfcag-esi9ck valve on the "failed" circulating "blower clo~ sua, and the amoist of gas delivered by the other blowers remaining in operation increases. This stage is represen ted on the blower operating characteristics (fig. 39) by the region Ag - Be (for the failure cf one of six cii'sula- ting blowers 'ind^r operation), by the region A, - B^, (for *h9 failure of twa of six operating circulating blowers)t etc. In the second stage the control circuit of the rate cf g=? flow г-чрг-^sents the control elements of the cireu- Iř.tlrsg blnwsrs trader operation and stabilizes the rate of gas flow on a gi.v=m -/аГл*} tti fig. 39 this stage is repre sented by the regi ;n B,- - A,- (for a breakdown of one of SÍT circulating blowers under operation), and by the re gion B^ - A^ (for a breakdown of two of six operating eir- cAlating hi;wars), etc. The second "controlling" stage is much longer than the first one (the change of the rate of gas flow is given by the speed of the servomotor of the flow rate automatic controller which, according to /46/, should equal 90°/112 в.). аг-v? changes of the quantity of gas delivered by the circulating blowers under operation, as well as the change of the rate of gas flow through the reactor corresponding ti the end of the first and the eecond stage of transient process are given in the following table: - 93 - Table 6 State towards the end ' otate towards the e* Humber of cf the first stage i of the second stage circulating blowers The quantity Rate of !The quantity Rate of failed of gas deli gas flow jof gas deli gas flo:: vered by the through vered by the through circulating the reac ,circulating the reac blower tor, blov.er tor, under under operation, operation, G(r)/Q(0) G G %/%v ZAD_ l/ D,ZAD One of six IЛ10 0,9246 1.0 0.8333 Two of six 1*233 0*8223 1.0 0.6667 One of four lft065 0Л990 1.0 0,7500 Two of four 1Л20 0*5600 1*0 0,5000 ! One of two 1,028 0,5140 1*0 0.5000 The table shows that in the case of one or two cir culating blowers breakdown on the nominal power level (breakdown of one or two circulating blowers from six in operation) the sudden ehř-Jige oí" the rate of gas flow through the reactor corresponding to the first stage of the transient process, is approximately half as large com- pare 94 ,. and J3% pcí.-л \-;-.re'.:! The sudden change of gao flow rat- rears approxi;iii»tti..y lo That which corresponds to the num ber of cl^ :ula+ing blowers failed* In comparison with the гез^Лз of -~\~ analysis (figures 26 through 30 and '-.br ;>g; .•* th« a -uTi 1лв of fuel elements maximum ^empíraí^re f :r the case of one or two circulating bio- •vara ^reaidowr. &% nominal power level would he lower and •^he system of control equipped Ъу the BKP wou3,d be able to maintain the fuel temperature below the upper value -rmit-ted,. Cfe the 66% and 33% power levels the actual •^empsrature excursions would Ъе similar to the results >? the analysis (figures 31, 32, and 38), and in the .ase of break.i>*:a of tw:. .ireulating blowers with the safety system inactive the fuel elements would become overheated,. í„2^4« Summary : ' e2,4tl* In case of a breakdown of one circulating blo wer on the 100%s 66%s or 33% power levels (it is the breakdown of one circulating blower of six, four, or two blov/ers under operation), the shut down of the reactor by the safety system induced by the blower failed is not necessary. Reactor control system equipped with, the block to correct the power according to %Це rate of aas flay (BKP) is able to reduce effectively где maximum excursions of fuel elements tempe rature t A-ith no excess of the short-time per missible excursion equalling to + 50 C# The ga3 temperature control circuit alone (without BXP) would not sufficiently reduce the tempe rature excursions even with an increased speed of power selector,, • Э5 .. In the esse of simultaneous breakdown of two os culating blowers the realtor is to be shut-down by the safety system initiated by the failed bio wers« If with two circulating blowers breakdown on the nominal power level the safety system wo^ -1 not act, the system of reactor control eq\iipme±r with BKP would maintain the maximum temperature of the fuel elements below a permissible value. With the 66% power level and the breakdown of two circulating blowers without any action of the safety system the transient maximum excursion of fuel elements temperature of about 97 С would occur owing to which ceitain fuel elements would suffer damage» „ The circuit correcting the power according to the rate of gas flow, containing BKP, must be a necessary component of the reactor control sys tem if the dangerous temperature excursions due to the breakdown of circulating blowers are to be reduced. It is very important that reliable and proper BKP operation be provided for all opera tional power levels. Basic parameters of BKP adjusting (from the viewpoint of minimum transient temperature excur sions as .«ell as from the viewpoint of a safe speed of increasing the reactor power) are as follows: the "magnification coefficient" Kp/Tjj = » 1, and the "dying away" time constant T-JJ^.120 s. * Brief evaluations, of consequences according го fig. 25 [The course 'of events | (fig- 25'• ! The ^emoeiatvr,- •-.; >d effe. :' all p-jv.er levels, he t^ani;iifit -?-4 i Étti.jBiTi-tíi.iýOřůtu: •i fue iau:-e i A t ^ r~° ~ | a) Failure of ЖР results 33% р^'вг level the „max / At,. mar > 200- С;„ The a-tun of the safc-Л. v ^ Ша & v>_-. tem induce.3. Ь^ gas temperature domes late owjii,- which the fuel elements suffer cjamage., b) Failure of ERT (with BK? propjerly operating is act dangerous provided that there will Ъе Г action of the temperature induced safety system or that the realtor is shut-down by the presc; button by 1 minute after the warning signal 1л i •-•eů by gas temperature. With the failure of BKP the safety system ind-ced by gas temperature алее not reáii:-e the maximum ? .п о temperature ex-r-uraiv^ (this results in fuel dam' ge owing to overheating" 2-5-9 The same as in th-= foregoing casr The transient maximum temperature excursion of 3. 6 fuel elements Л tT: пи.., < 50° 0 provided the a-v I dent signal delay £'£ < o/.; a) The temperature excursions on the nominal po wer level effectively reduced, At,, Шах^50° С» .^j b) The temperature ex-arsion on the*66% power le vel, &%Vi ma.v > ЬО- С - a proportion of fue] ele« ments suffer damage due to overheating. ft* Failure of BKP leads to 4tT, > 50° С и* max ' b) In case of ERt failure the BKP reduces the 1:•7-1 T 1 first maximum excursions of temperature as in (3 ^7-10), If the reactor is not shut-down manual1.v within i minute,, overheating of fuel x'esulte. F 25 THE COURSE CF EVENTS AS ASSUMED FOR THE CASE OF „BREAKDOWN"' OF ONE OR TWO CIRCULATING BLOWEFQ !—"" r-ř; i (2)\ 1—* Г. *" ' ' ; ř . -, — . re1j r | ~v :„ • ы -' 7 .''' J i •. ...._ r. i e. (1) ГА. * 4 r i aftř>,f. : v'. • F-JLl~h '; far i^lRIBi'T-k 'i'F doj A \ t . •, ) „. у . REACT* f/; Fr.SYHCHf-ONlZ-WOK 1 jH CAffTr .-•• ".-A' » (n) Fia.26 THE COURSES OF KS -ffO RECTOR FARAMErf*š ?•'.•>' .'•:'' BREAKDOWN OF ONE CIRCL-..ATING BLOWER nt, NOMINAL. POWER LEVEL Ю го to 50 V> ю 6c«9 0(0) 09 \ qe f ;[CJ itCj 'f-j -&j ;fii :t*j • ft; i: О •>-€ POWER AND rQiPER/ . _ CO t ос i ^ CWTRCL CIRCUIT aSCCHt^Tfn — ч ж : Ьоо54_|-гло ''-г<с>*^^; ?s i£E: rr-5j TOR /tti. CVt5£S átTC€)['C] 40 3 1 *~-..tr 2 20 \ líZ Ю F,?2? THEEF-ECT OF THE POWER AUJUSTH& DEVICE SPEED ON Г'' COURSES Oh KS-tSO REACTOR PARAMETERS ON THE A/aW;,v4. POWER LEVEL (Win; BKP DISCONNECTED) 20 30 *0 SO Г.О" ГС л] G/Ti G(0l o? o a ТСл] NlD Wů~) EftT BKP CURVE: tOi 1-е-} M (T)C'cJ •1*2- lil ill? I 0005.4 •2KJ* ~2 Iff* 5 22,12 100 55 2 0,005* •2.Í0* -2 Ю+ 5 79,00 К 30 — - 52 ICA] TCA] -20- АО г ; 2 At fr\!T] ~ i : 20 . •20 ', Fiq-28. THE EFFECT OF THE POWER ADJUSTING DEVICE SPEED ON THE COURSES OF KS-150 REACTOR PARAMETERS ON THE NOMINAL POWER LEVEL ( WITH ЭКР) Girl 0 Ю 30 40 ТСл! 0,9 0,8 10 20 30 40 NJTJ_ 1.0 - Ť[*1 N10) Л 0,9 Wy^^H\^^^ 0,9: ERT ВИР At,,V,ma, CURVE £*i л(, (rli'Cl ['Г'] [-1 [л J T'cJ Unnr .из 4 1*1 J 0.0054 0 -2.Ю"- 5 22,12 16 100 120 120 33 2 0,0054 0 -2. JO'" 5 77,00 16 30 120 120 33 TM Tt*1 Fi [ul re r ERT BKP CURVE SA к, J' W [-2 1*1 1 30 30 33 0,005 i -2.,6S •2 70* 5 77 0 16 30 2 120 120 33 t(*) tte) ui mccl Fif 30: THE EFFECT OF 9KP ON Т.чЕ COURSES OF KS-fSQ REACTOR PARAMETERS, BREAKDOWN OF ONE CÍRCJLA7ING BLOWER ON NOMINAL POWER LE''EL, TEMPERATURE CONTROL CIRCUIT DISCONNECTED о го 30 40 sc */*j ю -1- act) его; Ofi as t 0 x „J3O 40 50 ip о >° *. - «. - л л *[»} Fl too TW G(T) G(0) NlT) itjT)['t] 7 CH I ~r ERT BKP CURVE Тжг Tyy M j П C-l Сш] I £*] ! 1 I THE POWER AND TEMPERATURE м I I CONTROL DISCONNECTED ! г 0fiO54 \-2.10'\-2.fO"\ 5 22,12 « 100 ~ - 69 3 I i 22,12 16 400 120 120 41 ГСл) it <т) се) ТГл) utjnM л т во jí-*-*.-. О •- Т[л) --ř" 90 100 -го . r.f32: /='.9-/50 h'F.ACmr PARAMETERS CCLRSCS DUE TO BREAKDOWN Of .9.\c СЯС c./.'"VC BLOWER ON ' • •£ 33 '/ ?0W£R LEVEL 00 Г[А] W Т[л] M (1}ГХ1 iSo7^ •so- Г""1 T :>?r 1 £KP J, Л л.. -.... -„...... ' .-U4^J--.ULJJŽL...Í...^ .J,& r- r:tL' №b~Q 4-VÍ miPEZfi.TjRt: roo f- -j . .; • i f^o ;_ o 246 i -I • • ; - • - j - .JL. . ..L !.. j_f^? I № I :<^ j_ ^ 120 S3 ~.'R'i1 vC^y-if/i.'-' O" Г'й fí/f,'. £4/»' .'" rPřt/iTuR* Fíf33. KS-150 REACTOR PARAMETERS COURSES DUE TO BREAKDOWN OF м . CIRCULATING BLOWERS ON THE NOMINAL POWER LEVEL 6(4) 0(0) NCC) <*» N(O) rcu rcm h ERT BKP CURVE Tn T„ TB *%J*)fc] [V] [s] [*1 1 0 0 5 THE TEMPERATURE AW POWER1 m 2 CONTROL UNCONNECTED 66 3 -2.;ó' -г. A* 5 22,12 16 Ю0 — - 139 i 22,12 № КЮ 120 120 7* 5 REACTOR SHUT-DOWN BY THE SAFETY SYSrEh 43 i .... TW —tro *M =*»*/•: F^ji ГНЕ EFFECT OF THE POWER ADJUSTING DEVICE Z^EED 0Г. ^E COURSES OF KS-1SC REACTOR ťARAKEKřb' DUE TJ BREAKDOWN OF TV/0 CIRCULATING BLOWERS ON NGM'NAL POWER LEVEL вер 0(0) N(O) . ERT ВКР rr _, ... ri«V£ I*, к. N'] L'c'J 1*1 1*1 [Ч 22,12 Iff too П9 GOOW -1.Ю \-2.llf 5 2 77,00 « 30 - I2é *ы F 35 THE EFFECT OF THE TEMPERATURE FEEDBACK ON KS~lSO REACTOR * 'PARAMETERS COURSES DUE TO BREAKDOWN OF TWO CIRCULATING BLOWERS ON NOMINAL POWER LEVEL 0(4) 0(0) NPt) N(O) ERT BKP CURVC m M I'] Zf fo fc [*] Í--1[X] ) иг " *J*>N 0 0 U 2 BO ao a 5 12, a «- Ю0 3 - - us к ri' -2Ю' i ao eo в» «Л/ Г.аЗб THE EFFECT OF BKP AMPLIFICATION 17?}ON K5 - fSO *E*CJOR PARAMETERS COURSES DUE TO BREAKDOWN OF TWO CIRCULATING BLOWERS ON NOMINAL POWER LEVEL iW em ШР OJRVC If» [V] [V] l-i l"i h i 0 во иг 2 JO во US 3 to 0 •;«• 5 21,a « ко во н * no во n 5 ISO во n В 200 во n *w • во t -«o t Fief 3? THE EFFECT OF BKP ON KS-150 REACTOR PARAMETERS COURSES BREAKDOWN OF TWO CIRCULATING BLOWERS ON NOMINAL POWER LEVEL, TEMPERATURE CONTROL CIRCUIT DISCONNECTED , rca I кт T, сивиг Z ERT "' \r 4(s] G(0) Atu^vM \ | гс„ тс T, CRT ВНР л!., :uRvc;I/3, i r. .,j r, .,j c f T„ к. X fsl Is] POWER AND TEMPERATURE 110 i i CONTROL DISCONNECTED '2.1Ó1" 5 22,12 16 100 - - 263 "Л j _. 22,12 16 Ю0 120 120 97 < \REACOR SHUT-DOWN By THE SAFETY STSTEll 43 TW 200 &tm wpc] *>TW 200 ?w 1! 1/5ii »- r ! 02 I к К 1 С ' o' ' 1^ * Li- , * t EV ! tí ' 5> \%* "- 12 - •^ felv 6„3« b-ss of e*lf~^onauttption supply of tbe power station AscTvuaeá e<>«rs~ of this н> -i-tent event is shown on fig- 40. la this case the initiating svent is the 220 kV network disintegration followed by the loss of voltage in the 110 kT network and by the runaway of two of the three tur bines (c-r an emergency switch-off of the switch on the 220 kV A-l power station - Křižovaný line, or an emergency switch-off of the three press switches of the 220 kV gene- $at®-r-transformer blocks, followed by the runaway of two turbines as well)* If the 220 kV superior system failure is not accompanied by the disintegration of the 110 kV network, the self-consumption supply is automatically switched-cver on the transformers T 18 and T 19 (1 - 3)« The loss of the self-consumption supply of the nuc lear power station results in that all the circulating blowers become inoperable (1 - 2 - 4)« In this situation the reactor is to be shut-down by the system of safety (1-2-4-5) and transferred automatically into the regime of emergency cooling, similarly as in a number of other accident situations initiated by the "internal" fai lures in the power scation* The case of safety system fai lure (1- 2 - 4~ 6) is analyzed, as well as the effect of the control system (1-2-4-6-7) and its failure (1-2-4-6-8), The analysis of accident situations taking into con sideration the action of the reactor safety system has been made by the use of the A-l SHOCK program (Supplement 4)» while the analysis of accident situations with safety system failure by the A-l HAZARD program (K0B)# £..3.!..- Loss of eelf-consumpt' on supply followed by reac t: After the action af reactor ~afety system its therm?.* power decreases quickly below 10% of its original vsCue., With the KS-150 reactor, typical f sr а чету small thermal inertia of its core, the emergency shut»down is related with a sudden decrease of the gas outlet temperature, thus causing a very dangerous thermal sho:-k on the reactor ont-- 1st piping, the main gas piping, and the steam generator iaiet parts. According to the realtor design the thermal shook is to be softened considerably by means of timely reduction of the rate of gas flow through the reactor by partial closing of the regulating vanes and simultaneous opening of fast working bypass valves of the circulating blowers* Pig*. 41 shows how the reduction elements reduce the rate of gas flow through the reactor (the action of the re duction elements is considered to be simultaneous with the action of the reactor safety system)• With no action of the reduction elements (curve 1) the reactor gas flow rate rises moderately at the beginning owing to a decrease of its temperature, and then slowly falls due to the fact that the circulating blowers run out and their revolutions de crease.. With partial closing of the circulating blowers regulating vanes the rate of go a flow may be reduced du ring 5 e. to some 30 - 40% of its nominal level (curve 2)* The opening of the fast working bypass valves alone will reduce the rate of gas flow through the reactor during 2 s, to 25% (curve 3)* By simultaneous partial closing of regulating vanes and opening the bypass valves the rate of gas flow may be reduced during two s« below 20% of its nominal level. 114 •- Pig* 42 presents the courses of the averaged tempera 4 ture of the cooling gas ( -mean) on the outlet from the reac tor core after reactor emergency shut-down. The effect oi & sudden reduction of reactor gas flow rate is demonstra ted in table 8 by the use of maximum value of the transit " thermal stress caused by the emergency shut-down of the reaet>rfc This stress develops* in the "hot" knee of the 520 (dia) • 20 primary piping /47/• Table 8 The The method of reactor Maximum thermal Time course gas flow rate reduc stress on ttye needed of tem tion knee wall ijitoer to reach peratu surface the ma re ximum accor stress ding to ф ab— fig. 42 A To so^Ste r value о relative max (s) (s) [grad) ($rad) (kp.cm ) т4*ие (s) 1 45 45 1116 t.000 18 2 5.0 45 79 418 0.375 34 3 0„6 45 45 230 0.206 35 4 4.0 0.6 45 65 135 0.121 39 The table demonstrates that a sufficient reduction of the thermal shock (to some 20% of its fullj value) may be realized by means of a reduction of the ra!te of gas flow through the reaeter using the opening of the fast working bypass valves on the circulating blowers (two val ves 250 mm in dia». on every circulating blower) with no additional partial «losing of the regulating vanes. ''"':.# •:••.?- сJ -ore^aar^ ŘK.J.?S.IJ:Í in the reactor ;г^э- . .-ляГ'-.-г- af.'-x tří ^'lergrřncy ^hut-do'.vn is demonstrated on :g, .a."-, '/"'VÍ ч ..li i-'.-. ." а-.-Каты з:-:.в reducing the rate of • •"--::-£ oi:; '^" •*" •-'Л л^л +-he re" J- r --.auses a relatively s'aiti. Я- ••----..- f r-eň -^ i:/.-v gňs pressure by some .•..j.'t.t"",, »'.:. ^ лчу if a.'. + •.. ч'-i ex essi'/e r-"e.*ise of car- - •>'. 5.; «.; ie •""?'.•!. •'••.•.-? ы.аег*1 г or t^ a redistribution ox •ci-'ss-.'e .5..! "ferer.-.-es iu th- reactor, with whi^-h ar. inc- "ti>r,íj--. sť.es^-. ::§ -i'-i w• "че •-••r-'lirg of certain realtor com- ?^„ Д4 р:"ез^п1з the : trse ;f gas quantity delivered ">y the ?; ir-'Ula^ing blowers агЛ fig. 45 shows the courses ;.f gas pre ь.rare in i he sucticn and the outlet end of eir- -.-•;1а*;.^.ё lowers after realtor emergency shut-down. The rjicvrJiient :.-f tr-.s w;-rking poir.it of the circulating blower niter the гея-: г emergency shut-down for various methods f gas ii •>. Si red.i tion is presented on fig. 47# Pressu re bi;'i.g=3 •„•! p?ri. iicaL. nature occuring in the course of '}>•, геле :^lac+..,u through simultaneous opening of : ypas.- -'•sl.es and р^гча.. closing of regulating vanes du ring ".-'ÍVÍ fj'rst ? s„ of +he transient process (fig. 45) aay 'e dagger:^s in i hat the oil may penetrate into the prima- Г..З.-2,. L' as of self-consumption supply with subsequent fnilure of rea;-tor safety system (fig. 47» 48) let из ana^y?,* th- ;HOSÍ *mprr,>able case of the KS-150 /aaotor -л-л :ii'-?l..': safety oyst^ra Г?.Llure after the loss of power statiy/i self-consumption supply. Th- it?.-••--4..:- of g;,£, pressure in the reactor ia зир- p:aed t. ... ^tiii';.ia '-p '••:< ,0% .ď its initial level (appro- •/.шеЛ^лт* ••••" ,-.',j.„- • ii-,i4.ia^ i v-i /v/c o* ti?я circulating ЪIо- ..'-:?«'v. И >• •чгзеулэп": connect .:.-•_• л •, ř a pre-eelected circus-- r.g Ь.л'л'е ' од the Madunice's utar-Л-Ъу souroe)u Fig» i-7 эЬ<-мв h;>w the rontrol system (equipped with л1*"?'"' jr.ďý.-jenr-es *he course cf realtor power and the ma- ••Г.ЗКЕС f ^mp rat гге .-? fuel -j laments during the uniform '•across, i +h- r:--:* of ga3 fl-v. fr_->m 100 to L0%a It may • - 8ian thit »<:.\'a i, spee3 cf gas flow rate decrease of .;„?%•''.., Corresponding +o a maximum speed of decrease of ':>.? fi;...v.ti-.y ;f gas delivered during the run^out of the . iiv,...-,+ir.g il^wer-i;, the turbogenerators inertia being ta- ien J...-. а с'...i i the fuel temperature excursion during the first '00s.. -i.r-.+ ?rval is below 50 C, Of course, it 1з . 9o-.esзагу ':.-. taks into consideration that the operation of the ccntr- . system on the 10% power level cannot be stable, -3.v j.ii:: гЬе rea.rt- -r must be 3hut~down manually (it may be added that the operating personnel hao some 2 minutes avai lable } ;. The :ntr~:. a;,-stem without БКР is not capable to re duce suf f .1, lent Ly the initial fuel temperature excursions* Fig* '3 shows the effect of the control system wit hout БЕР on the course of fuel element maximum temperatu re fcr an .instantaneous decrease of gas flow rate (cur- res I and i). and for a decrease of 4<.5%/s (curves 2 and l7 >)9 HB iiej аз for a decrease of 1,8%/s (curves 3 and 6). lue :,J the "w?ll stabilizing circuit of neutron density 4'itr.:i. (v.hL-h responds very quickly to the reactivity •-hai-..^ft; ; .;э*т-а ly The hoating of uranium) and owing to -..ae ei~ •'Vv'",y-iv4i.'-..s of gas temperature control, the excur- •:_-.-.»n ••- i\--,?l +ft.:;rperature is too high and almost indepen- 'L'/Л .'.te.:;! -;f decrease of gas flow rate (curves 4 t-U'iW-.i.. •'-.ч- л •'.i-'-.tiVL't с-; •• ;ií *,-">ii ..•H-jaei vy 1?-í \Obd Ci ^v- -,>•;;. г*H.'.'..• tit-..Г- ..••vi-.^.ij.tion зирр1у is til? 2-;.-\ о Почи ,ae^ :ui.-' nee-í-з raat the KS-150 roa-ř^r I-- ah J- l .та vy -he ляг.г-у system^ ulid reqxlr^- -:'.|.,:-л х::з rest he&\ h.j ratably renrved., 7v"rto«- ^iei>:°i -,p гэ-i •t-'^ ri,-.\; ,-a+i • safety avitem f>-.ij\. + v-?v • :•: .^-./* w^vli It о-Лп :-antially softened 1л :"je a-.rr-.iáfcttt . rui^^rii t.y a control system e^t тре^ *5.'-h. "l*'" • V,, a •••ř.t;ř 'at reactcr wou* ' h^ve ťi he sli„: t-ď:iT-ii "i,7 ;jO;r.í'í two ffiirr.'.tis,. LT-':50 r-?a'-t-~r emerg'.ucy slnjt.-ďjí.oi саи^еь ••:, •thermal sl'o-'k on +:is --atlet parts of the reactor. 4 h; pri:xa"y ť'í,ir%s «.id Lhe inlet components of ihe tí-л-ый gin-rat era. The thermal shock may he s-if'fi •• ler-tly л i derate! Vy opening the fast wor king ':ypas£; va'vcs yi^oirt. simultaneous partial '..IcGraife 0'" th-i •,J. 1 jul-.tlag blowers regulating Vo» ,1.3 3, The frti-t reducti.cn c? the gas flow zio.te through. •he readier inaae with "he aim of softening the thftrm'; :^j-v-.ic -111 oa;^e a redistribution of pre.ia.iVi •7lfff.-ře".-ei3 -: a +ho reactor., аз well ^ peri :';.и:^я.'.. ргеззиге c-httr-ges iii the auction ык1 4e';i.¥ftry oidt :f the blowers, Tt is therefore -я,г K'.rabl .v t... investigate -these phenomena from the vievp••Ant of /•••ollrig anci 3+re:;Ging of the reac- '. "• ivv* '• ''lul p--írtar "iflnetica of oarhon dioxi-5* nrlfc-;-';: v;.', of th^ t;::>ae';^t:>-, penetrating *f Мчч* :" J,Y: ,:<< the v-'.,.'•-/•: , '^4' th'; primary .^Г^аЧ., •: л 6,»3.3*4» Brief evaluation of consequences according to fig, 40s The coarse of •vents (fig. 40) Consequences Owing to reacor shut-down by the safety system, a thermal shock ori ginates. This shock may be substan tially softened if the bypass valves 1-2 of the circulating blowers are opene in time (simultaneous partial closing of regulating vanes is not necessary) Sadden redaction of the gas flow rate is associated with rather high local changes of gas pressure in the prima ry circuit. The control system equipped with BKP is capable to maintain in the course of two minutes the maximum temperatu re of fuel elements below the permis 1-2-4 sible limiting level. Provided the reactor is not shut-down manually du ring these two minutes, progressive overheating of the fuel, resulting in its melting, would occur. Damage of fuel elements in a large 8 extent due to overheating above the melting point of both uranium and cans. - 119 - iO: THE ASSUMED COURSE OF EVENTS TN THE CASE OF LOSS OF POWR STATION SELF - CONSUMPTION SUPPLY IACCIDENT SIG- ] 1 iNALLING CWING SHUTDOWN OF [CONNECTING OF TO THf-FAILURES THE REACTOR BY(5) [PRESELECTED —* BLOWER TO 720 \OF PCWER STA~ THE SYSTEM OF iIIQN_pEyiCES__} 1 TRANSFORMER • SAFETY I ( LDSSCFXCLTAGE BREAKDOWN OF IN HOW NETWORKS?) ALL CiRCULAVNG (4) RUNAWAY OF **• CONTROL TWO OR THREE BLOWERS THE 220 fcV TURBINES 1—•"' SYSTEM (7) SUPERIOR (1) SYSTEM OF OPERATES SYSTEM SAFETY (6) DISINTEGRATION CONNECTING OF SELFCONSLMPTia FAILS TO T&AND T19 (3) CONTROL. TRANSFORMERS SYSTEM 18) FAILS F&.41; ТЦЕ RATE OF GAS FLOW IN THE REACTOR CORE AFTER THF EMERGENCY SHUT-DOWN OF THE REACTOR ÍC&Ěšbi *Ё^^^ЁеШ^&^^^вШ^^мМау^^Ш Fir 42- THE COURSE OF GAS TEMPERATURE ON THE REACTOR CORE OUTLET AFTER THE EMERGENCY SHUT-DOWN OF THE REACTOR 500 г CURVC AXefsJ **,['] 4>[ar*i) 4*[^*J Atm (%)fX] t - - «5 45 i Sfi — 45 79 3 — 0,9 4S 45 400 * *fl Ofi 45 «5 X» 300 Fie- 43: THE COURSE OF GAS PRESSURE EXCURSION IN THE INLET CHAMBER AFTER THE EMERGENCY SHUT-DOWN OF THE REACTOR &*>?(*£' /*p-c"í j -2.5 Г -V> 75 -ÍO 4/V *.S 15 X ?i FbjM- THE COURSE OF THE QUANTITY OF GAS DELIVERED BY THE CIRCULATING BLOWERS AFTER THE EMERGENCY SHUT-DOWN OF THE REACTOR CURVE **B(*J лт.г»; Vfr"*7 fbn^l t — — 45 45 2 *> — 45 79 3 — 0,9 45 45 4 *fi 0,9 45 «5 O.fX) 0,(0) *F*J '•if *C THE COURSE OF GAS PRESSURE IN THE SUCTION AND DELIVERY Sue OF THE BLCWERS AFTER THE EMERGENCY SHUT-COWN OF THE REACTOR , . у .... CURVE: A'Za[sJ it,/s7 i W'fgrad] • ^[grad] t — | — \ AS \ AS г 5,0 1 ~ 1 *S ! 73 ± 3 4 A,Q j 0,6 \ 45 ' 65 } i 3 NOTE p, (0) - 55,955 kp/cm' p„(0) "i3,Aii Hp/cnf 0.9 9.^ó: MOVEMENT GF >HE CIRCULATING BLOWER OPERATING POINT APrER THE SHUT DOWN OF THF REACTOR BY THE SYSTEM OF SAFETY — - j CURV!: **,,t*] Ufei'sJ | ^[gt-rtj i r*is*íij Í T" " i — ' — ! 45 45 Ž 5,0 ! — i 4.4 • 79 .? __!__.. 4 4,0 Fift?* THE COURSE OF REACTOR PARAMETERS ibRi.NG THE CONSTANT SPEED-DECREASE OF THE GAS FLOW RATE FROM 100 TO ЮХ 9(0) °* я[*] N(0) 0$ rcu ERT BKP WRVE том [X'J [X'J [*] I" ffj r0 Off fa h 1 0,045 -2 10* -2.16* 5 22,12 IS 100 BO 120 г 0,0» 0* 02 О 20 i :-iti4.\:J'" hc: "4/ ^ÍT : -LZW -'z~- FPQÍS to:':'. : "•'.'•CťVvJ:" "'^ 9е «tň/ S*° TJV TW wr 0,UW> Tft/ so n too ., 4 r~ ? • 4 5 6 s(V- SfOl / 5 ft) A • 0,2J. • С '> о *•+* R'Aptore of one of t*.« twelvn c"Ay.-. piling, or иЫ- :. pvw*»r station přiměly circuit on the nominal ?.с#ы level The : >arse Л events assumed for this acci.ivat 3itu.-,. • v;.r; is shows on fig, 49* The initiating event is я г ptu- rs suddenly originated in one cf the twelve main cot 1st .--.?• :i:ulet pipings of the primary circuit* As the cooling gas .escapes through the ri.pť.ire the gas ш*л.^ pressure in the primary circuit decreases, and the gas temperature on the ••"•=?a':-tr*r outl*- increase?.. By a redistribution of gas pres- в;.гз in the primary circuit the operation cf circulating *lowers плу r«.«a.i?i uninfluenced (I - 2 - 5), or failure of ;Г!УД1*Л1ТЛ£ >•:,-"wer1 in the loop with the piping ruptured ma* *.ле.го? (1 - 2 - 4), If the gae pressure in the suction •••••£ th*> sir-julaJ.i.iig "blowers falls below a given value or if tb.fi gas tenperature on the realtor outlet increases over a given val-ze, the reactor shoali be shut-down by its safety devices (1 - 2 - 3 - 5) or (I - 2 - 4 - 7) and transferred int'.- the regime of emergency cooling. In case of safety de vices fail, then .independently of reasons of the failure temperatures ezrarsions may Ъв reduced by the system of cont rol (1 - 2 - T"- 6 - 10 or .1 -2-4-8-12), or the ex treme case cf failure of both the safety device and the cont rol system (1 - 2 - 3 - 6 - 11 or 1 «- 2 - 4 - 8 - 13) юау <>,;;cur* In the framework cf this report the consequences of primary circuit .rupture have been analyzed with the use f the A-l APPORT program (Supplement 5>* 1'his analysis has oe*a mad* from the following points of VI?)WÍ - so«3d of the gas flow through the reactor., - +;::•?оз.'.^.c'% 'lynamieal f \ v-ces З',:;*:ац on •r':,y- Lan-u? parts of '- i-nfT.u^rj^a *f thtí fchroi + ling elements additionally Ъил!- is"7c tne inlet or outlet nozzles of the reaotcr to re~ ďa-ле the conse... nances of the accident under considera tion. The effect of the system cf control on the reda<-tic.t- of temperature excursions criminated due to tho primary cjrcr^t piping rapture has сзеп evaluated using the A-l HAUAfiD program (КОН). Scow important aspoits of the accident under investi gation are nrt analyzed. These are, for instance; the effect c* the pressure wave in the surroundings of the pipitkg rir.pt я"эс1?. emergency cooling of the shut-down reactor in case cf a low coolant pressure, penetrating of the air, water or oil into the primary circuit, etc. 6*4.1. Bupture of one of the twelve inlet gas pipings Let дв represent $пз magnitude of the rupture sud» denly originated in the piping by a parameter ^ defined as the ratio of the rupture to piping cross - section area (for total piping ruptare and the eecape of gas from both the ends opened, /r/ ~ 2). Pig. 50 presents the course of gas flow rate through the reactor after the cold (inlet) piping rupture in de pendence on the extent of the rupture. Curves 1-4 repre sent the ease without fch* action of reactor safety device, in which case th« circulating blowers remain in operation with nominal number of revolutions. Curves 1+ - 4+ are va lid for the case with safety device action when the circu lating blo7*ers are in operation with falling revolutions, the regulating vanes beir^g partially closed and the fast wortť.rg bypas» ; alvss opemed (the action of the safety de- v:4a :,.& hex* аавмч-а t-\ t.afce pla«e simultaneously with the - 130 - origination cf rapture ia tive p:-p'ng). The first situation is characteristic for a sudden ia.':refcse if gas flow rate through the rsa-rtcr (foT/tv •= '? during some 3 3. by 40%). Further gas flow rate decrease á.v.9 to the primary circuit discharging has approximately constant speed (foryv- 2 some l.lj^/s.). The excursions cf temperature caused by the sudden decrease of gas flow rate for.--t: é. i could be reduced effectively by the sys tem of control Cf or ,.**• s 1 the reap>tive sudden change of th» rate of gas flow is some l?%t at which the control system equipped with the ВЕР keeps the maximum excursion of fuel elements temperature below 50° С - see also fig. 26). If the dxmansions of the rupture are larger Cfor AvXL), the fuel elements would become overheated before the reactor would be shutdown by the safety device induced either by gas temperature or pressure. Failure of the sys tem bt control before the action af the safety device would result in fael elements overheating provided,» >0.5. If in the second case /и,? 0.6, the inversion of gas flow in the reactor occurs due to the fact that the circu lating blowers deliver a smaller quantity of gas. Рог а gas flow rate in the reactor core stagnating (for/w of about 0.6) the overheating of fuel would result due to the rest heat of the shut-down reactor. The inversion of gas flow in the core of the reactor would be extremely dangerous: a redistribution of pressure differences in the reactor would occur (the reactor core would be acted upon by a p "reversed" overpressure of about 1 kp/сш - see the curve 5 on fig. 54), the reactor inlet parts Ciacluding the flan ged ioiut of the pressure vessel) would be effected on by the hot gas, etc. A real course of consequences of a rupture suddenly originated in the reactor inlet piping would be obviously - 131 - the combination of oofch the cases тчпъ.-и^еа above. Just after the origination of the rupture fcha oTansient pro ces? would í-11 "••?.• t.he first situation, until the action •>f th=i s.sxet/ á*vi.' 43 indaced by $л& temperature or pres- а:.re takers pln^-э» After the reactor safety device action the crriat.'ty of Va*- cooling gas delivered by the circula ting bicw^rrs wouii b* reduced, ani the transient process would g) on in 1-xne with the second case. The analysis shows «hat in the oase of a larger rupture origination the phQosTpr.y of с-.oí-ling of tha shut-d-wn reactor, con sidered at present, is net acceptable and should be mo dified in such a manner -hat after the reactor is shut- -down tlie eircul-i*.:iAg blowers ought to continue their ope ration with nominal number of revolutions and without any reduction of the quantity of gaE delivered. The сопз q'lanoes of a rupture in the inlet piping could be reduced in principle by the use of Venturi noz zles built into the inlet nozzles of the reactor pressu re vessel as a substitution of the present orifice gauges (said orifice; ide a pressure difference of 0.32 kp/ 2 /car" necaEs; ^v r>--' the cooling of the pressure vessel, absorb:»asr rr.ds and the internal thermal and biological shields of the K3«150 reactcr). The course of the gas flow rate thr* ;^. th* reactor in dependence on the size of the piping гиртдге with the use of Venturi nozzles adjusted f-'T 8 cwui-tanfc pressure loss of about 0.32 kp/cm accor ding; to the Supplement 5» is presented OIU fig. 51» On the iigure it may b« sean the marked reduction of the gae flow rate decrease fc-r^. ,'- 0»3f but in the second phase of the transient process (after the action of the system of safe ty), the nozzles in their present design would not prevent in =.,].;. probability' the gas flow in tha reaotcr core from stagnate-n (:u: 'c.s 3+, 4 ). T.• avoid stagnation, the noz r zles w". лЬ". IA..7Z ---i "t* di íuftnsí onnd for lower critical rate « \*Л - of flow which, on the ; tv^r hand, would, be associated wi- a higher permanent i >ss i' " assure. 6.4i2-« Rupture of one rf the twelve cufl.et piping» fig, 52 sh^ws the dependence of the rate of gas flow through the realtor after a rapture of thd hot (outlet) pi ping on the size vf the rap tare» C-irves 1-5 represent th situation without the action of safety 3evi?8, while cal ves 1* - 5* are valid for a case with reactor safety devi ce aetion. In the beginning of the former case marked periodical excursions of the gas flow rate through the reactor are ob vious (for /и, .-•• 2 th* excursion during 1 s. is some 17%) which disappear in th* course af some 5 s., and the rata of gas flow decreases wlfch an almost constant speed (for .к = - 2 some 1.7%/s.). It has been proved that in the steam ge nerator in the ruptured branch of the primary circuit the inversion of gas flow f*r /к > 0.5 would occur, and the cor responding circulating blower would fail. The temperature excursions caused by the constant decrease of gas flow rate could be eliminated by the control system equipped with the BSP until either the action of the safety device induced by the gas pressure or manual shut-down of the reactor (see fig. 47). In the latter case a sudden decrease of the rate of gas flow through the reactor occurs owing to the partial closing of the regulating vanes and the opening of the by pass valves on the blowers. Again, a real course c-t events of an ouO.et piping rup ture would be a combination of both the cases ;>aid above. Just after the initiation of the rupture the rate of gas - 135 - flow through the reactor would temporarily ris« by as mach as 17% and the pressure difference on the reactor core by as much аз 37% (see curve 1 on fig. 54). For hig her size of the rupture С /řv> 0.5) a fail of one circu lating blower would occur, together with the inversion of gas flow in one steam generator. After the action of the safety device the decrease of gas flow Tate through the reactor to soma 25 - 40% of its nominal value would occuri followed by further uniform decrease in conformity with primary circuit discharge. Similarly, as with a rup ture of the inlet piping also here the present philosophy of reactor cooling after its shut-down by one circulating blower is not satisfactory for it does not secure a re liable removal of the r<*st heat with low coolant pressure. $he rupt;.""e oonsse/^n^es in the outlet piping could toe reduced by Vent art tabes b»?,ilt additionally into the pressure vessel outlet nozzles and dimensioned for a per- manent pressure loss of 1 Jrp/cm (see the Supplement 5). Prom fig. 53 it besomas obvious that the Venturi tubes built into the nozzles would eliminate the effect of lar ge ruptures, transferring it into a case of rupture with /H. = 0*5 (see also the curve 3 on fig. 54 and 55). 6.4.3. Summary 6.4.3.1. The origin of a large rupture in the primary cir cuit piping results in an extremely serious acci dent state. In case of a large rupture (/"*•*> 0,6) in the inlet piping, stagnation or inversion of the cooling gas flow in the reactor core would oc cur, which would lead to the overheating of fuel elements. The inverse direction of pressure diffe rences in the reactor cooliflg circuits could result in an excessive stressing of certain important parts -• 134 - d:ie to thermal stresses -r mechanical forces. Rup ture «if the outlet piping would he followed by a temporary increase of the pressure difference as high as by 37$ ir* th.e react or core, and, if д- .> 0»5> the inversion of cooling gas flew in the ste?;; generator in the raptured branch would occur (cor responding ^irculatiag blower would fail), 6*4e3»2« The control system equipped v»; rh the BKP is capab le to reduce effectively the excursions of tempe rature for & time interval of about 1 min. (with the inlet piping rapture for /*' ~- lj in the case of the ;*.tlet piping rupture up to {*• = 2). 6.,'+#3«3* Тпэ effects of large ruptures in the piping may be red'vriA-l o.y s-ii table throttling elements (Ventu- ri Л">??1ее or VeE*i*pi tubes) built into the nozzles of the reactor preasare vessel. On the other hand, their buildiag-in would increase the primary cir cuit pressure l*ss by some 10%, which would have a permanently adverse effect on the power station ope rational economy» 6*4»3*4., The present philosophy of the shut-down reactor emergency cooling by one circulating blower is not satisfactory for an accident state caused by a pi~ ping rupture, for it does not secure sufficient re moval of the rest heat in the case of low coolant pressure. • 135 - 6»4-„3w5» Brief aval.',.";. -^ í" the -;.о:^ • - .-.%*.-'^s ,.;/ everr •• ao. i-o Tiling t-> fj.g< it 9 J Table 10 The coarse of events (fig, *94 : Солаифитсвв a.,.? Ре-г ,ч- •• 0.,S in the inlet piping з ;v vзг3.1 on c/f gas flow in the reactor •••. ;:e wit a s-ibs'- r^at -ixtanbiva over- h-^ri^g of tha-iutšl» "с; F~r /-.. > 0,5 an the outlet piping 1 - 2 • - 3 • - 5 лrjversion -f gas flow in one steam generator and failure of one circu lating c..:'W-?r» '} Slat-down reactor emergency cooling by one circulating blower is not suf ficient for low coolant pressure. Substantial redaction of temperature I j 2 - 3 - 6 - excursions for approximately 1 min. - .10 (.then the reactor must be shut-down 1-2-4-8-12 manually) Sxtcuaive damage of fuel elements due 1-2•3^6=4,1 t: their overheating over the tempera- 3-2-4-8-13 tore of melting. - 236 .- r tq- **3- \ř£ CJURSE Cr EVENTS AS ASSUMED 0R THE .Abi OF THE MAIN GAS PIPING RUPTURE E; REAC T':R /-, \EMERGENCY SHUTDOWN BY __L ^i COOLING OF l/gj SYSTEM OF SHUTDOWN SAFETY \*EAC]pR_ CIRCULATING P) BLOWERS SYSTEM OF OPERATE CONTROL (Ю) OPERATES SYSTEM OF <6) SAFETY FAILS SYSTEM OF CONTROL m FAILS RUPTURE ESCAPE OF OF ih'LET OR (1) COOLANT FROM (2) OUTLET [THE PRIMARY PiPÍNC CIRCUIT REAC TOR [EMERGE r.cY SHI J T-'MK' к-- (?) j CC-Ci.'liG CF SYSTEM с; ISHUTDGVVK SAFETY BREAKDOWN OF ONE (4) CIRCULATING SYSTEM CF BLOWER CONTROL (12; SYS ГЕН OF (в) OPERATES SAFETY FAILS i— —n 5 S>::TT--4 Cr ? F,f 50: THE COURSE OF GAS Fi№ »'/WE ^HRQIJCH Г4t- /т.* /•«.. 'CR FOLLOWS 77TÍ" COLD" PRIMARY HF>/G RUPTURE 0(Я) GiO) 0,75 0,50 0,25 T -0,i5 R':iA7iv£ size of CURVES Thf HPING RUPTURE 4 ,1Ш- F .. 0,2 '*--*-•- 0,5 1,0 Li.,-:: -- F,f51. THE COURSE OF GAS Г LOW RATE . • w/?G/..'!?.» •'-•• Rl'^roR FOLLOWS! A RUPTURE OF THE JZOLC' PR-'MARY P,-": KLT;V<1'! NOZZí BUILT-IN >• to (.ÍUjtoly «die RELATIVE SIZE OF СУ/?И£Г5 7HE PIPING RUPTURE »,'# 0,2 2-.f 0,5 3.3* ',0 <>** 2,0 F,952: THE COURSE OF QAS FLOW RATE THHUJ ••€ REACTOR FOL THE „HOT"PRIMARY PIPING RUPTURE в(0) 0,75 0,50 t 0,25 RELATIVE S/ZE Oř CURVEb THE RPING RUPTURE '»'• 0,05 2:2* 0,1 3.3* 0,5 1,0 5,5* го THE COURSE OF GAS FLOW RATE THROUGH THE REACTOR FQLLQWH A RUPTURE OF THE uHOT" PRIMARY PIPING WITH A VENTUR/ TUBE 3b'! LT - IN e;o) Q5Ú 0,25 0 •— 4[sJ 0 50 100 150 100 fiCLATIVС St7E 0Г ZVP^F. 1HE PIPING RbPlURE л . ,/fM r F 1 0,05 2 0,1 3 0,5 A ""' "i.O If" Ftf 54. THE COURSE OF PRESSURE DIFFERENCE IN THE ^£AC- FOLLOWING A RUPTURE OF THE PRIMARY PIPING f^-2, Br PASSES •THfíc ТОП • яялг fíUPWPED BLOWER or Trtf и£*ЕЫ? /,y; PIPING VANES BLOWERS ThE » — %[,] Fif. 55 THE DEPENDENCE OF THE SPEED OF DECREASE OF GAS PLC RATE G AND MEAN GAS PRESSURE_д ON THE SIZE Q~ RUPTURE IN THE HOT PIPING RUPTURED (^..Ет.) (*l£« I - WITH NO THROTTLING ELEMENTS IN THE Р1РЖ 2-WITH A VENTURl TUBE BULT INTO THE PIPING mt" -L ?. ESTIMATION OF THE PROBABILITY OF OCCURENCE OF TSS EXTREME ACCJDEHT EVENTS 15M a*f.*tj of *h« nuclear power station is to Ъе s. :.tr*l f"t? all regimes of its operation». It may Ъе said ^Ьят p-pívii*ii th? p>wer station desig» is sound and put proper- 1 Г«.у in " p*%-t;\.-.»s then, under normál operation of the le- vЛ .ч*е nsltiwr the personnel nor the living conditions in t&« ís-fff.Tíadings woxld Ъе endangered by the radiation. Ne- 7*.r-^a*!*»*} the aixnrrmal accident states of thfj ..power sta- t-ť.ii ^A^sed by failures of the device, mistake of the oj>«- r-s^iíL* p*.--ía'*2Ma«.Ij, -:;т by an external disturbing action, may bn íang^Y.v;.!. Th* ar at part of the radioactivity of the reactor has its iM?,-*,gin in the prr-ess of fission of the nuclear fuel «ad tu.it is the reason why any devise of the nuclear po wer station,, the function of which is associated with eit her prevention or reduction of fission products release trtm th* tail into the surroundings of the nuclear power station», amy be considered as a safety device. Considered fr:»m the viewpoint of their function, they вау Ъе grouped it*: the following categories: - operating equipment which include the components, oo- r^řta end systems associated with the process of power generation (fuel elements,, primary cooling circuit, cir culating blowers, steam generators, systems of control); - eafety «ajjipment reducing the damage of fuel elements and -sf o1?h#r operating devices (systems for emergency thut-down of the reactor, emergency supply and cooling)| - -^fiXkbю rmmnb systems reducing the propagation of radio- «•*J/-;-. mnteriule into the nuclear роъзт etation surroun- •.5.,-.n;-.- iu *'h - 144 - 'l^e tbi'ee categories mentioned аЪг-'-'ь ir^ ••-• л- •-•:'-.*•- £-•?- ssardad %s three safety oarrltrs to рг^*в-м ;^e p:: '-a •-+s :" fi'.a^.^^n fr:'ia the.-т rei-aase cxb of tb? ; . :*atr, t h^ а-л- '.-f.=\'' p -we? is^at:, VE surroundings. Under .no;:'m?-* operatieg s.^i';'.'"^í the fission p-'^d.vts ?.ve, 'Slb-^"' i ti tb.« зр*'sí tina: iii7i'--4s (tha f^rst barrier) so tha^ < ^ 7 a very вл*?>Л • *t^d .--.see felly controlled amount cf radí--a •-tw* juaterials ни v ";.i-. ";inri. la ce.se of a failur* of said operating equipment tb<* syst«>n. of s^ie^y devices (the second barv-'qr) should act t ;.>' :IÍ'.'4:> (-..-; t. r*d-:*.e the dangerous «r •;•-:«£.% of radioacti- "-^ materials- The third barrier - the containment - is to 4it:..-,n if V-th *"he first end the second barrier simulta- lib •.'';.& -3 f'r\' I „ The prťVabi.lity of simultaneous rupture rr failure of :?;i t-h.e covrs* safety barriers determines the magn't'ids of the олщет nf a great radiation accident of a nuclear po wer station. It is obvious that this hazard should not be greater than those of great accidents in other branches of engineering already accepted by the public, ?;!.. TM prvbab.i Lity of failures of operating devi c*ts (the first barrier) Under nominal conditions the operating de-гл ..es operate .Ir. a ccnt.inuc.is manner„ Let us consider that the occurence vt a.r.-.;y fail arc is incidental and i ndepender.t on the other incidental failures of a device. The probabil:ity of tccu- ran^e of r -• failures in time interval • ' is g; ven by the Pv'iason law cf distribution /46/: p(ri --.-—^-j (3) Рог г - О / it is for the probability of failure-fee ope ration in time interval T /: P(o) -е-ХГ (a.) If the parameter Д denotes the number of failures in a unit cf time, then the number of failures in time inter val Г will be d=A-r (5) The probability of the not-failure-fjtee operation of the device is d f>=1-P(0)=J-e- l6) If d is very small (in case of a very infrequent occuren ce cf failures) the following relationship may by written: d PT ~ (7) Rsiating the parameter Л to the time interval of one year» then the probability of occurence of failure of said device in one year's time interval (therefore related to the so called reactor-year) will be as follows: řŤ *cí -X (year? "f (year") * Я This probability is called also the "hazard index" of a failare» Its values for the failures of some operating de vices as used in the safety analysis of certain designs, are presented in table 11 /14, 49, 52, 53/t - 3A6 Devi í<ň .а-а-?. •-"•"! .:ovi*o 3 ! MaLia supply system i ; Control system 10"** i /«.£,. Tli* prrbabiH + y of fnila^e f v-ii safety and con- tai-mft.:o,t d~vi -*s í.tb* a^ «ríi and i/nird barriers Under n'mal operation of the nuclear power station these devices a^e not -_?> a.:~". ;:it Lut this is suddenly de manded when abnormal .-p-^at-.ng * ' ^ist.! ?ns devfiop» РГОЕ the vlewpcibt of за?:»*-;/ -.s f g''"e-,t 1 лр-'-т-t-ance the proba bility of their fail'ire ř. wh->-h ^-.я^а th« probability of r--3carence if ше of the two possible output states of a giTsr. devi'-e - either fai • ira or cutting into operation ^n '-':mman.i« If a great п-дтг-чг of mutually independent "ex.f.r-x'iments" it: assumed, t>a I'^'-ijil;^ .:f failure of -artain device may be flxpress*} by nnsz-.s '* tin to n^miai Jew of distrlbuti-n /48/» ТЬл parameter cf distribution, denoted as p„ and express.1ng ^he relative proportion of the number of failures,. m*y f ? tb.e pn,rp?se under consi- d-.^ati-1^, be regarded t Mpr^p^i^- th- probability of fai lure of a device if it is '-•гясьчгЛ'эЗ to start action. Tab le 12 presents- in riant 1 tat! v o form the probabilities of failure of certain safety and ^xperxiaaroal devices /14, 52, and 5.3/г Table 12s I The probability I Device i >-;f failure | Main system of fast .0 * 10 * ; sho.t-iown of the reactor Secondary system of reactor Ю-3 sh-ťt-di-wn 2 Emergency system of injection ю- 2 Containment 10~ 7.3e The probability of multiple failures The terns "the hazard index' or "the probability of failure" shoald cover also the so called multiple failures which may occur on an operating, safety, cr containment de vice owing to the failure of some other operating device, Let an event denoted as A represent a failure of one opera ting device. and the event В be a failure of one safety sys tem» Then the probability of occurence cf both and В is the product of the probability of occurence of A and the probability of occurence of В in dependence on the occuren ce of A, so that Р(АГ)В)*?(М'?(В/А) - (в) - 148 Pr1 v.idsd >'?th ťaa eva.ats uM*r consideration are n<-t interdependent so that if the failure of the operating de vice A has no sabstantial effect on the ability of a cor- re.-rt action of nhe system of control, then PIB/A) * PCB) or, in another words, Р(АШ*РГА)'Р(В) C9) It is important that individual devices and systems should be designed in su^h a way that a failure of one of them does not ^ause failures of the other ones. 7-4, Estimation of the probability of the extreme accident situations of the A-l nuclear power station Let ив try to estimate the probability of occurence of the A-l nuclei:,' power station accident situations indu ced by the following initiating eventsi - rupture of the KS-150 reactor pressure vessel, - uncontrolled withdrawal of the control rods out of the reactor core, - breakdown of one or two circulating blowers, - loss of self-consumption supply of the power station, - rapture of one of the twelve main primary circuits. 7-4,1. Rupture of the reactor pressure vessel Т3а« possible failure consequences of the most reactor parts as well as additional devices of the power station 149 ^.TÍ 'Л«;л'.'..;. " " -" '*••' di *:1±1щ the davioes into partis" • imp myirVi^ •-. • (Í;Í.-IJ stand-Ъувз, ty tha s.^tion of safety devices -aridrf.ys-: .;.:.-з , at.;. Considered from this point cf view, -.-j ma-? *- č'-.;.d that a heavv failure of the reactor pi.-aosu-d "v ••'- - ií-- ал exclusion. It is beyond any qu^s- ti.0^ t.hr. ; .-;',..:-•; oi to.a KS~lrO reactor pressure vessel •» .;.Ci .тка • ' .... es •? +be ;;--.».Hng sf th/"! г-эа-.--£г..г се™ r 7'sa r'.^.p-:.-•-.•.•.:••. . ' .-.f ths safety system, and then tz the .:>ij._•'::'.:.:£, ,1' -. >. :-^;:~vr core with all consequences. That is the reason -л-гг;* it is vary important with the pressure •vessel о о ьесл>: <ц- extremely low value of the "hazard indax", or -,'•••>. -bahiiity of occurence of a large failu re * Titecď 4s л v':- лу smáli practical experience with the opsrstio.: ct ;-А-+;-г pressure vessels available so far, and the statri.c-.ics of their failure-rate does not exist. Helatively "oett-er sitaatior. is with the operation of stssuc. boi'l-:v-z /49f 50/• So far, for instance, the insu rance companies in the United States had in their eviden ce soma 530 '.--;.•!.' *^6 with operating pressures of the steam over 600 рв'я wh' •>•. bu V;T]Í.ÍI more than 4 000 so called Ъ&Иег-у^чггт, Благ, of them baing in operation more than 30 years* Th*ir statistics showed no serious failure of the drum. Xr- shoali Ъв aidsd that no drum under investi gation is made of tetter material against the pressure ves sels of present' Aaarioan nuclear power reactors. But direct appli1"^':'-• - f r>,; 3 íi.xp^ri-ínce to nuclear pressure vessels is ir/i'. p•• s* ;> ".••;, "'•'••-•4 -£)вя-я -pax-Vja ander different condi- Ыоххн 'i'iiip^j!s.'• .'• ъЬ::дв of steam boilers. In this connec tion, the " or^-re.-rii efvact of ЪЪч radiation should Ъе men tioned. The / — -0 r .^ctcr pressure vessel is designed for the fol'.:.••:•:,-'. = op--'"eting parameters of carbon dioxide: 65 kp/ca , 112° С. As to its shape, dimensions (4.8 m in diameter, 20 m in height), and the thickness of the cylin drical wall (150 sua) it is similar to a boiling weter rear- tor pi'essure ves<"vV with approximately three times higher thermal p.iwer (se# the review of reactor pressure vassals in /50A Is 'comparison with pressure vessels used in the w.?rli f=:r gas and water cooled reactors an important diffe- тяй-уэ is the low operating temperature of its wall (some 155r 0 «tga-l-nst? 25$ up to 370° С with the vessels of foreigc riglr;. That is why šach a high attention has been devrtei t the pr-jblems of brittle fracture in the course of Cse- •z-hcslovaic pressure vesel development. This concerns in the first place the determination of the crack-arrest tempera ture /55/. In order to verify the manufacturing technology and t- indicate hidden defects, a shortened 1:1 model of the pressure vessel has been manufactured and subjected to a pressure test (400 cycles with a pressure from 1 to 65 kp/ /cm). To ebtain information about gradual radiation effects, specimens of the basic material as well as those of Welded joints will be located in the X8-150 reactor. On the basis of the foregoing arguments it appears jus tifiable that the KS-150 reactor pressure vessel should ha ve the same operational safety aa those manufactured abroad, so that the corresponding "hazard index" should be (see tab le 11): Vessel) = 5 . (10^ - lO"8). For the probability of occurence of a failure in the course of its operational life (20 years) it then results: POL. vessel = 5 . (1СГ7 - 10"8) . 20 * * 10~5 - 10"6. - 151 - Aií-í-or-ding tc /50/ some experts propose in the recent; years That periodical pressure tests of reactor pressure vessels should Ъ-j ^гЛя > Their main advantage would con- sx*t in a red ..:.t:L .-_ of vessel failure hazard for a shcr- t--a:;' pra-set tims intervals and in the possibility to ta ke идя nesess- ;y steps *: increase the reliability of the v-sss-sl. As the main disadvantage, their high cost should he n-~ted» 7-*+«2- Uaccrr-ro.lled withdrawal of the control rods out r.f the rsa t:r errs The «i.e. tronics "f the neutron density automatic con troller c-rts'ns svaw 1 000 parts, two of them heing of a verj high imp••-."tar.••••5 sirce if at least one of them fails fcha oncost-rolled withdrawal of the control rods out of the reactor core would result /56/. Provided that an averaged aiean time interval Ъ'-ítwefn the failures of an electronic př?.rt T,_ ma„ - 50 h /56/, the frequency of failures of one of the two said parts is as follows: X - 1 ÓO0 ?30 - 5'5 • lcrl /7ear_1/ so that the corresponding "hazard index" is P'W> * i ~ •'* ~-2 - e"3,5 * 10~ s 2'9531 . Ю-1 In terms of order this estimation is in conformity with, the уя!>ле of the hazard index of control system fai lures affording to tahle г, 1. ТЪч f ", •• :wi:ag equation valid for the probability of ад шк-jntr-li-d withdrawal of the control rods during the ES-150 rector IT fe-time (20 years) gives - 152 - P 1 e s 0L, wp. "- " °*99909i The pc i.-'t-ab",'. a amber if uncontrolled withdrawals of c\;.ntr\l rods davins-T the 20 years of cperation *,.„<, ~ 20 •* » 20 , 3.5 . Ю"1 = 7 (1/20 years) '.r-. thess sitaatims it is the safety system that is *:•: avt to эЬ-л^-А^лт rae .rea^t^r.. Let us estimate the pro- ba'fcli:.xřy of safety system failure according to table 12: iace.) * ^naer the assumption that there is no interaction betTveer. the central system and the system of safety, the probability of an uncontrolled failure of the control system and ч simu^aae^as failure of the system of safe ty during the assumed 20 year's reactor life-time would be This valu* of probability should he considered as conservative because its deriving did not take into до со ant th-з periodical inspections and the operational trusts of the system of control» 7*4.5. Breakdown of one or two circulating blowers There are no data available concerning the failu re -rate of the A-l nuclear power station circulating blowers* Let us consider the average time interval bet ween blower failures causing it inoperable in the ran- Se yv. от, - 2 -• 12 months (it means 1-6 failures a year.. 153 ТЫ corresponding hazard index of a breakdown is *7--*-™MT^ = x ~ a"(i tC 6) • °»S5212 to 0.99752 Breakdown of one circulating blower is not dangerous i- the realtor (see the section 6.2.). The hazard index :~ -, simultaneous breakdown of two circulating blowers h'.2 M*««.> ' ^(blower) * 0.39957-0.9950* -..-,:; tí.-э corresponding mean time interval between the si- aaltanaous breakdown of tWc circulating blowers is T2 blowers, mean = CO.189rl.96) year, •3-; 'Liat during the 20 gear's operation of the reactor so- x;ir. 10 - 106 cases of simultaneous breakdown of two circu- -/;r>g bowers woiiid o^cur. Let us consider that the esti- nst;! order range of the number of breakdown cases includes r - the failures of one self-consumption e-ction distribu- . • whioh may lead to the breakdown of the two correspon- i;; -: circulating blowers» If two circulating blowers are i -л -j. arable, the reactor should be shut-down by the system Л «safety. Provided the operation of the safety system is v.."- dependent on the breakdown of the circulating blowers, •The probability of realtor accident occurence Cduring its 20 veers of operation) owing to the breakdown of two cir- - • atr'ng blcwers is given by the following relationship ?0L, (2 blowers, ace) * <10 to ^ * W"*. fco 10~Ъ * •в 10~2 te 10"*. 154 - 7-4i.4k Less of power station 3elf--^oi>sumption supply The initiating, event consisting in the 220 kV net work failure with subsequent loss of voltage in the 110 kV network and the runaway of two of the three tur biness has not bean analyzed so far in the analysis of the A--X nuclear power station safety measures. Let us consider, in a^-ordance with table llř that the hazard index of a failure of the main 220 kV system is 1 P,22C ^y. ••--• Ю" , so that the failure probability rela- ted t-j the twenty years of operation would be IVQT 220 kV)" ••= 0.8?» According to /51/ the 110 kV network failure pro 2 bability is PQ:Q kV^ - 1.65 . 10"* . To simplify the si tuation let us i'-ns.id*r the probability of at least two turbines runaway, ^(^urb ) ^ •*"* ^en the Probability of self-"consumption supply loss in the course of reactor li fe-time equals approximately PPL, supp. = P0L, 220 kV P(110 kVJ * P^turb.) s я 0.87 - 1-.65 • 1 У Ю~2, In this accident case the reactor is to be shut-down by its system of safety. Considering the safety system as independent, the probability of simultaneous failure of both the loss of self-consumption supply and the failure of the safety system would be P 10 2 t0 10 5 1СГб 10 7 0L (suPP; ace.) - ~ * U°~* ' > * ^ *o ~ >- 7.4-J.5* Euptare of one of the twelve main primary pipings Statistical data of piping ruptures in the chemical industry and conventional power stations processed in dif- - 155 - ferent co-ontries show a relatively fr^qasnv clearer» :e of з-ач.1?. vxpbúT'13 bit., s.'^j.i'^an-r^'i?'.;; •?. "V•;-;,-.. j.i'>frsq;isat. oc curence of larg-9 r-ptures due t;--- ov ". .s i vacfare, Accor ding t-? /43/ several thousands of hi.^h pw,•• ? -:e рлр!-:й£ i':, Anvsriiiar. р?^е.г štaciras have C^SJC и-.".аа'".----л va; - stati g- fri ?.з r-3cord.il f-'-r as long as 30 years. 0::...v fo'-i- v-.=Lses of hrí4"tl* f??a^ťi7'ň !въ±:1:щ т: tL. f'u.11 op-s:^-ur oi" tha piping ani. vhA d: 3 ^ha.'-ge cf *he medium through rh;.. douM* ;-:;.vss-' -SÍ:Í:;C area ha^a he^n records!, Ti^r? is ro :9ta"»-i3ti'".-8 oa p:p.".ng r„;i-ÍVÍS in nu.oi*a- pc-wye static аз so far Ба- it eeems ž'\f::ii*JoL*, on the I: a sis of experience from the operatior г? high - pressure рдрл i-gs in the conventicnaj power stations* that the ma ximum value of hazard iad?x according to table II may Ъе P, - - • . -.rr*. -», Piping ..! Corresponding pr-VbaMl ;< ty -f primary piping rapture during the reactor lif^-titc--'' (>0 .years'', is P . r 2 'T2, 4)L, piping ' As stated in section 6»4«, the r';.ptu:-ft of one primary piping Of th« B3-150 r*=l-s-?t-0r W'-.il J. 1«л:1 to .t; I'vi-'iistribli- tion of the dynamical forces a-': -:о..ц< or. ЬЪч internal parts of the reactor, Thus,, it is not fully ,ňast:.fiahle to rule .o-At an influence of the piping rupture -.^ tha compensating or safety rods to fanotioa proper! -j, Tf w-? use the rela tionship (6"» to calculate this ^rder of the probability of s multaaeoas primary piping rapture and the failure of the safety system daring the 20 years operational reactox li- f3--tiiu<3f we obtain (in terms of order); P iU i( 0L; (plpirr:, ace,) ~ ''•' -' - p6 The first value expresses the safety system failure due to piping rapture, while the second one represents the independents of safety system action on the rupture of the piping. 7»4„6. Evaluation of results Tha fЛlowing table presents the probabilities of occurence of the postulated accident events during the KS-I50 reactor life-time. Table 13 Th« probability of reac Accident event tor accident (during 20 years' operation) 1 Rupture of reactor pressure lO"5 - 10"6 vessel Uncontrolled withdrawal of 10"3 - 10"4 control rods Breakdown cf two circula io-2 - ю-* ting blowers Loss of self-consumption 10"6 - 10"7 supply Kaptx?e of one primary 10"^ - 10"7 piping i t - 157 The consequences of failures mentioned in the fore going table are not tha same for all :ases. Obviously, the mr>at dangerous would be those associated with, the rupturs \f р."-ч*за-?ч T^fSM" and the primary piping. In the remaining tirea* .•v.ases there would not be any '. >se of primary coolant рт:з-ssure nor an extensive Гч-ilaase of fission products b*yval the psimary c-rcoit wiúld е>:--**1г„ List -JLS tzy t: ооараге ths mortal dangee associated with -1?;!? KS-150 react».*? accident v;itk t^ht caused by a traf fi; %7M.SL.5 A-iiorling t'j /2/ the г-a would be 3 4-00 mortal ca ses dus to tb^ most dangerous адг eríil credible reactor (500 №Vt) a'cíó.ent under very adverse climatic conditions. Ascordiag to statistics! data there are approximately the same number of mortal eases 1л the Czechoslovak traffic du ring twi year». From this mey bs ds4-a*ed a IXr - 10 times higher probability for a Czechoslovak citizen to die due to a traffic i-raeh compared with that as*ociated with the KS-150 reaeter accident. 7.Í.5 • Summary 7.5.-Í-* Th« estimation of the probability of occursnco> of ac- '..Ident states in the prototype A-l nuclear power station is possible so far only on the basis of pub lished reliability data of devices of foreign nuc- Isar power stations or on the basis of our own very rough and queetionnable estimation. Therefore the results obtained are to be regarded only as a help to enable orientation. 7.5.2,. It is obvious that the most serious consequences wnuld be caueedby the rupture of the reactor pressu re vessel or the primary piping. - 158 - XArS'Jng i.ati а-.-сс aat a higber q;u tad probability of pvi^.ry p; p.liffs -.v.ii-=řui',9 in comparison with that of ~b.e r-ia.t ••, press,;;^ vessel as well as the fact vj.at. VÍ /-••. -•'• •:••-> KS- ',50 reactor it is not possible ¥ £.*.., .i.:-.-i /•.."..!у nhe -.ridxngsr of safety and compen- s.-r«ti..T*s r " 15>9 8. CONCLUSIONS 8*1. The advanced method of the evaluation f ni ea~ p wer stations safety which is in use atrial i.u va.;*mb years is based on the quantitative estimation of ths radiation endanger^ent • This endanger ment depends on the probability of reactor a^ci&^zxz occurence as w*II as on the extent cf its consequences. The appli cation of this method requires that the reliability characteristics of individual components of the nuc lear power station be known* The same g?-es fc>r the characteristics of fission products release and pro pagation under abnormal states of nuclear power sta tion operation. 8»2» The choice of the method of the KS-150 reactor safe ty analysis is influenced by the following specific facts; - the KS-150 reactor is a unique one in the ffgbrld concerning both its type and parametersь so that the utilization of foreign experience is very/Uimi- ted, - the equipment of the A-l nuclear power station con sists mainly of components and apparatuses opera tionally unverified so that their reliability cha racteristics are not available, -the characteristics of fission products release out of thé ES type fuel element under abnormal conditions are not known. Adopted method of the KS-150 reactor safety ana lysis, which is essentially qualitative and limited by the region cf accident prevention» proceeds as fol lows J - 160 - p^stulstion of the extreme initiating events, - analysis of transient phenomena in the nuclear po~ wa?' station caused by the extreme initiating events» the parameters #f reactor control and safety heing cheeked and verified, analysis of the possibilities to reduce the acci dent 9xaursicns manually, - design and discussion of certain changes of the •---. •.-instructional arrangement with the aim *f iiv reasing raactor safety, - evaluation of the order of probability of the extre me accident situations occurence, made with the pur- prse to provide a quantitative idea. 8.3. 5Dhe analysis of the accident situations of A-l nuc lear power station has been made for the following initiating events: - an on-:.->ntr.vlle4 withdrawal of both control reds at a maxlmam speed cut of the reactor core, --• breakdown of one or two circulating blovsrb, - loss of power station eeif-eonsumption faxpply, - ruptar* of ose main piping of the primary«circuit. 8.4. The main results of the analysis of the A-l nuclear power station accident states: - Provide! ал umaentralled withdrawal of the control r*is out --t the reactor core occurs, then, for no minal power operation, the reactor may be shut-down effectively by the safety system induced by power 6X04S6. Pallure of the system ef safety would result In this case in the melting of the prevailing part •it fuel elemente. For low power levels the system cf safety Indueed by reactor period is effective;