ML19268B852

Investigations on Comparison of Greatest Possible Accident Consequences in Reprocessing Plant & in Nuclear Power Plant: Critical Comments on Work Rept AB-290.

ML19268B852
Person / Time
Site:	Allens Creek File:Houston Lighting and Power Company icon.png
Issue date:	11/30/1977
From:	Bachner D, Farber G, Holm D GESELLSCHAFT FUR REAKTORSICHERHEIT
To:
Shared Package
ML19268B851	List: ML19268B852
References
GRS-A-59, NUDOCS 7906210421
Download: ML19268B852 (24)

Text

'

. . , ;'. s  ;

- ,. _ g g gT,.M N" / . I:

^ [*/g b WC PUObb, j

. , .,9 m

p rg)) b' I a- .

\

'4EC Iran 513;ien sais c 4' ,. .

- s.-  !-

?/b N'x.

~

.

• y'

, ORS i Gere11schaft f;er Reak:Orsicherheit (GKS) mbH it f

(Socie:y for Reae:cr Safety) 0.10 3"/ESTICATIONS ON THE CCl?AR:!CN CT f:

GREATIST PCSS!!LE ACCIOENT CONSEC"E'4CES IN A REPROCESSING PLANT A'*D IN A NUCLEAR b

PO'JER PLANT . {i t :.

• , s t ':.

(CRITICAL COMMENTS ON *.'CRX REPCRT AB-290) 2 h..._ ..

l t ":....x:. -

(Stamp: 15 June 1978) -

m.:

y. }, = :.

.- s- e GRS__- A - 59.(NOVEMEER 1977) pf L.i..

e= ,

x 2* W w$$'8' 9 .;

Dr. D. Bachner, G. Tarber, D. Holm, C' 8 S' b

%# $9 '

p:.

i' Dr. P. Neus se r , 'a'. Ullrich i.M.. _.i.

m Mn C :: ken;asse :

'C:0 ::1:;r.e .

ele:n=r.e t:::12 ::53-:

Telex i 53. ic ;rs : -

95005215

's' 9 0 gg 3 0 4,

. .. - _ _ . / [

i er

/'

S'JKMARY Work Report 290 of the insti:ute f:r Rea:::: Safety of the Tech- "E i

nical Supe rvisi:n L'nien e.. (!RS) showed tha:, in the even: ef a

1. e pos:ulated disregard Of all safety installa:1:ns planned f:: the Oe- ' " ' '

c:: issioning Cente (SEl) : presen: in the nuclear pcwer plant, the danger potential f or nuclear pcwer plants and f:: :he Oe:0 is- g .,,

sicaing Cen:e: is app: xima:ely :he sa:e. This gives rise to the b

enclusion :ha: :he safety installations f:: a nuclear de::::issi:n- ...

ing center, mere parti:ularly :he af:erheat discharge syster, mus: ,...

ii satisfy ::= parable criteria as for a nuclear p:ver plant. It was expressly determined in the reper: that the nucerical data given en the radielegi:a1 loads represent no risk state:en: under realis-ti: cenditiens and that an evaluatien as abs:lutt s: ate:ent is  !' '

inadmissible. In the same way, the numerical values are taken ever uncriti: ally f::: varicus sides, published and cr :ne:usly in:er- L preted. [

11 The A3 290 c:ntains, as was stated en the title page, only [

previsienal results. The Society for Reacto Safety abh as succes- .s.

24 so; s:ciety of the IRS submits in the f olleving a reworking and detailed explanation of the report whereby the latest state of art and applicatien data are used as the basis: f U

The design data of the nuclear dece:missi:ning [

center.

F r.75.? *:

~.'

The first results ef the Cercan risk study, i..

The lates data for release, spread and dese n ~

factors.

The results of the thermodynamic computations of :he extreme MF

.j:-

imp::bable cooling breakdevn in the fuel element receiving pools and in the high level radieae:1ve vaste centainers can be summari:ed as follevs: .

In order :o keep the water level during evaperatien to the /Il k.

level required f or aveiding a hea:ing up, the f oll wing replenish- h e.,

n men: is required:

b Tuel element receiving peels:

es. 50 m3/h after 10 days VaSte 00ntainC:s: 2*

ca. 21 :)/h after 2 days.

+

• hu :ers .n :ne ::;n: car;;n in ::a:e p ag.na:::- n :ne ::ig na; :ex:. t

' V\J

~

Q f )

W ox

' dem>..w--

Accordingly, there are either 10 0: 2 days available in ordct '~

to replenish the relatively slight veter quantities f preventing melt. These c:olan quantities per time uni are so small that they can be even brought by tank :ar transport. In :he same vsy, an 10- ,.

prevised hese conne:: ion to a lh" to 2" line (for exa:ple, fire s#s hy dran:) is sufficient to supply :he required coolant vster quantity at the customary pressure level in the supply network. ,

(

i' Tar :nis reason, a melt acciden: is ou: of :he questi:n for the ,

fuel elemen: pool and the 11AW centainers is out of the question F I

since the possibly required emergency measures are easy to carry y cut. p.

i  !!

A disastr:us fission product release and the high radiati:n ex- ,.

pesures ::nne::ed therevith are, for this reason, imp ssible. The w U

corres; nding table . in Work Repor: 290 are f:: :his reasen not

. suitable f:: taking state:ents en the radiologi:a1 danger. .

-E l

A nuclear tel: ac:1 den: in the rea::ct is not censidered in the Atomic Legal Li:ensing Procedure and is net centioned as an event Oc h be considered in the standard Safety Report (" Arrangement of 1: ems [

and Organi:ation for a Standard Safety Report f or Nuclear Pcver L Plants with Pressurized Water Reactor or Boiling Water Reactor", i-appearing in the Joint Hints:erial 04:ette, Edition A, No. 6 of b 30 August 1976, pp. 418 ff.) since, as a result of the costly emer- b; ,

/1II !.k; gency cooling system, the probability of occurrence for such a case g

is so small that it can practically be excluded. Nevertheless, if' i

radiation exposures were computed or taken from technical 11:erature .r. -

for just such a sequence of events (Chapter 7). g l

p The own new computations shculd show, for the sake of example, l

that the resul ts published in Work Report 290 cannot be used for p absolute statements with respect co radiological danger. Ep. .

b 'T. .

The dif f erences with the results of other authors (WASH-1400, fj;

="

KFK 2433) can be explained by the limiting conditions solceted.

i More exact, risk-relevant statements or evidence must be tsken f rom the Oerman " Reactor Saf ety Study" presently being written.

As has aircady been explained, a core melt is practically out of the question in a nucicer power plant. Por :his resson, computa- e tiens of ::diologi:sl effects are also not required. If such state- T="

ments, however, had to be made, the relative numeri:n1 values men-tiened in AB 290 need net te used bu: :ne resul:s achieved in Chapter F(

7 of the pres en: report. :n order :o be able to make a s a:ezent concerning the risk ::nnected vi:5 a nuclear pcVer plant, i: is n ece s s a rv :: analy:n n:: :niy the scope :f damage bu: als: :he 4

D 95005217 e

" w -. ,

p:0bability of e::urrence. These risk-relev:n: state:.ents are cade --

by the Oerman risk study presently being werked on.

Society for F.es::or Safety (GRS) =tii (Signature} (Signa:are] .

i Ke11er ann  ?:cf. Or. Birkhofer s i

i r

e-i

[. .'

y8. .:

t i:;: ..

l :::

' ' ': t

• e-3 4

,..a.'.'.

l*.id'.

$~. .

I.S.

g ... 7.

i.

1.'..

e 95005218 es

.-+ wM esag g e .N ew es.e . e e ees --w-

. Ae ,.e. e,, r. r- .,. . e_ s. . >

Face StiO'.ARY ,

. . ,, .R ....w

. . . ,. . .4 3

. .avr

.. . . ., .,. 5 ... ., . - J.. w.

. ne. R..

, .... .y 0

.. ,.u..,4.......s

. w w .s .

i

~

3. INVESTICATIVE 'm*CFZ 9

.s.r.a w-e.

.. .e h. e.r.. . .+ P . . . . . T C C O LA . . C + v...

.. . . .i + .. a-

5. TittE LAP!ts S THE Asso'ED C0ct:SO 3RE.w;CC :S THE T'.'!L 10 ELE".EST F,ECE;VISO PCCE A';D THE HA'.' COSTA;SEE CF THE DE-C0!:".lS$10:i!SO CESTER . :=

16 E

6. ACC10EST CCSSIDERATICSS i;

7. RADICLOGICAL ETTECTS 17 i ..

v:

1. '

IJ EX?I.A'4ATION OF ABEREVIATIONS 22 Id+

. u:.-

BILLICORAPHY 23 9.w.

.F.:.:.

e

.'.7: '.

h I

i f .'"

~

95005219

..>wa

/,-

4. ....r

.,. _ . , CT . ... .. ..

The inter?.a1 */erk Ee; ort 090 had :he task Of :: paring the danger potential Of a large reprecessing plan: vi:h :ha: et a nuclear power plant. The g:a1 Of this ::: paris:n was :: determine whe:her ::: par-able safety insta11a:1:ns are ne:essary f:: the re;tecessing plan: as for a nuclear power plant. Fer this reas:n, the :entra::ing ac:hcri:y specified tha: the investiga:icns v:uld :e carried :ut using the lie-iting :endi:1:n that no safety systems are either present er effe::ive.

Ac:ordingly, the probability of occurrence Of su:h a 6equence of events was also net considered but a :::a1 cooling breakdevn was as-suced which always leads to melt and, in the case of the specified limiting eenditi:ns, to high radiation exposures. !n the case Of accident analyses for a nuclear power plant in ::nne::1:n e.th the A:::1: Lav L1:ensing ?recedure, a :celan: 1:ss a::ident vi:h subse-quent core melt was not censidered since :he probability of o::urrence for such a ec=bination of events was so 1 v tha: it can be ruled out of the questi:n. In the case of a coolant less accident, safety sys-tens are used whi:h feed water into the reactor pressure vessel in order to again raise the water level in the pressure vessel and ensure the leng-ter: coeling. The folleving syste: functi:ns are involved h e revith :

p.:

High pressure feed, i' E..

Accumulator feed, til The low pressure feed for flooding and subseiuent circulation activity.

Oepending en the specific accident (large, medium-sized er [..'

small leak), the above-described safety installatiens are placed in operation. As a rule , four subsys:e s are specified for a required system functi:n whereby the effectiveness of e:ergency cooling is "-

ensured by the operation of two subsystems. On the basis :f this design principle, it is used as an assumption in the licensing pro-cedure en a worldwide basis that these safety systems are sufficient to avoid a core ec1:down.

The quantities used in ::=puting the potential radia:1:n expo- /1 sures, for example release altitude, pr:paga:icn condi: ions, etc.,

had therevich only a medel character and were selected f;ce the viev-y peint of " comparability". They are not suitable for achieving a 'E:

i realis:i: esti ate of radiological ef f ec:s.

, u.

Fur:her, f or ressens Of a simple r :::;arabili:y, an ine:nsis-t i

i tency in the irves:1;a:1:n was :aken in:: :ne bar;ain inas uch as the thermal engineering :: pu ati:ns st arted f r: an ale:st-1..:a::

building vnereas the ::::uta:1:0 :f :he radi:1:;i:a1 effe::s started t

e 9

95005?20

'd hM ww- gm en e

6 for pra icai purpeses from a faulty building. *he buildin; stra:-

tu res , as s af ety :entainment , were not censidered in the :: puta:1ons f:r A3 290 alth: ugh they ever in a deft:::ve sta:e :learly reduce the activity relassed frc: the core en the way to the outside by deposits and similar effe::s.

~

For the abovementioned reasons,1: is a::ordingly ee ;1etely in-admissible :o value the radiation expcsures centioned in Werk Reper:

290 as an absolute s:stement or as advisory p si:1:n :n :he radio'.cg-ical ef f ects to be expected in accordance with the abovementioned .

events.

In the mean:1:e, a :o=prehensive safe:y repert f:: the planned nuclear de:e=-issi:ning center is available. The data of the nuclear de::mmissioning : enter are essentially differenti ::d fr:: :he ft::1-ticus data f or a reprocessing plant used as :he basis in Werk Report 290. In the -eantime, there is available fer the decer.missioning center a jcint pcsition of the React:r Saf ety 0 mmission (RSK) and 2+

the Radiation Frote :icn C : issien (SSR) in which both ::::issi:ns make pcsitive state:ents on :he safety-engineering feasibility.

The following-described events show tha: a meltdown accident is out of the questien in the case of high active vaste centainers I(

(RAW) and in the case of fuel element receiving poels since suf fi-cient length of ti e is available to feed in the relatively slight y" a-water quanti:1es required for centrol. For this reason, 1: is un- /3 ;E. .

necessary to compute a :1 dent deses for a ":eltdevn situation". s F

The core meltdown acciden: in the nuclear power plant with pressurited water reacters (Ok'R) is to be ruled out of the questien . . .'

as has already been explained. Never:heless, with respe:: to pe:en-tial radiologi:a1 consequences, ec a recent results are reported which :ake in:o consideratien compu:ations of ether authers in order 05-to avoid a further misuse c.' the numerical values intended to be relative which were menti:ned in A3 290 and to indi: ate essen:ial "i- -

' dif f eren:iations such as short-ter and long-ter deses.

I il; r

k...

""E...

[ ; ^ ^'

--~~..

S 95005221

• - ~. ,

'm' ^w om o

t iD!!!!NC COND:TIONS IN k'ORX Rtp0RT 2 0 /4 1.

The foll: wing li iting conditier.s were spe:ified by :he centra:-

ting authority:

The initial assun; 1:n is :o be :::a1 breakdevn _ ;"

of all cooling and ventilation syste:s. _

The release of radica::ive substances f rem :he ecitdown results without redue:1:n mechanisms L.

directly into the environment.

The release takes place near :he greund.

Addi:1:ns11y, the f olleving :ce.servative -easures are taken which led c::puls:rily :: a f urther esticate of pe:ential ef f e::s:

An infinite irradiation time was assumed for the fuel elements stered in the fuel elemen: storage, s ,

The heat release by convection and the heat accu- -

mulater capa:ity of the building were disregarded. t-u The heat'ing up of fuel elements starts after '  :

evaporating up to fuci element upper edge. y&;,f+

r The release f acters were assumed conservatively. e=:.

The release duration was negligibly shert. jg.7.;

m :-

The prepagation was computed conserva:1vely with as.x

'"7~

the parameters of Pasquill f or stable and addi-tienally neutral weather c nditions, es 2.=

--~;

Increased dilution owing to swirling in the build-ing a: the time of release by deposi: during -+ +

33 transport into :he atmosphere as well as by flue-tuations in vind dire::1on was not taken into E- -

consideration.

The mest unfavorable values were always used for l the dose f ctors. These involved rn:hcr cid d :: #*~

in which the mest recent inf ormation v:s not yc:

taken in:o censiderati:n. -..:-

Qi 95005222

- ~ . _ ._ . _ . . . . . . . _ . _ . _ _ _

b 1

1

3. .

"'/E5710AT:'/I SCOPE l5 ._

centra:t The :enservative f a:ters used as a basis in a : rdance are :: prehensively explained.in Work Re;crt 190 :t for with the reasens of sir.;1e :::

insa y p'"

described which systems and c epenents 290, it is have to be a devn in spite of reliable design before there can ecas breken ened up in the fuel element or vaste container. cur any heating the n0v vell-knewm data frem the defor the pestulated of : eling), using event (tota

~5' sequsi es for the heating up process:cm.?lssiening : enter, the time required for preventing meltdevn are redeterr.inedand the cooling water quantities On the basis of these time and quantity data given of th.s possibilities available for reliably , a descriptien prev teltdevn is

, enting a::1 dents in the fuel ele:ent receiving pools and incentainer, the FAW

h. ~

le In order to arrive at a technical evaluation e sequencesof th s

of statements and design data to arrive at made in the Safety Report e basis ant statuses concer

!U danger cor.nected vith these sequences of eventsa qualitative statement concernin 1:ed water reactor, the latest results e are present dFor t pressur-although this improbable and, for this reason, is net ,censidevent procedure. is quite ccebin ered in the licensing I pg*

l R g

Qx

e. ._ ._

i 95005223 e .

m * . . . . . "'

i

\ .

I i

i i

w 4 SHCR* ttsCR!? TION CT 00CLA.'!? C!T.CU!TE /6 The fuel element storage pools in the RA*.J containers are fitted with cooling systems fer rer tving af terheat. ,,

in the case of the fuel element poc1, pool water is pu:ved to the heat exchanger using a circulating pump. The heat exchanger dis-charges heat to the inter:ediate coelant circuit. From the inter:ed-iate coolant circuit, the heat finally arrives at the main :ccling water system through a second heat exchanger. The main cooling vater is reccoled through a vet eeeling tever. Replacement water is removed item a self-contained c cling pene. The coelant circuits are designed

! as 3 x 100% circulations. In (1 eut of 3) eperating mede, the pcol j te perature is kept, according to the Safety Repert. to ca. 'O' C.

l :n the case et a (1 cut of 3) mede of cperation, this temperature

, rises frem 40 to 60' C.

The standard power supply for the cooling installations takes place through two separate independent feedins. Ecth censist of overhead wires which are supplied by separate nodal points of the i p l an t . In the event cf breakdown of cennection 1, the decem=1ssien-ing center is further supplied by connection 2. Only in the case of the quite improbable simultaneous breakdown of the two independent feedin systems would a situation arise where emergency power drop e occurs.

t In order to supply the safety-engineering important censumers, a diesel emergency power plant is specified which is divided censis-tent with the process engineering concept into 3 x 1000 e:ergency cooling loops. Each loep is supplied by two diesels with 50%.

! The cooling installation of the RA'n' centainers is built in the same way. Nevertheless, a coolant circuit is lacking. The intermedi-ate coolant circuit takes the heat directly from the container :entent to be cooled using an overdi:ensioned cooling loop. According to the Safety Report, the heat discharge installations correspond to the single error criterion.

.1. . ..

m 0

95005224

5. TISE LAFsEs IN THE assumed tooLINO L;.E g;07: In Tat r';tt ELE 32xi /7 D

y _r C r '. ". '. ". . " * ~ ' '. x .s' ". D ~. 2. .r. "s'.,- 'vv^"..'.A'.'."...

. 07 ~. w. . .r 'a' ". ' ~. '. 5 5 t 0 ". '. ". v' CESTER The fictitious plan deal: vi:h in Work Reper: 290 differs es- '

sentia11y f rom the data Si ven for the de::==1ssiccin-* center in the -,

Safety Repor:. The f olleving table contains a number of the differ-ences decisive for the heating-up process 2 Fuel Elemen Storage Tool

'e:canissioning 3 ,9g Center Number of s:ored fuel ele- 1400 6544 ments i Jecay time of fuel elements 200 435 (d) -

Af:erheat capacity (MW, 25.4 31.0 [,.["

(infinite irradi- Eh..

7 a:ica :1=e) s.E bb c.

Water volume in total pool 12038 13806 lis:1 to be considered (m3) EEEe:

v::

r.= ::;; -

Total pool cross section S42 (vich subtrac- 1062 hifE

(=2] tien of discharge lT pool) [

Difference in height avail- 10 (evaporatien 10.5 (evapora- =.

sble for evaporation (=) up to core upper :or up to core gd*6 edge) eenter) i p ..

Waste Container jf.p:

l b=::.

i Specific afterhea: capacity 16 12.8 @k of vaste container (W/1) ](([

Wa

er volume in total pool 339 339" cc.c._.

(:J) j.A...

~

to be : nsidered T::a ;oe: -.ss se::1:n : 2' :3c  ! 256-

"These da:a ref er :o the vas:e ::ntainers w :'- .a::i :: hea: ::pa:i:"

95005225 s

,a . ~m., . . ~ . - . -

E" .. .

4

\

\.

pools can beDisturbances in :he ,

cooling processes turbed cooling be coped into poolwith by shifts fr m feel s wit

"" a carried out of water is through compar:h intact cooling of elementindividual pools with whereby fuel idis-element f-is kept

• at s:andard operating valuemade . with cperationa Further, replenishmentnsula:1on "'

lease of radionuclides f cant in the temperature Herewith, the temperature s or limited an impression to 100* C. The re- E and not improbableconcerning the time Neve r:h elet s ,

available. rom the fuel nt into the coola I is insignifi-gard of :he above cocooling distu sequence in :he in order to E' case of an obtain commissioning cen:er,un termeasure,rbance, u pool:

there assumed ernere the foll: wing using picture thef ac:ual da:a forres lts wi:h a disr a de-Heating up to or the fuel element >

Evaporation evaporation temperature of coolan: ca. 1.3 d ""

Adiabatic heating up to inventory start ca. 9.3 d (1700' C) mel:down a" y Heating up to meltdown into art st ca . 1. 0 d 3;r=

ability of concreteconsideration cap- the storagetaking ca. 21d 7.C[

time-limi:ed t ranspo rt irradiatiThese o

through a ven:n period (thra taking in e rmined time data were det ""4 vg.wE

=

a only affect meltdown on) was no: startof nonfunctioning The e outsideconvective a he t ventila WE

lla:icn system cannot beon case n=="

7p with increasin system (seals, a nonessential basis. considered g :emperatures, damagesassumed would in the heating uo 9?

sin more. flexible An intact phase since ven-the capaci:y of connections, cannot be $_.

a meltdown but only lengchtheven:ilation thermal expansion)avo'ided intac: hole on the "

Further-en time system I

The heat the following si:uatioengineering computation prup to start n

as, for example, pumping ut oand even here possibleovides fore was:

( '}

t~

Ucatin; up of concentra:into intact

o 100' C e disregarded: containers such wercou

[h EF c from 60 Evnpora: ion of C3- e th0 C 13 d Nea In Concentra:e /':

jacke:g conten:s20d eVape!a'ing :he eccl:ng ca, 1, 7 i 00' ,

[ . s .er, s s

'W W e t4 "*-- 0, 4

we.

9...*.* a m

w 95005226

--w- sus.en ~  %,w..e.. i..>w-o. ~e- -- -- w.. . .

Also in the cooling withou t consideration givsal:s) case of a total b, a meltdown (residual is only "7-en countermeasures.reakdowr, of -

up phase The scavenger air system is as althcugh it has no significancerespect withsumed to be intact in the heating In to heat removal, I lrequired:

level required for avoidingorder to maintain the a heating up, uring evaporation to :he ~' t ..

} ,,,,,

the following feedin is Fuel ca. 50 element m3 receiving pool:

/h after 10 days Waste t containers:

ca. 21 m3/h after 2 davs The re in order to feed in the relativelyare accordingly _I or 2 days 10 venting meltdown. available as the case may be

'P"'

small quantities of water for pre -

set-the question for the fuelFor this reason, a meltdown a the emeraencv element pool and measures cossibly the needed are eccident asv RAW containers since l of can be This finding is to accomolish, cribed below (Chapter 6). confirmed by the accident consid t eratiens des-plantsWithin the scope ===

,,' coolant loss accidents were inves:igatedof the licensing

=.,

German , risk study, the c ear power cours e fur:her analy:ed.tionally a assumed but improb bl of In the the /10 accident in th into :his e breakdown For this reason it of emergency cooling iscase of an addi-center

aine rs and, more particularlyseries is necessaryof problems to go fur:herhere. , Insof are concerned ,

, a number the fuel element pool andar as the decommission

. characterize the progr,ess in time waste con-of parameters should be s:c aced whi h t

of an assumed dis:urbance.

95005227 44 0

=

~ ~ -.--+mee ,-

...mo- - om -ms.

l tlUCI. EAR 1)ECOttfilSS10Nlt1C CEtiTI~.R Fuel Element St orage Pool Concentrate Container l ._

tio pressure; low temperature  !!o pressure, low temperature Syst em piessure and temperature level ca. 1/har/; ca. 50* C level 1/har/; ca. 50* C instituting counter- During the first' 11 days af ter Dur ing the first ca. 2 days h Time f ..r af ter failure of any after-me.e sn i e:s failure of any aftercooling installations, constant tem- cooling inst allations, con-

• perature level at ca. 100* C stant tempe ra t u re level with ca. 100* C In the range of 1 to 2 days, in the range of 1 to 2 days,

! Altenheat almost constant = ca. 31 FM almost constant = ca . 13 IMa e _

v ._

1(c.pi' r e I feedin rates and 11 days after breakdown of ca. 2 days after failure of p.is s i b l e improvisational ;ool cooling, the follouing cooling, the folloutng fecd-me;is.ii cs feedin quantity is required as in quantity is required as replacement for the evapora- acplacement for evapotationi tion (in the time peri <nt up (in the time period up to ~

to 11 Jays, no feedin needed) ca. 2 days, no f eed lii is i 3 reilui red) m3 50 m;a-- ca. 21

• ca. h-

• ~l le e w d t .. refer to t he IIMJ containers with the greatest heat ef ficiency.

L.D C

C Ln s v

CD

!!i ;

n gig gi g e - n .-.

, 3, .

^iU!!i  ;. ! 8.

• Hi:l

. . . . . . ~ . . . _ . . . . _ . . . . . . . _ _ . . . . . _ . . . _ _ _ -.

f - , . . . - - - -

i I .

i

.-.~.

Peipi f i c.t Juel Element StorageNtfCIEAR DI'Omt11SSIONING impi m isat lena lfeedinand ratepo:,s i b le Pool CI'NTER nat s.m) mtasntes '_

(cont len-Uitlethere ing, totalisbreakdown of cool- Concent rate Containe r

• a ulde spectrum Wi t h of measures which lead to to beprovised im i ng, total lereakdova thereby preveution ofcontrol and conditions t!ere as resieltwith similaof cool-t down ptocess the mel t - elen ent since: s t o rat,e poolt .he f uel

a. are time the differentiated onl They y in r

Available so great thattime Interv a is t f eedfit rates. in te s va l an.1 In t he measures are only (ca. required after <t tiys a. Time l 10 days). interval b.

Coolant Isatto Ing ional measuresfor improv-unit of timequantities per lead-that are by preventioncontrol and these-even be this quantityso can smsti process, ca. of meltdown 2 days.

d car t ransport. supplied I,f tank le .

Feedin LD an improvise.1 hose 1,ikeutse, rate f<ir coolant g tion is connec- only half of c3 stif fi c ien t to a Coolant t he t e.pelred I!" to 2" line (for exam-plc.

i fuel elementquant ity for t he is ply theftre hydrant) to s up -- P'"d - s t o rage N s quanttity vitti thereqtitred coolant pressure level in *hecustomary Ite t a .. t I t ply networ k. .

sup-posstht1Itles -

- Increase fuel elementof redin dancy of the pool.

no problem I.I ke and feasibleCooling is out pool. fuel element storage with-great effort at any tim e.

,,,.y *S""

illI" INN .. !, . . . .:m I r q'*.I"*T Ha .

I tlUCl. EAR DECN!!!!SSIOtlltJC CEtifER ^

l'oel Element Storage Pool _

~~Concent rat e Cons alner t:ool ley; wat er supplies Water storage quantity 10'# m 3 in Water storage qtiantity 10 l a pond on the operating terrain. In the same pon.1 as for the tiithin a time range up to ca. fuel e l eme n t. storage pool on 10 days, an evoporation cooling the on.erating terrain. Within of the fuel element can be main- a time period up to ca. 1.5 to j

tained by institution of improv- 2 lays, an evaporation coo l in g Isational measures over a very of t he bot t om set t lings (fiss-long time (up to ca. 800 days). ton products & salt s) of t he concentrate container can he ma int aine.1 liy inst i t ut ion of improvisational measures over a very long time (P.OO days).*

e.

t re The same situat ton results as When heating the water storage quantity (10 6 m) 3 with the fuel element storage from a maximum initial temperature of 35* C to pool, and there atuays result 95* C, i.e., without evaporating, as a consequence of lesser there results the possibility of afterheat longet time perio.Is keeping the fuel element to a for the cooling of ca. 230 d.

temperature level of ca. 100* C (Aftetheat assumed constant.)

• over a time period of ca. 100 d.

^'t hes e *la t a refer to the IIAU containers wi th the -greatest heat efficiency.

Lrt O  ?

tri v u

O m:!  ;;p;

, mm-r -.n n . . .

..2;

3. : '

'! i'

pj e .

~

6. ACCIDENT CON 510CP.ATICN5 /14 The operating and e:ergency cooling installa: ions for :he fuel ~

elemen; s:orage pool and the RAW containers specified in the nuclear decomissioning center are designed such tha:, wi:h :he help of one **

of the always-three cooling loops available, :he cooling without evaporating of water fro: fuel elemen: pools or RAW :on:ainers can "1.. "

~~

be maintained.

The short description (Chapter 4) shows tha: :he emergency cool-ing installations of the nuclear decommissioning cen:e: correspond to ,

the state of safety technology in the case of modern power reactors.

Since with respec: to the power reactors for con:rol of cooling acci-dents in :he nuclear dec:: issioning center always more : ice is available (cf. Chap:e 5), a mere contrast of the reliability of safety systems in nuclear power plants and in :he nuclear deco ==is- -

sioning center is only advisable when the time f actor is given value. ~

The following are a number of examples in this regard.

Emertenev Power Care Fuel Ele en: 5: orate Pool The time period available for reestablishing :he energy supply r.y a=oun:s to 10 d. There are no probabili:y data available for a ne:- L-n work shutdown over :his time. A network shutdown during the entire al.. .

time in question was not observed. Even the shutdown of five emer- .IE......T gency power diesel sets of the :ype : hat cannot establish over 10 pgs days the necessary (2 out of 6) redundancy is to be charac:eri:ed 3E as extremely improbable since the average repair ci=e for a diesel . yg a:oun:s :o 20 h according to experience. A nonavailability less than 10-8 is to be anticipa:ed. *"+

Emertenev power Case RAW Containers k5ky l

is-

[

l The period available fo: rees:ablishing :he energy supply amoun:s l to ca. 2 days. Also for :his order of magni:ude, no reliable prob- sgj_,.' ~ ~ ~

i abili:y can be given for a large area network breakdown. The non-

="

availabili:y of :he power supply coming from :he of fi:ial ne: work and :he e=ergency power diesels is, as an estica:e, 10-c. #

9 Air::af: Crash /15 syg An assumed aircraf: crash on che cooling towers would put no rmal =3;s cooline out of operatien. Since however. the cooling of fuel cic- syg ment stora;c pools and RKJ ontainers :n bc cddition lly maia aincu by a :colint pond, :his ac:iden: has no effe : at all r. the :ocl-abili:v. The fuel ele:en: p ol s::ue:u 2, the re:::cessin; buildin; and :he e c:;ency pewc: diesel buildin; wi:h all ass: cia:ed ::nne::-

in; lines sh uld be ; o:e::ed agains: air::ad: ::250 and effe::s :f de::is.

-. ---- .=

w.

=;.

c.. .

Breakdevn of !ndividual Coolan: Circuits These acciden:s have no effect a: all on the safety of :he plant since each one of :he three coolan: :ircui:s can main: sin :he required =~

cooling, i.e. , a limi:ing tempera:ure of 60' :an be kept in the fuel element storage pool or in the RA'i container even with operation of .#..

only one circui:. Only :he breakdown of all main coolan circuits '

leads to exceeding the limi:ing :e perature of 60' C. 55.:

~-sh n.. '

Ereakdown of All Coolant Circui:s and Failure of All Roosir and E=ereenev Measures only the hypothetical case that, following a simultaneous total breakdown of all coolan: circui:s, :he repair of :he broken-desm components within the :ime available also f ailed and in additi:n the emergency measures such as, for example, :he set:ing up of mobile pumps and wa:e feedin from the available water reserve through fire hoses would f ail, could lead :o heating up :he s:cred fuel elemen:s 4m=

and :he high level radioactive licuid waste. The probability of these hypothetical accidents is so low that a numerical statement is unreasonable. The accident is, for this reason, ruled out of the -

question. p

7. RADIOLOGICAL EFFECTS /16 U F###

'=r On the basis of available investiga: ions, a =el:down acciden:

can be ruled out for the fuel element receiving pool and the RAW BM#s containers of the nuclear decommissioning center. There is pracci- frh+

cally no si:uation conceivable in which it is not possible in the i ==

timeavgilable(10or2d) or 21 m /h). For this reason, :o guaran:ee it is not the needed to reasonable wa:er compute feed (50 radi-

~

ological effec s for this. The radiation exposures for the fuel element pool and the RAW containers mentioned in **'ork Repor: 290 are TC?

l TT j for this reason groundless with respect to an actual danger. They i were used a: tha: time always for purposes of cemparison in order  ;;;

j :o be able to derive requirements f or safety installations of the s~~

! nuclear decommissioning center.

The following depicts rather new findings for a nuclear ecl:-

down accident for a pressuri:ed water reactor. These findings -- .; _ .

publica:icas of c:her authors and our own computations -- should pre-vent a further misuse of the relatively in: ended numerical values jjf men:ioned in Work Repor: 290 and indicato necessary dif f eren:iations such as sher:-term and long-term dose. The computa:i:ns used as a basis always the extremely improbable even: combina:icn of cool- @@r

~"

an: less a :1 den: wi:h subsequent c::e meltd:wn and early ::n:ain-men : f ailure '.eading :: releases whi:h, in :he American risk s:udy

. ('4A3 h 2 2 ) , are cla3sif*ed in release Oa:Eg ry p%7. 2.  ?%? 2 has 6 8 95005232 4

4 y. . waeee -e.ima ..ges.a ++==mmap. pia.e u. * * = = =

in the group of nuclear mel:down acciden:s at:ording :o resul:s of the American study a 1:w probabili:y of occurrence vi:5 simul:ane-ously the grea:es: radiological effects. The corresponding acciden:

sequence (coolan: less with failure of e=ergency cooling and contain-ment failure) leads according :o resul:s of the German risk study ava'.lable so f ar in the German plan: :o a later overpressure failure '

and thereby :o essen:ially lower releases. To the ex:en: tha: re-leases in :he quan:1:y of the American category Ph*Ts 2 can appear _. __

~ ~ ~ " '

from c:her acciden: sequences, this is s:ill being inves:igated at the presen: time in connection vi:h the German risk study.

According to '4A5ii-1400, Volume VI, Figure V1 13-7 (p. 13-9): . . .

u I N i w

\

g 6 ,. \

t u . ' \ ~ ~.~ ,i s '

i -

\ -

> \ E

( l l 3u - 1 _

1 1

.t  :::: .

u wm . EM I r::.

L:::

K li 1 '

\ \

\

u -

\gw w g

\

.+

ow \ em e

i ,Nm e s

, \~ a <

e e s e o a h A sen Render Figure VI 13-7. Mortality probabili:y for an af f ected population versus dis:ance free reac:a for two hypothetical wea the rs : stabili:y ca:egory A, wind speed = 0.5 i.

=/sce; s:abili:y category F, vind speed = 2.0 m/sec. ,:...

I m:

i The American risk study reveals : hat, in the mest unfavor:ble "

s=

enne (wca ther cond t :fons F, release near ground), radiclogical of-4~'

f ects hich con load in a shcr: time to the death of the affec:cd -~"

persons are no: : be excluded up :: distances be: ween 7 and 9 miles (12 :: 15 km). In the :sse :f the e.cs: fav::cble :endi:icns (waa:h-er : lass A.: , the limi: n; radius is ab:u: 2 . 5 .i ' c s (5 k::. .

a 95005233 O

e==..- ,

-=* e-

I * .

t  %-,

s' According to the Risk Study", KFK 2433 accident consequence model in Protection,and SafetyAnnual Reper: ,

s The , May 1977, Figure 1976 7/16 (Department (p. for Radithe Ge . . .

~

meltdown radiological effec:s 116)): ation #-

accident Reactor accident Safety wasStudyconsequence determined r fo =odel (s:atus 1 .,.n.

one German locuion nuclear usin th e P'A 2 corresponding to WThis is characterized by) selecte

.. Asit-1400 release man .

,_4a,

+

category I

1 i.' - e whole body i_

s g ,,

5-e i . . _ . .. .

g..,--

' bone ~ . ; m .-

marrow ,,,, .

Idfe l4 4442 .,,..-:

u$ - I

, 4

• S.4 s- ..n::.

.t

- - - - U[

l  :- r =

l -

e..  :::n g ,. m

• M

,.*

• eJ t , .:. =

d' 4 ,

~ti;;"

') u?*

1 4

W 4 d'> Figure 7/15. 4

]pn i W. _,

~

Bone = arrow dose D and p .ari radiation death with e 24-hrobability acute T for "h e our dwell.

repo r: .

. of 12 to From this, thereresul; is published in Figur . . _ . .

'~~

I D) vi:h rain. 20 k: in the caseresul:s a limiti e 7/15 of the above annual E:'.

of average wea:ng radius for mortalityh down accidenOur of for a evncarried estimates ou:

were /

thermal ;if condici0n i, ecs:near-te rround rele as

he entire < assi:nWind 2/s) and Ve100107un censidera f avor2ble t-

/19 propa in e .

range vi:h early ncr::ali:- In order 00nStan:1:n :nditi:nshe(ves::ica r

/ analog:ur, to make the Vinda dire 0:10n during .

o the ^,e tan centerning :he and Ame rican risk b) 95005234 "W-um

- _ . _ .~ . , . - - .~. ~ ~ . .. . . _ . .

studies , f or :he abovementioned lini:ing condi:1ons, :he bone carrow dose oving to inhala:1on was computed for an in:egra:1on : ice of 30 days D3 v. In addi:1on, :he radiation exposure of the whole body ow-ing to 'e'xternal irradiation f ro: fission produ::s dezesi:ed on the ground (depositica rate 10-3 =/see for aerosols, 10** =/s ee f or tod-ine) was ec=puted for a dwell ti=e of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Dog and :he resul:ing ,

su==ation curve was prepared.

• The results are depicted in the following figure: ,,s 10'-D Dg o

/ rem /

i Dm 10 3 -

T T 1.0

[

i -

0.8 -

0. 6 m 2

10 . .

0.4 0.2

=

i ' " '

10' 1 10 100 /km/

=

95005235 m ,e. .-~ . . . . . - .

• f i

The limiting radius f or early mortality accordingly lies in the /20 area between ca. 3 and 15 km. Herevi:h, nona of the effee:s reducing -

he consequences are considered such as, for exa:ple:

Stay in house or in dwelling or in the open (shield- 67 ing of buildings) . -

. . . . ..;;. . . . .e Emergency protective neasures (f or example, cc=pu-cation of external irradiation over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />), ._

Thermal lif t and the thereby-condi:ioned reduction of the concentration near the grecad.

In addition, the whole-body dose and the bone marrow dose owing to inhalation was compu:ed for the abovementioned lici:ing condi- ,

tions for an integration time of 50 years in order to explain the discrepancies between the compu:ed radiation exposures of Wo.k Report 290 and the resul:s mentioned in :echnical 11: era:ure. The values co=puted herewith produce radiation exposures for a distance of 10 k: which a human being keeps for a period of 50 years af tet a one-time absorption of fission products through inhalation.

AB 290 New Computation

+E Release Release from core = Release factors accord- . . . .

release f ro: plan: ing to WASE-1400 ((]

a Propasation E c.

Pasquill Vogt .u.

para:e:ers Dese contents Original data Data according to US NRC Regulatory Guide g g ... .

1.109 Whole-body dose /

re /integra:1on 9.4 - 10*, S.4 - 10,-

1:e 50 years Bone =arrev dose / ,

=5s+

re:/integra:1on 1.8 100 3.3 103 eine 50 years

=7.g .

. The differences in :he doses dete '->> dn the c'd . Work Repor: /21 290 and in :he never c: puta: ions are :o be a:::ibu:ed :o the above-mentioned differences in :he release, propaga: ion parametars and dose cons an:3.

= 950052 %

" " * * ' ~ . .e . ._

g . g. .g..we we este 4 e ... gym . g.

* =,

1

-pm 74 ..

I I

The release f actors according :o '.' ASH-1400 :ake in:o considera-tion the deposition ef f ects always presen: during transport of -

nuclides fro: ex1: fro: the fuel through the con:ainment until pass-ing into the outside envirenzen:. This ef f ect was no: taken into consideration in Work Repor: 290 because, vi:hout an accura:e de- -c tailed knowledge of the nuclear deco ==issioning cen:er, to corres-ponding factors would have been derivable for :he nuclear deco = is-I sioning center.

The 2:coagation carameters of Vog: better reproduce in accord-ance vi:h the mos: recent knowledge the propagation conditions in the Federal Republic of Germany (rather considerable rough :criain).

They are a cons:1:uen: of :he " general ec=puta:1ccal bases for de:- ,

ermina:1on of radia:1on exposure owing to e:ission of radioactive -

substances with exhaus: air" which was :o be used by :he Federal i

Mir.is::y of the In:erior as a basis for a legal ordinance :o Section ~=

45 of the Radiation Protection ordinance. They are also being used at the present time in certification practice.

l l

The dese constants cre derived from the curren:ly valid NRC , . .. .

51 Regulatory Guide 1.109. Dose constants from the newes: model ec=- st.e-putations as they are used, f or exa=ple, in ASH-1400 would reduce the above dose values. (E t "_f The new cc putations show, for example, that the results pub- '"~~

lished in Work Repor: 290 -- which at that :1:e were only :o be used l f or purposes of comparison -- cannot be used f or absolute state:ents a.se+

gjgy with respect to radiological danger even on an approximative basis.

55 6 At this point, ref erence should again be cade to the conservative 0;;

lini:ing conditions used f or the new co=putations. The co:puted 5%.

doses are not sui:able for =aking state =en:s concerning the risk

'i of nuclear energy plants, rMME

/22 EEE EX?LANATION OF AB3RE7IATIONS Nuclear Ceco ==1ssioning Center  :'-["l i

NE' u<

3E Fuel ele =en:

RAW High level radioac:ive was:e ikr

== +

DWR Pressuriced wa:e: resc:or . : :_

95005237 o

A es,y.. .ee

9 i

I:

j. . c L

3!!LIOCRAFMY t

. "Untersuchungen zum Vergleich gross:noeglicher Stoerfalifolgen in einer Wiederaufarbeitungsanlage und in eine: Kernkraf:verk." ~

! (Investigations for Co:parison of Wors: Possible Acciden: .

f Consequences in a Reprocessing Plant and in a Nuclea Power .,_..

f, Plan:) I?.S- Arbei tsb erich: 290, Augus: 1976. . . . . . .

~

I

! US N3C Reactor Saf e;y Study, WASH-lLOO, App. VI.  :

Cesellschaf fuer Kernf orschung. Abteilung 5:rehlenschu:: und

[

. Sicherhei:. Jahresbericht 1976. (Socie:y for Nuclear Re-search. Department for i',adiation Protec:len and Safety.

Annual Reper: 1976). KFK 2433. q

??

l I . :!

e };;;'

l:

j ..

h i=*: ;;

~

t  ;:::::.':.

n+

e

. . ?:?'??

p--

R t

4,.

h".***.h

!9. .. X-

u: :

e 0

Y.

95005238

- e. . . . . . . . . ..

. . . -