ML20027E556
| ML20027E556 | |
| Person / Time | |
|---|---|
| Site: | 05000447 |
| Issue date: | 06/30/1981 |
| From: | Cowan R, Duncan J, Rastler D GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20027E536 | List: |
| References | |
| 81NED343, NEDO-25420, NUDOCS 8211150489 | |
| Download: ML20027E556 (136) | |
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. S'.'N RESSION POOL f,CRL'BSI.43 FACTORS FGR TCSil! LATE] 30'. LING WATER REF. TOR ACCIDE.1T C0iOITICliS 1
1 l INFORMATION REPORT c
i D. h. Rastler l
Approved:
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R. L. Conan, Manager R.' H. bucrSolz,?.anager Chtmical and Radiological BWR S; ste:S Licensing Engineering f % 1 __ - p b bY'qv%
- 4. D. h nra r., Ma na g er L. F. F idrycn, Acting Ma'ager NI frograms Plant Design and Analysis tetJC t # H I NG'5E3NG DIVIS' 7N
- Gf NCH L E LECTHIC COMPANY
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.ieners' Erectnc Company nor any ct N conenbuto.s to tns coc.rme,r A Makes any aavants J revese~ratm. eurvess or um.ved wrin respe:t to the a:cu n:, cc~p e'e ess oru s e'a es s :' ~e . -*:--r : . : 'a 'e: ~ r's .~.vu-rnen* or tw the ute of any ne*armtoon a sceses on tm cxunent map oct on**,~;e p r>a'et, c n ~ec t gnes c-0 As. tames anv respCV's Or'.tv f0* f atht!V C* T'S'P9 'e Of b nj knd wh.Ch f~ay retur it:m t"? yse ci p y o"* em e%:e C.s L?se: n tf>s Cxument I
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we. memu~ co r.cunum hED0-25420 ABSTRACT i
Ihe inco poration o, a suppression cool in the Mark 1, 11 and ill containments with a Boiling Water Reactor provides a mecianism that will De highly ef fective in absorbing fission products shculd they be relea'ed f / ora t: e reactor vessel following a ds. graded con e accident. The Mark 1.
11 and III containments are configured such tnat f or v L ascident sequences fission p oducts released from the reacta- vessal will enter the drywell and be transported to the suppression pcol where they will be absorbed or sc ubbed. This report focuses on the scrubbing efficie9cy of the suppression pool for fodines Ond particulates.
l Iodine and particulate ratention factors are presented which account for the ef fects of suppression pool rceubbing corir.g p;stulated PWR accident scenaries. The expected behavior of iocine is based on:
ob';ervaticos during reactor accidents ano dcstructive tests; nornal SWR eperating experience; sad small and large scaie scrubbing tests in water pools. A brief review of iodine chemistry is presentec to explain the chenical behavior of icdine and the expected chenical form of iodine relea'ed from LWR fuel. Jased on the datT prese:ted and the expacted BWR transport conditions, Juporcssion pool scrubbing factors greater th:n 10' for elemental icdine and particulatra , aro factors ruch greater than 10 for c sium iodido are currently justifiable fer cuccooled pools. r or satura:ed pools, scrubbing factors of at least 30 for clctantal 2
iodine, and 10 fer partic' slates and cesium iodice are currently justifiable.
Scrubbing f actors s*veral orders of magnitude larger may be demonst rated af ter core experirental testing. L'se of these attequation f actors fer the aralysis of postulated SWR accidents will plJVide consenatiValy realistic estimates of t..e consequcnces of icverely degraced accidents in perfor:ning probabilistic risk assessreats.
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i TABLF 0F CONTENTS
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{
ABSTRACT
{
LaST (;F FIwuRES i 4
f LIST OF TABLES f
1.0 Introduction
- 2. C Summa y of Results 3.0 Cor.clusions 4.0 Dominant BWR Fission Product Transport and Rcmoval Mechanisms 4.1 BWR Cescription I 1
4.2 Fission Product Transport Hechanisms During Degtaded Core Accidents f
$.0 lodine Chemistry and the Chcmical form of locine in LWR icel '
5.1 Review of lodine Chemistrv 1
5.2 Chemic41 Farm of locire in Lv'4 Fuel i 6.0 Pehavior of lodine Juring Raictor Accidents, Experimeital Destructive Tests, anJ SWR Normal Operatic.) f 4
6.1 Fission Product Releases From Jamaged Reactor Fuel 0 6.2 Experimental Reactors Tested to Oestruction j 6.2 Behavior of lodine in Operatir g PWRs il
- 7. 0 Removal of Indine and Particlas by Pools o' Water (
i 7.1 Scrubbi.10 by Subcooled Pools p J
f,-
- 7. 2 Scrubbing by Boiiing Pools I;
- 7. 3 Summary of Pool Scrubbing T.st Data l P. 0 Applicability of Pool Scrutbing Test Data to EWR Transp,rt Nchanisms I I
9.0 Application of Poot Scrurb.ng Factors in Prcbabi'istic Risk Assessments -,
10.0 REFERENCES
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y 4.1 Direct Cycle %R P.aclear Syster 4-9 N t, . .
4.7 5..apressicn Pool !cre bing Path.avs 4-10 ]
- 4. 3 (tsnaard Mar, III - 238 Quencher Arrancernent 4-11 E!cvation h gq r-4.4 Mirk III Que c9er Flan View (Typ.) 238 Plant Arrangement 5 : n 4-12 ]y b
4.b tr.R t'a r k III .indaro P11nt Contain~ nt Design 4-13 6.1 Fissicn Predact Iran; port Pathway at l!-2 6-19 [
s4
- 6. 2 se-133 and .*odine-131 Levels in Contairment 6-20 A*nosuhere at Crystal R%cr 3 y
..y n.3 1-111 Di,tribut ic i at Crystal River unit 3 6-21 h c', 3!:150 I?
6.4 H0CI 5 pts. Aimplif mt PA.!C - Trst Centigt ation 6-22 0
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7.1.1 T ansient Pressure Test f acility 7-43 [j 7.1. 2 Scale Modes of the 5 J.H.W. Reactor kented 7-44 Str** Suppressi n dysten h
- t. ,
1.1. 3 5 all scale Steam suppressica Rig ficw Dia4 ram 7-45
! 7.1.4 !.arge Scale 5tt'm suporessien 'iig Flew Diagram 1-46 I) t-e.4 7.1.5 Pemo. 31 of Element il Io.iir : Frc: a Stean/ Air 7-47 f.
S'i stare by He. ins of a Watc r Lur c hd 7.1.6 Removal of Wrticles and ladire corrooands 7-4A 6
- t re.n a stea./ Air Misturc by Heans .P a Water p;i
.i Lute R I
7 1.7 Removal of Ele.?cntal Iodice by a CO, Carrier 7-19 H
, Gas II-J 7.1.S 'le .ovs1 of Elemental !cdir.0 by 5* ?adAir 7-50 hi.stures {-
1 7.1.9 rission l'r:Juct Trapping facility 7-51 f .j EA
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1.1.'G L Cross Sectional view of tne Naniken 7-52 L Contair. ment k hi 7.1.11 MarviAen Sarpiing Points and Corresponding Saiple Lines in '.ietwell an; Lo.er Crywell 7-53 i1
'. l .12 flow Scnenatic for Si.mulation of Container 7-54 n:
Venting Lincer Sea Water b
- 1 7.1.13 5.-all-5cale Test facility !., in a Helium 7-b5 Carrier Gas ]/
J p
, 7.2.1 Principles cf Large Scale Experirnents on 7-56 g Iodine Deconta;aination J.;
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- 7. 2. ? Releise of lodine f rom Boilina wter 7-57 3
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- 7. 2. 3 Iccine tielease frcm a kater Pral a "C 7-5S t,
, into which is Injected an 10 dine-5 ixture l.2.4 Deco 713mination factors V5. Iodire .en- 7-59 l';
trat:en in a Saturated Pcci (IC07,)
fi)i Bf 7.2.5 Test Ac.:iratus to Petermire t.'emont of I.,' 7-60 kI f e ce a Soiiing kater Colonn I 1
cistr 7.7.6 Cependence et N g . ,.: ie on pH valu? 7-61 f' at D'tferent " " Pressores ps11: 27 g<
7.2.7 Dept en e of K distr of Icdine c6 pH Vilue 7-62 l-a*. Siffereat "IP preesures psia: 57 f; diste fy 7.2.8 Depencence of A of Icdine on pH Value 7-63 i<"
- PU Pressures psia: 142 at Different
.a 7.2.9 Dependence of Fraction of Unhydrolyze1 lodine 7-64 k on the pH Value of the Solution at Various Icdine Cen entrations in Solution f.
[.
8.1 The Ratio of tne Diaveter of Nacl Solution 8-9 -
Droplet to the Diancter of the haCl l'ar ticle f rc:3 =hich the Droplet has been formed at
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Various trvels of Halath e Hemidities f
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Table No%er Title Pace 3.1 Minitre surport ble and Potentially At tainable 3-2 L Si:pnrossion Pool Decontanination Facters fo" lodine and Particulates .
4.2.1 BWR Fission Preacct Transport Medas and 4-8 Transport Parameters Durir.g Degraded Core fa' Accident Scenarios
..Y 6.1 ledine-131 Measure :ent P.ccults at IMI-2 6-10 m
L . .' Chemical Forms of Radioicdire in BWR Pri:r.ary 6-13 M Cooiant M, 6.3 Chemical Forms of Radiciodine in the condensate 6-14 a.
6.4 Radiciodine species in Off-Gas at the dJAE 6-15 f1 I.9 6.5 Co.Toarison ef Iodine Activity Cercentrations l1 6-16 f. "
in EWR Reactor Water and Oftgas [
6.6 Icdine-131 Separation Factor in BWR Hot-Well 6-17 h g,;
sy
- 6. 7 Torus Airborr,e lodine Activities and Their 6-18 ;
Gas / Water Partition s.
v 7.1.1 Decentamination factors from Fission Product 7-18 i'-
Entrainment 5tt.cies 7.1. 2 Icdine recontami. nation factcis Against Sica, Flew 7-19 ,
and Proportion of Air 71.3 Indir.c Decontamir.ation factors Against Other 7-20 ;
Variables ,p 7.1.4 Decont.rnination f actors f or 0.06am Ni/Cr 7-U pt Aerosol g; I :
1.1.5 Decentamiration Facters for Elemental Iodine 7-22 in a C32 Car Ier Gas .;
7.1. 6 Re.no al of Elerental !adine by a Pool of 7-?3 r<
Water for Various Sicoown Rates, Blowdown j ,,1 Mixtures, and Icdi e (once ntrations p-l 7.1. 7 Patbyl iodide Test facility Parameters Comparison 1-24 f. !
of I'easure. tent with Design Specifications t o
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Title I'fc3 !.s l.1 7.1. 8 At, sorption of Methyl iedide in the 1/10,000 7-25 W 5cale Mocel Suppression Pool {.,
7.1. 9 Initial Conditions for tne Marviken Blowdown 7-26 '.y fests !)
7.1.10 Exoerinental Test Conditions for the Marviten 7-28 Blowdown Tests d Si 7.1.11 Spatial Distributien of Pethyl iodide in the 7-31 4.'
Containment at 'various Times Relative to the y,j Slo.edown Start, Run 4 g 7.1.12 The At.mosor tric ['istribution of I-131 in 7-3u .-
tt'e Contain-ent at various T'res Relative to {,',j the Start of the Slowdown, R.n 8
- c ,.
7.1.13 Tr.a Spatial Distribution of Elemental Iodine 1-33 i In the Containment , Run 15 j 7.1.14 The Spatial Distritation of Ele.tentc) lodine 7-34 {)'
in tne Containment, Ken 16 r.'
7.1.15 Decontamination factors of Pethyl iodide Between 7-35 l Gas and hater During Blowdown Period L .,
7.1.16 Decontaminatien Factors for Ele ental lodine 7-36 Durirg Post Blowdown Period 'I,4 m ~
7.1.17 Fission Product Re.moval by Sea Wr.ter 7-3' 7.1.18 .
Decontamination Factors for Container Venting 7-38 ,)
Under Sea Water F I's.
7.1.14 Results of Sr.all Scale Test with lodine in a 7-39 Helic: Carrier Gas (3 ti 7.3.1 Su::rnary of Pool Scrubbing Tests 7-40
8.1 Mini.mun Supportable and Fotentially Attainable 8-8 Suppression Pool Occontanination f actors for lodine ky '
and Particulates ! .;
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9.1 Applicaticn of Pool DFs for Each Postulated S-3 Accident Scenario
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In degraded core accident scenarios, the presence of water in the fission g r
product transpurt pathways provices an irtportant .nitigator to minit::ize )j e-the quantity cf air 56.ne fissinn products, lhe General flectetc Boiling * .]
w rp W=ter Reactor (M) uses the pressure suppressicq primary centain .c..t ]
syste'i to provide a water carrier to fissinn product nigration. Itius, l:p significant retention in the cool of radiciodines and other fission 'j ;
products, except noble gases, is expected and must be accounted for in c t* -
any realistic risk assessment. jb hb r <
In the ER primary consainment configuration, fission prodccts releated 'Q q in a degradeo core accident would be transported by steam from tn? -1 4
reacter vessel to the suppression pool. Tne routing would be via the [j drywel' and the horiZental vents from the drywell to the suppression pool in the case of a large break accident. During a transie.,t, the 'o.
fission product transport with steam from the reactor vessel would be "j directly to the suppression pool vid the safety /telief valve discharge h lines and quer.ch rs in the pool. During a co-bination of a transient f)]
and a nall t'r?e accioent, transport wou11 be via both pathways.
- lhe historical licensing basi,, as ese.plifisd by Regulatory Guide 1.1 [
4
.b for a loss of coolant Occident, has never recognized fission product e I i
retention in the surpression F. col. lhe Guite specifies that in a design a
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basis accident, it shall tv ns>umed that 2$$ of the rtdioiodines in th? I' cere are instantaneously transported to the prtrary containfvnt air p! '
space. This quantity is specified as teinq 9L% dier. ental icdine, 5% l[
particulates, ar.d 4% tethyl iodide. Ccrrent infornaticn indicates that iodine released frem overneated fuel is in the form of ic.creanic iodides, '
and that tnc assumptions in Elegulatory Guide 1.3 of the quantity of f, }d elemental iodine and fretnyl irdide are far in excess of any likely {j expectation. E sperimental test data indicate that cheetical formis similar ~ *f, to the inorganic iodides and particulates that would be expected to be j s
released f rom ccgradeo cores would be retained in pools of water line l'-
f the suppression peal end would not escape into the prirary containcient. ;
air space. l f'
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0 A review of reacter accidents and cirerieental tests indicates that the l'i p.
preserce cf ater plays in i.ncrtar.: rote 'n I m t ra the quantits at S iodine and particulate fission pra ucts 3vai'aole ?or leakage to te .
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enviro went. In additicn, .itural e ocesses, whict re not acendert on J cperator acticn. also li.rt t he t r.'r spe rt o f r .ss t ticsien preda: 6
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daring the various c.ccioent scerarios. Sucn pro: esses inciuce niatecut, d fallout, washout, and retention in leas paths. The overalI decentamination n
... .f factors fur postalateo Bd't scen3rios result fra? tre cc-D'.'ed ef*ect! of 47 pool scrutbing and the other re.wwal ptoresses. I',e primary focas of ,i 1
the present review is en suppiession pool scrutbirq. ';[
,F y
In this study relevant data fro.T. the literature en icaine and p Fliculate
!. . A transport throu'h ponis of w ner are preserted and cccu'.cnted. These C 1
dats cre used as a basis for est3plishing iodine ar.d part.iculate pool 2 scrubbing factors for use in prct.abilistic risk a s s e s s.te nts . For the 74 q
purpose of this report particalates are detired as all fission products ,V cther than rot,le gases cr balecens, escept tnat irorganic icdices are ! ,;M evaluated either as a sapor or as a particulate. L .5 m .:,4 t.
4 This stucy focuses (n t'te following areas: (1) de inant EWR fission I product transport and remo,al rechanisms; (2) iodine chemistry and IH ;-
locine cehavior in w3ter; (3) the espected chemical form of iodine ~
released f rom LWR fuel; (4) behavior of iodine as observed durirg actual reacter accidents tna e=perimental destrLctive tests; ($) tchavice of iodine in operatin.) EWRs; (6) indire 3nd particulate decontamination f
V f actors from experimental tests related to suppressien pool scrtbt ing; l;;
i.
and (7) postulated EWR accident scenarios and capected decontanination 3p factors.
u..
This study dernonstrates that t!'e Draisure suppression ponl in the S'AR id containment is an irracrtant t,arrier which significantly li. nits the /
w ntity of iodine and p.trtico 3tes availabic for release to the environ- .
'i neat during postulated reactor accidents. Properly accounting for this (
aart ier =111 sharply reduce tte of f-site consequences calculated in ' e d..
prcbabilistic ri% assessrents. (l t?j
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A liter 3ture search as cenactec to ioentifs and decawn? the ctem ca? [
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forn of iodine in f t.el: tre tehavior of iodine durirg reactor accidents, j 4 experimental destructive tests, and normal par occration; and the renval I;.
5
- 4. -
of iodine and particles by pools of water. The a.ailable data base was '. 7 evaluated A st. mary is given below' w sh a detailed evaluation presente d in subsequent sections of this report. ]
lJ4l t'
.J 2.1 Chemical Form of Icdine in LWR Fuel ,m
- a 41 it'e che .ical form of iodine released frora LG fuel has been reccm- . ?j mended by three independent intestigators to be Csi. During SWR 1 n fj,1?]
accident conditiens, the cor' version frcm Csl to I is nct expected ca 2
because of the prederr.inantly reducing environnent in the reactor .' 1 vessel and dry. ell. I.r
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2.2 9ehavior cf Iodine Curiro Reacter Ace' dents, Esceri: ental e' :
Destructive Je st s. and CwR Norn31 Operatior. l - ]1 i
e J
t Obsarvatio".s frci three lignt water reactor (LWR) accidents shows y that iodine released from tne plant was al ays much less than the available iodine released froc the fuel. The release of iodine was
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[j never instartaceous but occurred over a period of seserel cays, The presence of water in the plant systens during the course of '.ne f.d
[. .c4 accidents limited the arrour.: of icdine and particulates released. r.
Significant a curts of io'line were released when a wet redacing '
enviren. Tent w s s not preselt.
k i The predomirant chemical forms of iod'ne in the reactor water t' 4,
durtrig normal plant operation are ! , HIO. .snd 10 lodine carryover [ <;
3 f rc'n reacter water to steam, as r.easured by relative concent-ation ' .k .i in condensate to tt'at in reactor water, varies frem 0.2 to 2.%.
14 dine (H10) scrubbing factors in the vain turbine cordenser during
-]
nor: al plant operation are on the order of 10 . Suppressicn rool [I.[jt.
scrubbing factors for H10 have been rieasured to be greater than ten .
thcusand. 3
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[ 2$Ydd[tb u NEI'N'* 53N21$5E.b $dQ1hMbM.I.b[d () ;
C 3-2 W O L,
"'.4 .p lr 23 Re~.eral or Ni na aad e'erticles t'v Poals of bat.>r f'
- .A l.5 cut one hund ?d tect.iical capers were revie.ec f rc.m which elever. k JJ applicable ey crimontal tests were ider tified which in nstigited I v4, t r.c cif ect , of fissi:n crecoct retentir
- ty p-essure suppressim $~j: :-
g ro.)l s . ihe caj^rity of the tests useo clevental iodine (i ).y Triere als-) were scoe cata en the acrubbing of CH 1. ril, HIO, and 3 [.g1
- call insoluble p.srticles. There were no reported data en the fj transport of Csl tnrcugh a pool of . cater, however the bd.autor of h. n ;
Cs! as a sacor or as a particle may be inferred from test ve dts y- .
using similar chemical forms. Tarle 7.3.1 presents a summary at d w +g 1 the empc Sent al test conditions and . measured decontaminatic ,"A n
foctcrsi for each test. 3 '
2.4 Evaluatio,cf Dat.i Race ,2 j'.<
Review of the e=perimental data base inoicates that it can be used ' ,..;)-.
J to conservatively bound cWR transport and retention conoitions 4g existing as a conse:;aence of degraded core accidents. Whiie several [.jj tests were rease, ably representative of :ne transport phencmena h$f exper.ted darirg a postulated accident, the experimental conditions .!)
(e.g. , shallow t sol, small particles, volatile !.) led to lower i paol cecontanination factors (Dfs) than vould be expected in SWR
.f certain. tent syst*m conditions. Comir nt EWP. fission nreduct transport k mechanisms during accident scenarios were identified and the esperi- , ' t'] y, rental Jata were assessed for applicability to expected cenditinns.
Minimum pool scrubbing 0Fs 5.hich the current data base can sunro*t.
l )a 3
and the pctentially attainable CFs which could be supported by I : ;n -
further testing, ar e presented fer each dominant transport seqt.? ace.
[ -]
The application of pcol decontamitation factors in risk assessments lbW is presented in Sectica 9.
a l-Fo.)tnote 1: The de. onta.T.inaticn fictor or " scrubbing f actor" is defined }o.cc4 as the ratio of fission product mass trtasported into the u001 to the W
]
mass which is releised f rcat the p)ol surface. 6
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2-2
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3.0 CONCLUSION
S Suppresrien pool decontaminstion factors apprcpriate for use in EkR risk assessments are oresenteo in Table 3.1. Based en the data presented cnd the expected BWR transport cor.Jitions, suppression pool Cecontamination factors of at least 102 fer elemental iodine and particuletes, and 103 for cesium iodidi are justi'4able for subcaoied poors. For saturattd pools, decontamination factors of at least 30 for elemental icdine and 103 for particulates and cesium iodide are terrently justifiable. It is felt that these minimum values can be incre.. sed savera' orders of magnitude by ,
further testing for cen11tions more representative of the suporession pool during post accident ccrditfor.s.
Natural processes such as the agglc . eration of solidt, plateout, '
depresition, washout, etc., also play an important role in limiting the gaantity of fissicn products available for leakage to the enviroament. The over.11 attenuation factor applicable to BWR oegraded core accicent sce.arios includes both the effects of pool scrub'ing and of such nctural removat processos that will occur in the various volumes o.' the SWR process system and its multiple contai ment system
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SUI e>RE5510N P001 0FCONTA'4f NATICN FAC10i:5 r0R L9 IODINE AND PARTICULATES .
s N
1ransport Pathway Hinimum Supportable Potentially Atta' table and Associated F. vent (s) DFs DFs (3) y Subcooled Pool (1) Saturated Pool (2)
Td Reactor pressure vesse!
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105 Csi, l ,HI Ifir ; articulates (4) 105-10s c.s!, ! , HI Li to pool via "fety relief 10" particulates 30 1 2 102-108 particulates 9 valve and quencher 102 g, ic2 103 1 2 A (Transients) g}
~
~ L' Reactor pressura vessel to 103 Csl, I , til 102 particuletes (4) 104-10" Csl, l , til h pool via vents (Transients 102 particulates 30 1 2 103-108 particulates b following RPV depressurization, 102 g, 102. j ,)2 g p jy) or LOCA post bicwdown period)
Aeresol Transport to Pool 102 particulates 102 particulates (4) 103-108 particulatEs z N Via Vents (Cure-Concrete Vaporization Re1 ase) 102 1 2 30 1 2 102 103 1 2 k4 M
= :1 3 h V.J n.
N$
N (1) During these conditions, complete condensation is expected when the pool is subcooled.
(2) A subcooled pool is at a temperature below the saturation temperature corresponding to the pressure in the contaire.ent, while in i saturated pool steady state boiling " steaming" .is. occurring.
(3) Potentially .ttainable by further testing (satuiated-subcooled pools).
(4) Incicdes Csl 4
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l 4.0 MAJOR BWR FISSION PRC')l'CT TRANSPORT AND RJMOVAl. MECHANISM In this section, a description of the maj;r SWR design features is presented. ' Postulated BWR acc.oent scenarios, fission product transport and removal mechanisms, and expected transport pararceters are presented in ords to assess the applicability of the pool scrubbing experiments p Nsented in .rection 7.
4.1 BWR Description General Ciectric Boiling Water R: actors use a multiple cor.tainmen*
system featuring a pressure suppression primacy containteent of the Mark I, II, or III neometry. A schematic of the current design GE-BWR,6 Mark III system is shown in Figure 4.1. The direct cycle begins with teedwater entering the reactor vest?l. Recirculation pumps and jet pucps circulate reactor water through the core region where boiling occurs. A saturated stes.m/ water mixture at PDout 1050 psia and 550*F passes through a series of steam separators and dryt.rs . The dried steam flows cut of the reactor vessel via the main steam lines to the turbine. Steam from the turbine is exhausted to the main condenser. The resulting condensate is pumped threugh a full flow condensate treatment and feedwater heater sy;te a where it is purified and preh sted prior to re-entering the reactor vessel.
In tha Mark III design, the reactor pressure vessel is enclosed by a drywell structure. A pressure suppressien pool containing about a million gallons of water together with the drvwel. forms an additional barrier between the reactor vessel and ti.e primary containment air space. Both the drywell and the pool significantly retain fission products released during postelated accidents. The primary containment and the secondary contale. ment shield building fully enclose the drywell, the suppression pool, and the reactor system.
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NED0-25420 in the BWR cc.1tainnent systea. tission products released in a degraded core iccident would be transported by steam from the reactor vessel to the suppression cool. 11 the case of a loss-of-coolant accident, the routing would be via the drywell and the horizontal vents from the drywel . to the wpression pool.
The five-fc,ot thick dr;well wall is penetrated by sutecrged horizontal vents (27.5' in diameter) which have center linas located at three *.evels belod the surface of the pool, with a minim"m depth of seven feet. There are 40 circular v2nts on each of the three levels.
During a transient or small break event, fission product transport with steam from the reactor vessel would be directly to the s'Ip-pressio t poo' via sece of the 18 to 20 safety / relief valve dischrge lines. The steam is released into the suppression pool via quenchers as illustrated in Figures 4.1, A.2, 4.3, and 4.4 Each quencher ensures rapid condensation of steam in a subcooled suppression pool by discharging the steam through more than a thodtand small holes wnich are approximately I cm in diameter.
4.2 Fission Product Transport Mechanisms During Degraded Core Accidents Degraded core accident scenarios address events where the reactor core may be damaged significantly relecsicq substantias quantities of fission products from the core. Degraded cora accidents are extremely inprobable. If an event is initiated by a transient or a primary systen break, extensive or total failure of sa'ety and emergency core cooling syste.r must also occur. This section presents major BWR accident scenarios censidered in probabilistic risk assessments and describes the fission product transport mechanisms expected for each scenario. These scenarios are of two trajor types. Class 1 and 3 considers those where core damage initiates loss of prieary containment integrity by overpressure 4-2' l
A ew3msomwnm;yw;xya:r,wg33:, iT,,myrygggm gyggng m.;gorn wn hE00-25400 and crackinn. Class 2 and 4 considers those where loss of prinary containn.ent integrity by cracking initiates core dr.3ge. In Class 1 and .1 scenarios, the suppression pool is subcooled while in Class 2 and 4 the pool is in a saturated condition. At the ti.te of core damage in Class 2 1nd 4 scanarios, the pool is not expected to t'e turbulently boiling, but rathe~r in a " steaming" condition.
Wichin each of these tw) major accident types, BWR accidents can t'e further grouped into two categories: (a) discharge through the horizontal vents, and (b) discharge through the S RV disch1rge lines. Figure 4.2 shows the expected fission product pathway for each of these categories.
i During a postulated design basis LOCA a double-ended bream of a
} recirculation line is assumed, resulting in the rapid depressuri-zation and blo,d'wn of the reactor vessel to the drywell and then to the sup,'ression pool by way of the horizontal vents, lhe
. r'sulting high drywell pressure signal will initiate the upper pool dumptothesuppressio7[001.1 For degraded core conditions to occur, the emergency core cooling system (ECCS) must be assumed to fail resulting in reactor fuel heatup and subsequent damage ano
. melting. Prior to clad melting, hydrogen cay be produced from the (Zr-steam) metal witer reaction. If degraded ECCS performance is assumed to continue, melting of the core may continue to propagate, and th? melted portion raay slump into the botto.n head and result in additional steam goneration from the water remaining in the lower portion of the reactor vessel. As the cladding melts, fission p;oducts released from the fuel cay be transported (with the steaat and hydrogen) to the drpell by way of the break in the recir:ulation piping during the post-blowdcwn period. Fission product transport in a predominantly steam environment can occur from the drywell to the primary containment air space through the suppression pool.
The fission products would have to pass th ough approximately Footnote 8: The upper pool d mp is initiated of any one of three signals:
hig* drywell pressure, low reactor water level, or low suppression pool water level.
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l-4 thirteen and ene-half feet of watert bef ore reaching the suppression {
pool surface. Flow or ly through the first lesel vents is expected f as steem flow at this tir.e late in tte scenario is only sbaut five q pounds per second per vent. A r ronimately ltft to 20% of the vent cross-sectional area on the drywell side is expected to unco *,er, initially pecducing bubbles cf various sizes. Near'.y all of the volatile fission products in the fuel are expected to de released (in a steam environment) 1caing this stage of the accident sequence.
For postulated transients where ECCS makeup water is assueed to be not available, fission products released from the fuel to the reactor vessel may '>e transported directly to the suppression pool e in a predominantly steam environment via the safety-relief salve discharge line and Quencher system. In these events a low reactor t water level or a high drywell pressure signal will initiate the f upper pool dump to the suppression pool. Ir these ever.;s, the h fission products must pass throLgh apprcxi.tately ninateen feet of [
water before reaching the primary containe.ent air space. Nearly j all of the volatile fission products in the fuel are espect ri to be l released (in a saturated steam or superheated steam environment) :
during this star'e of the accident sequence. In the casJ of small f or intermediate size creak accioents, both transport pathways occur.
Following either a LOCA or a transient event, the noiten core ar.d !
other naterial .may penetrate the thimble, on the reactor vessel ;
bottom head nd drcp to the floor of the drywell in the restricted }
area directly under the vessel. Under LCCA sequerces this is not a ;
v olent event since the reactor vessel is alreacy depressurized. .
Howeser, for the transient cases where no large LOCA breat exists y and wt.en the operator has failed tr depressurize the vessel in h
7 Footnote 2: In sequences where ECCS operates as designed, the expected !
depth rf submergence is about 8\ feet c 4-4 E
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order to make the low crosrure ECOs available , there may te additional 2
tron 5 port f rce the vessel to tbc drywell and suppression pool via i the horizon *.al vents. Vessel de;tessurization is net expected to q cccur instantaneously and violen'.ly, but insteaa tne depressarl2ation
[
should cccur rather slowly as the core material slowly penetrates through the control rod and instrument thimbles and perhaps eventually 3 through the bottom head itself. 4 Following the vessel penetration sequence, a moiter. core-concrete interaction may occur on the portion of the drywell floor of rectly g under the vessel releasir; aerosols and nonconcensable gases such ?
as CO 1 2 . Figure 4.5 illustrates where this interaction may occur, p Note that thJ very substantil reinforced concrete structure arcu.;d '
this central portion of the r) ell floor area prevents any direst i
~
access of this molten core-concrete mixture to the region of the kg suppression pcol in the BWR Mark III cent.Inmeat system. In addition, fv features assuring the st;uctura! integrity of the suppression pool r includ2 the five-foot thick drywell wall, the ten to fifteen foot "
thick floor under the pool ard dryvell, atd the eight foot thick -
wall arour 1 the pool outer radius. Ground elmation is above the h pool wate- level, as indicated in Figure 4.5. b P
tl The noncendensable gases penerated will slowly pressurize the
[
drywell and eventually vent through tne horizontal vents (top row '
only) to the suppression pool. Gnly a small fraction (about 1*.) of j a
the vert cross-sectional area on the drywell side is e pected to be il uncovered producing relatively stall bubbles. During $his stage, > d. l fission products (vaporization release component) attached to aerosol particles may be carried O'y a CO2 -stean nixture through the vents and depending upon the accident scenario released under 8h to 1 13\ feet of wa:er. 3 4
l.
Footnote 8: The Emergency Procedura Guidelines instruct the operator to Si depressurize the vessel if the high pressure energency core cooling j systems are not maintaining an adequate vessel water level. E fi LU 4-5 Y
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!E00-25420 JL cJ ,
A l For those fission products v*iich are not absorbed by the water .
-l prior to reaching the sup,3resson pool surf ace, transport may occur in the primary containment building. In the event that the co-'tain- y i e 1
- a. cot pressure continue 3 to increase due to loss of decay heat t rernoval or non-concensib'e gas ceaeration. General Electric analyses
)
indicate that catastrophic failure of the containment is not expected.
]
The most probable cause of loss of primary containment integritv li for the fiark Ili Standard Containment design would be cracking near kc the top of the containrent structure. Buckling and cracking would g be expected in the area where the transiticn from tne vertical wall j to the c'sme is located (refer to Figure 4.5). Therefore, fission ]
Il products (gases or particulales) released from the pool surface must. travel for approximately 130 feet f rem the pool surf ace thrcugh a tortuous path of stairway gratings and cornpartnents with significant ]9 plateout surfaces before reaching the location of the cracks in the f primary containeent steel. After leakage through the cracks, h) 3daitional travel through the snield builaing will cccur, with r( -(
exhaust througn the standby gas treatment system and/or directly outside through cracks or other leakage points. Significant areas '
for plateout and deposition are available ia the shield building g and the other volumes withir, the secondary containr:ent zor.c. '
A summary of the dominant BWR fissien product transport r. odes and sigr'ificant transport parametcrs is presented in lable 4.2.1. The g transport and subsequent removal of fission products by the suppression [d pool is an expected and highly probable occurrence during m.ost ]
postulated degraded BWR accident sequences. The pool is a large passive !,
,c D
system and its capscity to retain fission prcducts should not be ;
undarestinated or neglected in any risk assessment.
Ih In addition to removal by pool scrubbing, natural processes in all J',l BWR cerrpartments within the reactor system, the drywell, and the h' primary and secondary containtrent air spaces will limit the quantity l of fission prcducts available for release to the environment. 1 These mechanisms include: agglomeration in the reactor vessel, j i.
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plateout/derositier., aerocol aggloneration, washout, and retention il in leak patt.s. The oscrall fission pi aduct atteruation facto- for ji a BWR will include tne e:fects from both pool scrubbing and all It*
such natural re: caval and etention processes. Li y
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. **'f.4.~. e .
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g r. n.y ,, h LOCAT80N OF CORE CONCHETE 9 *0 *m". INTE R ACTION Figure 4.5. Ek'R Mark III Standard Pirnt 4-13 /4-14 1
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.h$ 4 ,,._9_- o NEDO-25420
- 5. 0 lujine Chemistry and the Chemical forrs of Iodine in_ tWR Fuel 5.1 Review of lodine Chemistry The behavior of iodine as observed in reactor accidents, norm .1 BWR ooeration, and during experinental tests invoivir.g suppression pool scruubing, can be better t'nderstood by a brief overview of the chem;stry of iodine. A complete review of icaine cheliistry is beyond the scope of this report. howcVer, a summary of itdine c*ienical forcs and ' heir expected reactions, solubility and vola *ility, is ussful in assessing appropriate decontamination factors to oe used in degraded core accident analyses. In such i snalyses the chemical forms of iodine of interest include elemental -
iodine (i ), rganic iodides (e.g. CH I), cesium iodide (CsI),
2 3 hydrogen iodide (HI) anJ hypoiodcus acid (HIO).
Elemental acine (I2) is s latile at room temperatures, dissolves slightly in water, and may be rapidly hydrolyzed in aqJeous solutions by the following reactions:
. 'aq) + H O 2
+
H+ + I' + HIO(ac) or 1 (aq) + OH' + H+ + I" + 10' 2
3 12(aq) + 3H;0 < 5 I'~ + 103 - + 6 H' The degree of hydrolysis depends on 21 c ncentration, pH and temperature.
Dissolved iodide (I') might be oxidizec' to molecular iodine in acidic solutions by oxyge.. ;n the contain ent atmosphere, by oxygen dissolved in the w; iter or by radiation effects ia the wat2r.
Radiation induced chemical reactions in aqueot.s iodine solutions c .
l have been re,norted (5.1).1 At concentrations <10M (3 ppm), I icns can ba nearly quantitatively oxidized to 10'3with hypoiodous acid, HIO, at a possible interr:ediate, in solutions with r i below 9 ur.Jer 6
the influence of high intensity (5x10 R/hr) gamma radiat on. (This i
l t
Tootnote 2: The notation (5.1) refers to Reference 5.1 5-1
n w- - -
.rw . . . . ,- ;. w .>. - , ,,. x - . .
- r. x , c. : . . ;, , s <. ... 2 C O-25420 mecnanism is believed to play a dom:.' ant role in producing volatile
!10 during normal BWR operation as will be shown in Section 6.3.)
Under degraded core conditions if all the iodine is conservatively assumed to be released from the fuel and reactor system, the maximum iodine t.oncentratioa in a BWR suppression poo! would be approximately
-5 3 ppm or 2x10 H (5.2). The pH of the suppression pool is exoected to be s10.5 under such an assumption because of the lar,e quantities of Cs (eleaental Cs, Cs oxides, or CsI) also released from the fuel (5.2, 5.3). Also, the noel would not be subject to a hign intensity radiation field. Thus, air oxidttion or radiation induced oxidation of I' to 1 is 2 n t very likely to occu- in a suppression pool.
The formation of organic cr.c; nds of iodine has bean studied by several investigators (5.4). Elemerital iodine (12) was the pre-dominant chemical form of iodiae investigateo in these studies.
The co.iversion rate of 19to organic iodide would ba expected to be much larger than the rate for I~ to organic iodidt because of differences in salatilitier.. This occurs because high airborne concentrations of 12 promote the conversion to organic iodica *,hile low conceitrations limit the conversion. Organic iodide GH 1) 3 was ,
reported to rear'ily hydroly;:e in basic solutions of water at high !
temperatures (T=100"C) (5.5, 5.6). l 1
i As cesium iodice is non volatile and highly soluble in water, it i should be efiectively retained in a suppress.on pool. L'nder degraded ,
core conditions, cesium iodide vapo. 'to' aid be expected to condense ;
into a particulate form at temperatures less than 500'C. This .
condensation h:ay occur on the inside surfaces of the reactor pressure vessel or onto other particulate surfaces' airborne in the reacter vessel. The transport 3nd retention of a particulate by a suppression ll pool then heromes impor wnt when considering the removal processes of cesium iodide.
Hydrogen iodide (HI) is volatile and is rapidly hydrolyzed in water with a partition coefficient near infiaity. Thus, it should be effectively retained in a suppression pool.
5-2 1
. , w ; .. . . . . u, o. .. ~v. u nw - *v NED)-25420 1
l Hypoiodous acid (HIO) has never been chemic-Ily icentified. It is a volatile and highly soluble cc.cound which is rapidly hydrolyzad in aqueous =olutions. Thus, it should be effective'v retai ed in a supp = sion pool.
5.2 Chemical Fores of Iodine in LVR Fuel The chemical form and oehavior of fission p.odccts in LWR fuel has been asses,ed by Adamsen (5.7). The fission products Cs, Rb, I, Br, Te, and Se are classed as " volatile-rsactive" and their behavior resembles that ?f the gaseous fission products Xe ar.d Kr at +eepera-tures above 1400"C. The " volatile-reactive" fission products tend -
to migrate dcwn tenperature graefents in the fuel through cracks, etc. , ar.d co.idense or react in tne core fuel regions (<1000'C). .
Since Cs and I'are both aggressive chemicals, they may combine to form CsI. The fission yield of Cs is 10 tines gecater than that of iodine, therefore, the probability that all the iodine could combine with Cs is relatively high.
The boilin'i point of iodir.e (12 ) is 184*L, while the boiling point of cesium iodide (Cs!) is 1230*C. Since the average clad temperature of thR fuel is 300 C,1,2if present iriside the gap, should behave like a fission. gas but Cs! should not.
In gap purge experiments at ORNL (5.8) fission products were purged from irradiated commercial fuel rod sagtents at temperatures up to 1200 C. The conclusit,as from the experiments were that the dominant chemical form of iodine in the fuel is a metal iodide, probably CsI. Crystals of Csl have been icentified on the internal cladding surface of an irradiated commerical fuel rod (5.9) by X-rr, fluore-scence. Other papers also support the conclusion tnat the chemical form of iodine in LWR fuel rods is resium iodide (5.10, 5.11, 5.12, 5.13).
5-3
- HED0-25420 An iodine " spiking" release phenomenon from c' adding defects has been obrerved (5.14) when reactor power or pressure is stddenly changed. The spiking of fission noble gases is distir.-tly dif-ferent frri that of iodine, indicating that the chemical form of iodine releas'd from the fuel is not gasetis elesental iodine.
Organic iodides exist in extremely small amounts under accident
. conditions. In the TMI-2 accident they a.nounted to about 0.003-0.01%
of the core fission product inventory (5.15, 5.16, 5.17). It is the dominant form of iodine in the gas phase not because there is a large amount present, but because the species that are present in large amounts (the iodides) distribute overwhelmingly into the water phase, leaving only trace quantities in the gas nhase.
Radiation ef fect- in water .3ay contribute to the formation of organic lodides but the reaction rate is relatively slow. Peasure-ments of iodine in operating BWRs show that very little if any organic co.mpounds of iodine are released from LWR fuel (6.9).
Based on the available data in the literature, experience from reactor accidents (Section o.0), operating 8WRs, and the expected moist-reducing environment in the primary system and drywell, it is concluded that the chemical form of iodine released from LWR fuel is cesium iodide aad this species shculd be used in assessing suppression pool indine scrubbing factors during BWR degrad 1 core accident scenarios, i
l l
5-4
2 . . . .. .. _ . .. .. . . _ . . , ... ,.
. .s.- .. .
NE00-25420 6.0 BehavioLof Iodine During Reactor Accidents. Experimenta.
Destructive-Tests, and BWR Nornal Operation In this section observations are presented on the behavior and transport of iodine and particulates daring reactor accidents and small experimental reactor destructive tests. The behavior of iodine in operating EWRs is also presented to illustrate two significant removal processes in a BWR: removal of iodine by condensation processes in the condenser; and scruboing of iodir.e by the suppres-sion pool.
~
6.1 Fission Preduct Releases From Damaced 'esctor Fuel, 6.1.1 The Accident at TH!-2 THI-2 is a pressurized water teactor (PWR) and its containe.ent system coes not have the pressure su,npression pool which is an integral part of the BWR design. The loss-of-coolant accident at THI-2 resulted in the release of large amounts of fission products to the reactor coolant and the reactor t. Iding containment. A stuck open pressurizer relief valve resulted in tiie transport of primary system water directly to the contain:::ent where it subse-quently flashed (see Figare 6.1). It is postulated that extensive core damage occurred when the water level dropped balow the top of the active fuel allcwing fissan products released from the fuel to be transported in the saturated water-steam mixture directly to t.
containment. Although the magnitude of the radioactive snaterial released during this accident was significantly higher than any .
previous accident in the ccreercial nLclear pe,er inuustry, only "
modest amounts of iodir,e escaped f rom the primary containment.
The President's Com'iission on THI-2 (6.1) reported that the total release of iodine 131 (I-131) to the environment fror. March 28
[
through April 27 was s17 curies. This release was insignificant j compared to the 7.5 milliun curies of I-131 which was discharged l 5
i 6-1
NEDO-25420 from the RPV. As noted in Table 6.1 only 2000 Ci of I-131 became airboine in the reactor builoing. khile measurements were not available imediately af ter the acesdent, measurements made several months later showed tnat the only iodine airborne was organic iodine which is relatively nonreactive. It is evident that the activity remaining airborne was a very small percentage of that rcleased demenstrating that iodine released into a hunid environ-ment ic readily removed by natural phenomena. Also the majority of the iodine-131 activity remained in the water.
In view of these findings, the President's Commission recommended that "a study should be made of the chemical behavior end extensive retention of radiolodine in water, which resulted in very low releases cf radioiodine to the atmosphere."
6.1.2 The SL-1 Accident (5.2)
. The SL-1 was a small natural circulation reactor (3 MWt). The fuel elements were constructed of highly enriched uranium-aluminum aleoy [
sandwiched by aluminur alloy cladding. Compared to a present day commercial pceer reac*or few engineered safety systems existed in this prototype portable nilitary reactor. A sudden canual removal of a central control rod by maintenance personnel caused a rapid [
I reactivity insertion which lead to a power excursion and extensive core melting. A very large fraction of the core was celted and most of the water was ejected frcm the reactor vessel, yet less than 0.5% of the I-131 ccre inventory was released to the outside atmosphere. k'hile SL-1 was contained in a " Butler" type structure, which has ninor containment capabilities compared to containment structures associatt.d with commercial power reactors, only 10 Ci of I-131 were reieased to the atnosphere during the first 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> ana 70 Ci over the next 30 days. The expelled water was h hly contami-nated with fission products, including radioiodine.
. 4 6-2 j
k.I" t4EDO-25420 I h
0 In fd .
6.1.3 Crystal River-3 Event (f>.3) p 4
n Crysta! civer Unit 3, a PWR, expe:ienced an event on February 26, H 1930, in nich 40,0C0 gallons of primary coolant water was discharged onto the ficor ;f the closed containment building. An electrical g equipment failure resulted in a stuck open power operated relief p valve. With the reactor scrammed and the power operated relief ;
valve open, the reactor vessel pressure dropped as the reactor coolant discharged into the reactor coolant drain tank and then ,
flashed into the reactor containtrent building af ter the drain tank d
- \
rupture disk actuated.
6 o
.h The reactor water contained both I-131 and Xe-133. The Xe-133 was h;.
immediately released into the containa nt atmosphere, but the I-131 I/j was observed to be retained to a large extent in the water. The I-131 in the conta n. Tent at.osphere peaked in about four days as i
i,(
shown in Fioure 6.2.
N, following the event, the containment air and the water in the Y contair. ment sump was analyzed for I-131. Approximately 1.8 Ci of I-131 was airborne on March 2,1980 while approximately 85 Ci of I-131 was in the suTp water (6.4) as shown in FigJre 6.3. Thus the t .)
decontamination factor was approximately 55 days af tec the s' 7
incident. A higher DF may have existed on the day of the incider.t K but this could not be evaluated due to lack of data. ,
'1 The primary phenomer.on in this event and in the THI-2 accident fl}j w
appears to be the flashing of primary reactor water with tt'e y sub:equent release of fission prcducts to the containment. In spite of this flashing phenomenon, cecontamination factors en the f.f.9 G.L ,
order of 50 were observed for iodine. ,'
G i~
Ol La 1 '!s II m
l n
6-3 4 '
siili k l
Y' y
NEDO-25420 j A
k n
6.1.4 Windscale-1 ,
i 1
The accident at Wincscale-1 (a once-through, air cooled, uranium- )
graphite reactor) occurred when the uranium metal fuel overheated, p/
1 caught fire, and released fission products to the air stream C owing lg up a tall stack (6.5). The stack filter was estimated to have gj rercoved 2-5x10' Ci of particulate irdine, however, 7.0x104 Ci of [q gaseous iodine was released from the stack to the environment. The
}
released iodine represent?d 12% of the available iodine inventory j r in the core. Water was ng present at Windscale since it was an i air cooled reactor, thus accounting for the large fraction of iodine released. ?
a 6.2 Experimental Reactors Tested to Destruction i y
r Three experittental reactors have been deliberately tested to destruc- i tien te verify that-large reactivity excursions were self-linitinn '
.I and would automatically terminate the nuclear reaction. The tests : -
were designed to violently disassemble the core and melt or vapori2e ]
part of the reactor fuel. Dispersion of the radioactive material
]
was monitored to provide information to assess the dispersal areas, f Three tests of this nature were the 809AX-1 test (6.6), the SPERT-1 test (6.7), and the SNAPTRAN tests (6.8).
The BORAX-1 experimental reactor was loca.ed in a Icw pressure
.)
water tank partially sunk into the ground. There was no building q over the reactor. Motion pictures taken during the tests showed .
that the water tank holding the experiment burst anJ most cf its 3 contents were ejected into the air. All the fuel fragnents could $
be accounted fer witnin 350 feet of the reactor. A wind speed of [j 8 inph was present during the test yet the phenomenological mechanisms
[.]
limiting dispersal operated to attenuate the release of radioactive fj fission products. v n
al h
R 6-4 c
t,
&..s NED0-25420 g n
a The SPERT-1 destructive experiment was also cenducted in an open :fl 1,
tank facility. It was covered in a lig5t structure not intended for containrent purposes. h the test approximately 35% of the bo core was melted with nearly all the fuel elements in the core W experiencing melting to some degree. Metallic debris approximately j 100 pm in size was recovered in the water. Only noble gases escaped 5 to the at:nosphere from the reactor tank. The radioactivity released f to the atmosphere was estimated to be 2.4x105 Ci; which represented a less than 1% of the fission product inventory in the core. Icdine h
- o.
was detected only in the reactor water. ;
cs The SNAPTRAN series of destructive tests weic conaccted to evaluate the risk of launching the SNAP-10A reactor (a U-Zr hydride, NaK a cooled reactor, designed for space applications) over water or p
j land. In the SNAPTRAN-2 test, no witter was present and 70% of the q iodine was released to the atmosphere. In .ontrast, when the d
.s reactor was placed in an open tank of water (the SNAPTRAN-3 test), 1 there was no detectable iodine released to the atmosphere. Nearly 'q all remained in the water, in the fuel or plated out.
}$
6.2.1 Su.. mary of Observations on lodire Behavior In the reactor acci e nts and destructive tests involving an initial I water barrier very little iodine released from the fuel was released
to the atmosphere because of retention in the water which was present. Iodine tended to remain soluble in water rather than ! .f<
escape as a volatile gas. This behavior is consistent with that l .
observed frcm suppression pool scrubbing experiments where iodine 3 (in steam) is passed through a pool of water. Results from these [:?
tests will be presented in Sections 7.0 and 8.0.
Oh}
.g 6.3 Behavior of Iodire in Operating BWRs '
l$
f(d,.
The behavior of iodine in operating BWRs is presented in this section and includes: the chemical form of iodine in BWR systems; W
ii!
6-5 -
O;.
. . . 'udi
' ~
i.
1.'i ?
NED0-25420 j i
renoval of iodit.e by condensation processes in the condenser; and j removal of iodine by suppression pool sceubbing during Hig1 Pressure O Coolant Injection Surveillance tests. Decontamination factors for .j each of these processes are presented to illustrate how the separa- [
tion of iodirte (HIO) from steat f avors the water phase rather than the gaseons phase. This behavice would also be expected to occur =LG' during BWR accident cormitions when steam carrying iocine (HI or 5%
1 Cs!) would be rapidly condensed and scr0ced as it passes throug'i e, a
the suppression pool.
a 6.3.1 Chemical Behavior of Radioicdine in SWR Systems (6.9, 6.10) ,;S d
.]
An examination of the chemical forms of iodine at operating BWRs //
>:2 during normal plant operation and during shutdown conditions indicates ;p that I ,10 , and 3
perhaps HIO are the predor.inant species in reactor af water. Only traces of 1 ,2and rganic iodine (CH 1)3were detected and are believed to be from contamir.etion in the analytical procedure, d[
ry Results of analyses of reactor water for iodine are shown in Table ,y 5'. 2. During reactor shutdown with no steam distillation, most of b the iodine was found in the iodate form in the reactor water. This i
. L-is probably a result of radiation induced oxidation of I. to 10, or d 10},sinceitiswellknownthatl'canbeoxidizedby10- 3by the N]
influence of radiation in aqueous solutiens (5.1). T y
E.
The amount cf radioiodine carried over by the steam has also ceen ' i, investigated. The term " carryover" is defined as the ratio of iodine concentration in the condensed steam to the concentration of ir.;
i the same isotopic species in the reactor water. The moisture ,9 content in the steam is very strall as a result of the action of the .2 moisture separators and steam dryers in the BWR design. Based on ,
the distribution of nonvolatile ionic species, such as Na 24, g between the condensate and reactor water, the iodine carryover by a mechanical precesses, e.g., entrainment, is estimated to be < 0.05%. @
LcJ The observed iodine carryuver, hcwever, ranges between 0.2 to 2.5% rq (6.11), therefore, mechanical processes alore cannot account for 7 p
all the iodine carried by the steam. This suggests that some #
r J volatile iodine species must be involved in the carryover process. J 6-6 2' kh
g n
).Q NEDJ-25420 id
? '
i.
0:
Volatile H10 is believed to be the major icdine species distilled $] 'bY from tne primary coolant during normal operation and is belie.ed to d.g be produced by the radiation induced oxidation of I" found in the w?
p primary coolant (6.9). Essentially all the iodine species which N r . s.1 are carried by the steam from the reactor water, through the turbine, .g and to the condenser are condensed with the steam in the hotwell. Q sq Only trace noncondensable iodine species are fou:.d in the gas phase dj which consists mainly of radiolytic gases. Since HIO is expected .h]
to be the major volatile species in steam carryover, it should be expected in the steam condensate. This, indeed, has been shown by j examining the chemical forms of iodine in the condensate of operating M NRs (refer to lable 6.3). Very little 1,10},2 or organic iodide j were found in the condensate samples, and essentially all the lodine was I' or HIO. Unfortunately, the chemical identity cannot r,d1 uL be distinguished between the I" and HIO forms. @
..v h
The iodine species in the offgas have been sampled and results are l'3.1 e ., i shown in Table 6-4. The majcrity of the airborne radioiodine is in L..
4 the organic form. The source of organic iodine and the mechanism of its formation in the offgas has not been precisely identified. l:p The offgss iodines, however, are usually " older" than those found flij L..~
in the reactor water and condensate. This is shown in Table 6-5 : '::
4 where the isotopic distribution of the offgas isotcpes are compared 'N with the distribution in the reactor water. This apparent aging is M,. e probably due to tne delay incurred between the iodine deposition on ,b) material surfaces, such as carbon steel, and the subsequent formation ~
l1r p].
and release of organic iodine.
uy
[
, :q Organic iodide fctmation may be related to the production of CH4 , ,]
as a result of steel corrosion and carbide / hydrogen interaction "3l (6.12). The conversion cf crganic iodide from 1 2in 80*. humidity air in centact with galvanized steel at ambient temperatures has Q p
been recently reported (6.13). (24
.y$
6-7 R w&1
NEDO-25420 t,
if C
6.3.2 Separation of Icdin in the Condenser 1
}
(V' As was shown in Table 6.3, the r:ajority of the volatile radioiodine carried by the iteam distributes overanelmingly into the water ,
phase in the condenser leaving only trace quantities of organic ,' ej iodide in the gas phase. The icdire separation factor (S.F.) in l :n$
the cor. denser is defined as: ;
g.7. _ transoort rate of iodine into condensate (uCi/sec) -
r.;
release rate of iodine in non-condensables gas phase 'a (pCi/sec) ].fl a
The separation factor has been calculated (6.14) from field measure- @
ments at operating BWRs (6.15). Table 6.6 shows that separation factors for removal by condensation range from 140 to 5580 for inorganic iodir.e (HIO) and from 60 to 2460 for total iodine (1 '
2 HIO, and organic iodine). M kO The removal of iodine in the condenser by condensation processes is .."
expected to be very similar to the process whict occurs when steam ?
carrying radiciodine is rapidly condensed by tne suppression pool.
Suppression pool scrubbing, however, is expected to be more effective ~, -
in removing iodine because condensation in the pool will be very rapid thereby minimizing the centact time between steam and water. .j Once the steam is condensed iodine cust travel sertically through ;
J 6\ to 19 feet of water before it can escape f rom the pool. In - lJ addition, separation of iodine in the condenser occurs under a ,
-]
vacuum (1-2 psia) which will f avor the transport to the gas phase. .3 The separation factors or DFs for suppression puol crubbing thus g j are expected to be larger than those for condensation processes in ,
-1 the condenser. This indeed is true as will be seen in the next s
/a section. if
'M
- l 6.3.3 Torus Scrubbing of Iodine (6.16) 4 MS b? f Radiological measurements were made in a Mark I suppression pool .(
during a 20 minute High Pressure Coolant Injection System (HPCI) ,' l 6-8 '
N
E21 NE00-25420 W) a,-P, bd surveillance test while the plant was at 71% of rated power.
$R During the H?Cf test, steam from the reacter drives a turbine and L% '
is then discharged to the suppression pool as shown in figure 6.4.
,q Measursents of torus activities in the water and in the gas phase j!'.i were made both before and after the surveillance test; torus water [.$
phase and gas phase iodine concentration ate sumarized in Table 6.7.
Before the test, the primary iodine species in the torus air was
}_ f organic iodide (97.8% with only 2.2% as HIC. lodine activities in /di
. n:t the water before the test we e found to contain 85% as 1, and 15%
NJ
- i.%
as 103 . Elemental iodine and organic fractions were not detectable
$ieh in the torus water. Following the test the gas phase contained ";E 3'. -45% HIO. Partition factors defined as the ratio of iodine concentration in the gas to that in the wa.ter are also presented in f5
- p Table 6.7. .J y i til e,
Decontamination factors for pool scrubbing can be calculated frem: cy i w;;
0F = l-131 activity transported to rool '
.dn -
s change in I-131 activity in the gas phase g
y where the change in gas phase activity it, taken as the maximum E!$
a-observed charge in organic I-131 activity. This will be conservative, i.e. , predict a smaller 0F than if HIO concentration changes 2 were j)N used. -
@l
,N 2
bihi Relatively high background levels coupled with low iodine activity in the ste.m resulting from minimal fuel cladding defects produced
'[
h10 concentration charges which may have been on the same order as
.$.h QL the error in the analytical procedure. Thus the use of the maximum c. /
- G,8 change in cganic iodide activity will provide a conservative W estimate of the DF. ,a A
- ((
Nfj an
. s e t,l
_ .y
.- p gh
+ vj 6-9 ? >;
,0s$d
4 1
"2 NE00-25420 2
7 f
h The I-131 activity trann,arted to the pool is calculated from the reactor water I-131 concentration, tha .t.easu.ed 1 131 carryover, and total F. ass of steam transported to the pool.
From data in (Ref. 6.16) t?.e amount of I-131 transported to the
) pool is:
(5.6pCi/kg)(0.004)ll.44 x 10 6kg = 3.2x104 pCi 1-131 L
u The maximum change in orga.iic I-131 concentration from Te51e 6.7 o is:
I l 5.2 x 10 3.1 x 10-10 = 1.5 x 10-10 pci/cc h
H E The voltrae of the Monticello torus air space is 102,565 ft#from s
- Ref. (6.1.), then the activity in the gas phase is 3
j 3 (1.5x10'10 pCi/cc)(102,565 ft )(28,314 c_c) 3 ft b;l '
= 0.44pci I-131
,2 h
7 and the OF = 3.2 x 104 pti = 7.3 x 10 4 d 0.44pCi b
d
,b Based on this Ir.r;;c 0F, the suppressien pool is e'%.cted to effectively scrub radioi-dine ('*0) carried by the steam. !bt. frolt small
' ),.._.<7 scale oco? : rubbing tests (presented in Sect on .0) will provide i
v[,P addit...al information on the expected magnitude of the decentar.inatio'
.n lactor for dif ferent chemical forns of iodine, i.e. , I , CH I, HI 2 3 s and NaI.
t a .
I 1 The measured carryover was 0. C (Ref. 6.16) s a
C w
W
- i. C-10 >
[
I 6.3.4 Suwmary of Observations on Todine Behavior Operating BVRs The behavior of iodine in an operating BWR is well defined by measurements taken during plant tests. Radiochemical measurements at operati g plants indicate that gaseous iodine concentrations (HIO) are significantly attenuated by condensation processes s"ch
- as , hose observed in the main condenser ano during steam condentstion by the suppression pool. In view of these observations and the respective solubilities and volatilities of H'O and CsI, the removal '
factor for Csl by the suppression pool during a postulated accident ,
I is expected to be much greatve than the removal factor for HIO. ;
i.
2 e
e 1
9 l
6-11 9
l l NEDO-25420 -
i I
3 Table 6.1 a
131
[ 1 EASUREliElli RESULTS, T!11 - 2 (Ref. 5.16) I J
a FRACTION OF l LOCATION I (DECAY CORRECTED TO 3/28',(Ci) CORE INVENTORY l
l R.B. Sump 570,000 gal. at 0.012 pCi/mi on 8/28 = 1.4x107 Ci 20% .
l Aux. Building Mair.ly RCBT's A, B & C 0.21x107 Ci 3%
Liquids i Letdown Leminer- 112 gpm from 0730 to 1900. 3/26 ,
alizer ' at 1.7x104 Ci/g = 0.5x10' Ci 7% !
6 3 i R.B. Atmosphere 2.2x10 ft at .025 pCi/cc on 3/31
= 2x10 Ci 0.003% :
1 RCS 80,000 gal at .025 ;:Ci/g on 8/28 ,
= 0.4x107 Ci 6% !
R.B. Surfaces 1.31x10 8 e ,2 at 5.8 9Ci/cm2 en 0.4%
-6/1 = 2x105 Ci 2 8
1.3x10 cm 2 at 0.0041 pCi/cm on 8/30 - 3x105 Ci Auxiliary and Charcoal Filters - 124 Ci %. 00'V.%
Fuel Handling Building Surfaces - 130 Ci haildings Er.vironmental Release (3/28- .
4/30) -38 C1
(#
OTSG B Liquid 23,000 gals - 440 Ci % .0006%
TOTAL 25xlC Ci 37% ,
i 4
(.0 R.E. = Reactor Building (b) R.C.S. = Reactor Coolant System (c) OTSC = Once Through Steam Generator .
i 6-12
NE00-25420 TABLE 6.E CHEMICAL FORMS OF RADIOI0 DINE IN BWR PRIMARY COOLANT (%)
REACTOR (DAiE) 1 2
I 10 3
ORGANIC CH 3 Itt nURING NORMAL OPERATION BWR No. 13 (7/1972) <1 80 20 <0.1 -t BWR'No. 2 (11/1972) 8 9 10 4 0.03 I
BWR No. 9 (12/1972) 1 65 34 <0.1 <0.02 BWR No. o '.1/1973) <0.1 52 48 <0.1 -
BWR fio. 9 (3/1974) 3 79 18 <0.1 -
BWR No. 16 (7/1974) <n.1 /7 23 <0.1 -
BWR No. 15 (9/1975) 4 04 12 <0.1 -
BWR No. 34 (C/1977) <0.1 67 33 <0.1 -
DURING SHUTDOWN:
BWR No. 9 (4/1973) < 0.1 2 98 <0.1 _
BWR No. 9 (3/1974) <0.1 4 96 <0.1 -
BWR No. 16 (9/1975) 3 67 30 <0.1 -
BWR No. 16 (2/1976)** -
18 82 - -
- Hay include HIO
- DurirJ hot standby i Not measured it Determined by distillation, the contribution from the isotopic exchange between I and carrier CH31 was estimated to be s0.02% of the total iodine activity. ,
(Ref. 6.9) 6-13 l
NED0-25420 TAP'E 6.3 CHEMICAL FORMS OF RADI0IODIt'E IN THE CONDENSATE (%)
REACTOR (DATE) If I'* 10} ORGANIC CH 3 Itt l
BWR No. 13 (7/1972) 6 82 10 2 -t BWR No. 2 (11/1972) 2 85 9 4 -
I
- BWR No. 9 (12/1972) 8 88 3 1 0.06
=
L"JR No.15 (7/1974) 3 93 4 1 ,
i Hay ir.clude HIO t i40t .aeasured it Determined by distillation, the contribution fro. the isntopic exchange between I and carrier CH31 was esticated to be s0.02% of the total iodine activity.
4
- s (Ref. 6.9) l 6-14
NEDO-25420 TABLE 6.4 RAD 1010 DINE SPECIES IN OFF-GAS AT THE SJAE* (%)
REACTOR (DATE) 1 2
"HIO" ORGANIC BWTs No. 10 1.5 16.9 81.6 BWR No. 13 30.2 34.6 3f.0 BWR No. 16 5?.3 18.3 29.4 Bk"1 No.11 4.5 20.2 75.3 BWR No. 12 6.7 22.6 70.7 EVR No. 13 (7/1972) 30 -t 70 t BWR No. 2 (11/1972) 2 -t 98 i BWR No. 9 (12/1972) 12 -t 87 t BWR No. 9 (4/1973) 12 6 82 BWR No. 34 (5/1977) 10 48 42
- SJAE = Steam Jet Air Ejector t HIO not determined; the organic fraction may include HIO.
(Ref. 6.9) 6-15
NEDO-25420 TA3LE 6.5 COMPARISCN OF IODINE ACTIVITY CONCENTRATIONS IN BWR NO.13 REACTOR WATER AND Oc GAS
- 150TCFE REACTOR WATER OFF-GAS CONCENTRATION CONCENTRATION pCi/1** RELATIVE TO I-131 pCi/1** RELATIVE TO I-131 1-131 (8.04 D) 3.35 1 0.0058 1 I-133 (20.8 H) 9.55 2.8 0.0139 2 I-133 (6.59 H) 20.4 6.1 0.0251 3.7 I-132 (2.29 H) 25.T 7.5 0.0233 3.4 I-134 (52.6 M) 53.8 16 0.0499 7.3
- Date of analysis: 7-31-72
- Volume measurco at ambient temperature (Ref. 6.9) 6-16
NEDO-25420 TABLE 6.6 10 DINE-131 SEFARATICN FACTOR IN BWR HOT-WELL (Reference 6.14)
Gas Phase Radiciodine Separatign Chemical Compsition,%a factor Reactar (date 1 7
H10 Organic Inorganic Total d
BWR No. 9 (12/1972) --- --- --- ---
1296 BWR No. 10 (1973)' 1. 5 16.9 81.6 690 111 BWR No. 11 (1973)' 4.5 20.2 75.3 2418 498 BWR No. 12 (1973)' 6.7 22.6 70.7 1940 407 BWR No. 13 (1973)3 30 2 34.6 35.0 141 78
~
BV9 No. 13 (7/1972) --- --- --- --- ---
BWR No. 16 (1973)' 52.3 18.3 29.4 55S4 2462 ;
BWR No. 34 (5/1977) 10.0 48.0 42.0 477 243 p i
F L
d iij
- a. Measured at steam jet air ejector sample point.
g
- b. See text for definition.
}
I. '
- c. Including 12and HIO. g f,;
- d. Chemical torm not deterrained. _
a h
- e. Data obtained from Ref (6.14). 3 d
9 A
H
,k 6 ,
k y
d
,4 6-17 -
b
TABLE 6.7 Torus Airborr.e lodine Activities and Their Gas /Wter Partition Sample Water phase. Concentration in gaa _
location prt/rr _
Cas these,ycl/ela Partitlei. Concentratsan in e.ater siid Time ,1 1,31 1-133 1-133 I-It) _
I-111 1-133 orgente Mio . Oraule uno Total Inorgente _ Total Ince gang Torus Sample **
Stat?on ,
2/2 7, 9 : 30 - 2.%s10'
' - 3.68s10
-10 8.16ml0'I 4.42:10 -12 -
1.5 10'
- 3.3s10' - -
2/28, 9:50 (17.8%) (2.21) (sl001)
-10 C-2 1.8:10 - 1.6u10 8.21stD'I - - 1.34:10 4.56 10 - -
3/3, 1825-1955 (66.!!) (33.91)
-0 -10 C-2 2.46 10' -
2.76s10 2.26:10 - -
2.04:10 9.2s10' - -
3/3, 2026-2058 (55.2) (451)
-0 -10 C-2 1.87s10"II'5.84:10' 2.b a10 3,gg,33 3.45sto'II 3.19:10 *II 1.9a10' 6.la10- 1.3:10' 5.5a10' 5 to 3/3, 2155-0015 (67.5%) (32.51) (521) (481) g
-10 *II *II 2.65 10 2.4s10' C-2 ,3 ,3 1.18:10 4.8?a10 8.51 10 3.69:10"II 8.9:10" 7.2a10' 3/4. 1121-1421 1.87a10 5.12:10 (69.9t) (30.11) (69.8%) (30.21) 4 N
-10 O C-1 - -
3.74s10 , , , , , ,
3/3, 2026-2058 (t ot al)
C-3 2.36:10-4 - 4.81 slo'IO - - 2.05 10 - - -
3/3, 2076- 's (total)
C-4 2.36 10*I" - 5.2 10 -10 - - 2.2n10' - - -
3/3, 2026-2053 (total)
C-5 2.82a10 ' -
2.4n10
-10 - - 8.$s10 3/3, 2076-2058 (total) .
- No iodine actietties in 1 2or Particulate forms were found.
tThe I-131 activity in water was found to be ~ 85% as I", and-15% as 103 , and both 12 and organic fractions were not detectable.
ttAvg. of three samples taken during the sampilng period.
- HPCI Test. Occurred on 3/3.
Ref, 6,16
-x;gmsg ; . gr.o.e,.ame-ve m % % a w-v. .. + w .sm*- w 5 .
NE00-25420 ty h
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ta
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Figure 6.1 Fission Product Transport Pathvay at TMI-2 i5 in d
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2/263E; 2/27'6s 2/2E/8 2829/F.. 3 ,13 0
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DATE
'1 1 Figure 6.2. Xe-133 and Icdine-131 Levels in Containment Atmosphere j,.3 at Crystal River ,<
(Ref. 6.3) ,;d
,e e.
4
..t h I Lf
- 'O
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,t,t*i II6
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s 13
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MI NE00-25420 '.4
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Y k k l -
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LwLl= "9 a'
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- .a
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~
(-T ,
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N
- _-- f_z ::f , .j,\
x \ I [
b 40.E paa OF PRIMARY W ATER ', '
(85 Ce l 13 0 (d .
Figure 6.3. 1-131 Distribution at Crystal River l' nit 3 on 3/2/80
,g.,
?
(Ref. 6.4) 4h a
l'
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li s.-
'0840$ CONDENSATE l
STORAGE 2[
4' g i .7
! 9P l j,i:v 2
l Jk j' ) Pl *
- - - s s
/
l
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,1 v HPCI evue
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S 9
7.0 lodineandParticulateScrubbingByPoofsofWater hp
?
?e1 An extensive review of tNe literature has t,can made to identify and 'b docu: rent observed iodine attenuation factor, from pool scrubbing Lj2 experiments. This searco identified severa, papers that are applicable ,y to BWR~ accident mechanisms. ,
h 1 The majority of the tests used iodine in the elemental form I .
M 2 hd
,o However, several investigators used HI, CH I, and small diaineter 3,Q 3
insoluble particles. Attenuation of particles is of importance ,'
..s M.
because fission preducts released during an ac.ident may be in the p ,
form of small particles or in aerosol form. The assumption that an h 41 aerosol exists has been supported by Sandia tests (7.1). If a core melt occurs, the subsecuent hign te@eratures will result in vapori-zation of certain metals (structural components, trace metal in the k fuel,etc.). As these materials are transported from the core @@j region, they will beccme cooler and condense into an aerosol. This it) s the basis for examining aerosol transpcrt in suppression systems. j
- 5. ,'
P e.,
Su:,naries frem each of the relevant pacers are presented in the d j
following sections. The pa;:ers are grouped into two categories: 9 scrubbing by subcooled pecls and' scrubbing by boiling pools, hN' 7.1 Scrubbiro By Subcocled Pools 3 W .
7.1.1 Fission Product Entrairment in Pressure Sur,pression Systems (7.2) y m
, :l 8 t
In 1959 General Electric conducted an exploratory test to evaluate og c
the effectiveness of a pressure suppression system to act as a 3.;
barrier to the release of fission products following a reactor l
accident. The tests were run with noble gases (xenon and krypton), @
M halogen (iodine), a soluble salt (sodium iodide), and insoluble particulate matter (2pm florescent zinc sulfide particles). Q]ie, d
.n
', .'.i $ *
[
,M n
N t
ve l &
( 7-1 YU w
l
\
t
q' NEDO-25420
,1 E
~ .1 As shown in Figure 7.1.1, the test facility consisted of three Ma interconnected pressure vessels which simulated the reactor pressure
,f vessel, drywell and the suppression poo" Water in the reactor f wessel is heated to saturation conditions at 1000 psig and then f dischargad through an orifice pli.'e into the drywell by breaking a rupture disc. Part o+ the water, after it flows through the orifice, flashes to steam which in turn discharges from the dryvell through vent pipes into che water " ol where the steam is condensed.
][ A more detailed description of the .. . f acility can be found in 1 Referer.:e 7.2.
q
[i] Fission product trace s were placed inside the reactor vessel which
.u y
a was filled with a chargc of wi.ter. The ater was then electricolly ,,
{b heated to saturated conditions, at which time it was dis t rged to tne drywell and suppression chamber. Concentra6 ions in the wetw?ll ni air space were then determined by taking grab samples. Test
,i results presented in Table 7.1.1 show thct th<. DFs ra"2ed frem V-2 C M 10"-1Y for ludine and insoluble particulates.
[ The test results show that the su,pression pool effectively retained 3
all but a very sr-11 ar eunt of the : dine, Nal, and 2 pa pa-ticles, N.s while a significant pnrtion of the noble cases were released tecm the pcal.
j Although t he tests 6.*e.e exploratory in natuec, the test res';1ts f showed taat a suppre:,sion pnol is an extremely effective carrier to j the release of fisi'c.) products following t. reactor acc' dent.
5 C 7.1.2 Iodine .*!emoval By a Scale Model of the S.G.H. W. Reactor Vented Stea, Suppression System (7.3) a
[ Reference 7.3 reported results from a series sf tests in which J
malecular fadine and smai. nickel-chromium particles (0.06 pm) were
] re: eased into steam, air. and steam-air mixtures. These mixtures j
.q l [1
~
72 .
-~ -- ~ ~
NEDO-25420 were then allowed to pass through a pool of water and pool .decontami-nation factors were measured. The experimental apparatus, shown in Figure 7.1.2 was 6 one-third scale model of a section of the SGHW di: charge pipes. It was designed to reproduce primary system
~
failure conditions by properly scaling discharge pipe, pool surface area and steam mass flow. A correctly scaled amount of heat passed 1 to the s apression pool, but gas velaities in th( discharge pipe 1
and depth of discharge pipe immersira in the pool were full scale, 7} The mass of 2I used wcs equivalent to the release of a few perr.nt j
of the content from several reactor fuel asser.blies.
1 Tables 7.1.2 and 7.'. 3 summarize test results with iodine. Decontami-nation factors (DF) were determir.ed by taking the ratio of iodine in the suppression pool water to that found frca sampled air above Is - t .
y the pool. Tht values obtained varied depending on the, air ta steam ratio. Decontamination factors for 1 ranged 2
from 20 to 320.
l Iodine release time, pool terrerature, and depth of ditcharge pipe i irT,ar. ion were found to have little effect on the DF.
J 1 Table 7.1.4 presents results from the 0.06 pm particle tests. The
] DFs for particles were found to vary only slightly with air / steam i!j ratios and were hignest at the larger ratios. Lower DFs were f
obtained at the high steam flow rates than at the lower ones (compare
] tests 21 and 22 with 17 and 18). However, at the high steam flow rates a :::uch greater proportion of the to.rl activity was found on the demister, so the smaller DFs may have been due to sprcy carried up under more vigorous conditions. The very large DFs found with 75 wt % air are much greater than ey.pected (refer to Table 7.1.3).
b_
a A possible explanaticn of these results is that when cold air is mixed with saturated stecm the air may become highly supersaturated.
l Under these conditions condenstcion niight occur on the subnicron particles which would grow into droplets r.ach langer and more s easily removed than the aerosol itself (7.3). In this study the air was cold and the steam was saturated at the entry of the mixing 1
. 4 4
7-3 d >
1 F
NED0-25420 8
41 r
[ chamber. In other tests, when the air was not preheat:d, signifi-cantly greater DFs were obtained (7.3). Tiis mechanism may also
,, explain tne large proportior, of the activity found on the demister, f since i+ would remove droplets although it would not remove the 5 aerosol it:alf. Decontamination factors were reported between 15
/ and 1680.
b 3 7.1.3 Iodine Clean-up in a Steam Suppression System (7.4) ni Experimental nrk presented in this Canadian paper was similar in j scope to that of Hillary et.al. (7.3). The objective was to assess j the effectiveness of a steam suppression pool in removing fission j products released during a reactor accident. Eiemental iodine q (12), methyl iodide (CH3I), and hydrogen iodide (HI) were used in L the experiments to simulate the behavior of gar.eous forms of iodine likely to be present following a reactor accident. In cddition, a aerosol particles (0.06 p diameter), cimulating the behavior of 3 particulate material to which iodine might be attached, were also tested.
s Tests sere run.using both staall and large scale equipment. The
- i test equipment is shown diagrammatically in Figures 7.1.3 and 7.1.4. The small scale tests useu 3 mm discharge pipes immersed in 6 cm of water. Air coritaining either an iodine species or a parti- ,
j culate ceroso. was injected into the steE?. fles and tha r.ixture 1
9 passeo through the discharge pipe and into the water pool. Vapor velocities in the discharge pipe were atout 34 m/sec. air was drawn from above the pool surface, collected and analyzed.
The large scale test equipment consisted of a stainless steel tank 41/2 feet deep, 3 feet long and 11/2 feet v :de containing the 3 water pool with single and mu1*iple discharge pipe arrangements.
The discharge pipes were 50 mm in diameter and 50 cm below the pool
,; surface. Air containing either icdine species or particulates
! mixed with steam was passed through the discharge pipes and a.r samples above the pool were sampled and analyzed. The pool temperature was 50 C (122 F) and the pH was 7.
7-4 i
~ .
NED0-25420 Results, in the form of a decontamination f actor, are srown in Figures 7.1.5 and 7.1.6 as a function of discharge pipe diameter, air /hteam rates, and final concentration. OFs ranged between 10 and 2000 for eierental iodine. DFs ranged between 10-1000 for HI, between 50-500 for 0.6 mm particles and between 1-5 for CH3 I, and were dependent on pool depth, velocity, and % air / steam mixtute.
7.1.4 Diffusion of Iodine in Water. The PIREE Experimant (7.5)
The purpose of the PIREE Experiment sponsored by the French CF' __.
to investigate the diffusica of radioactive iodine in pLre water ertrained in a flow of hot pressurized carbon dioxide. The experi-
- mental apparatus shown in Figure 7.1.7 consisted of a 42 m3 (11,160 gal) stainless steel tank (6 meters x 3r. dia) filled with demineralized :
water. A loop consisting of a heater, heat exchanger, compressor, f and CO c
, cylinders was connected to the tank so that co.ntanineted g .
CO 2 could be injected int the water at a pressure of 20 kg/cm (26u psi) ano 400 L.
The C0 2 carrier gas was contaminated with iodine-131 (I2) pr duced by the oxidation of active sodium iodide. Altogether, fourteen experiments were performed with the following variatfor.s: duration ,
of the injection varied trer 20 to 80 seconds; concent.ation of icdine in the carbon dioxide at the point of injection ranged fro 7 ,
10
-8 6 g/g; temperature of the circuit ranged from 100 to
~0 300*C; total 1-131 in t9e water was varied from 10 Ci to 15 Ci and flow rates ranged from 2 to 20 g/s. The air s'Jace aoove the tank of water was analyzed and the transfer factor of iodine through the water was calculated as follows: .
Aa K=
Ae+Aa 7-5 .
a
HED0-25420
. where Aa is the radioactivity of the iodine which passed through the water, and Ae is the radioactivity of the iodine trapped in the water, measured as soon as possible after injection.
Measured transfer factors, i.e., the retention capability of the water, are shown as Table 7.1.5 with the average error calculated on the basis of the uncertainties in the determination of the activities. Pased on the results presented it was concluoed that the transfer factor is independent of the parameters invettigated and thus variations in activ.ty in the water from ~3 -5 to 15 Ci did not have any noticeable effect on the retention of iodine. The transfer factor, on the other hand, decreased with increasing height o' the water column, and with decreasing flow of the carrier gas. Fo.' a flow of 20 g/s the trensfer factor varied from 1x10'2 to 6x10'4 as the level vari.d from 2 to 6 meters, while at a flow rate of 2 g/s under the same canditions, the transfer factor varied frcm 10 to 10'4 .
The tr:nsfer factor is related to the decontamination factor DF by the following relationship:
DF = 1,X where the OF is defined as the r-tio of total activity in the air above the water poo! to the total activity in the water prol.
Decontamination factors tor each of the 14 experiments were calculated and results are show.1 in Table 7.1.5.
4 DFs range ' rom 71 to 10 and are depencent on the height of water in the tank, and the flow rate of the carbon dioxide carrier gas.
The report concludes that " water has a very high scrubbing power" and that iodine releases from a water pool would be only a small fraction of the total iodine released to the paol.
7-6
3 NED0-25420 7.1.5 Elemental Iodine Retention by Pressure-Suppression Pools (7.6)
Experiments desigr.ed to provide information atout the retention of 12by pressure suppression pools were conducted at the Oak Ridge National Laboratory. Tast conditions were selected to conform as closely as possible to the design basis acc! dent. Elemental iodine ,
was used as the simulated fission product. The experimental apparatus shown in Figure /.l.8 :onsisted of a large aluminum t6nk which contained approximately 260 gallons of water. Saturated steam at 125 psig was supplied to the tank through siculated downcomer pipes. The downcomer pipes wera 0.680 inches ID. Elemental iodir.e was injected into the steam in quantities sufficient to produce concentrations of 0.5-10 ppm in the pool.
The investigators cons;dered the following cases: (1) saturated steam (containing elemental iodine) injected into water initially ccataining no iodine, (2) saturat 1 steam (containing clemental iodine) injected into water having a knowa initial concentratica of iodine; (3) the injection of saturated steam plus 2 wt % air (this mixture containing elemental iod ne) into water initially containing no iodine; (4) the injection of saturated steam plus 2 wt % air (the mixture containing elemental iodine) into water naving a known initial concentration of iodine; (5) the sa.ae as Ca:e 3 except the pool contained 1000 ppm C/0 2 and was at a pH of 9; (6) the sa.t.e as Case 4 but in addition the pool contained a known concentration of iodine; (7) the same as Cases (3) and (4) but with multiple down-comers.
7-7
i, NEDO-25420 Cases 1-4 involved final iodine concentrations ranging from 0.5 to 10 ppm in the pool water. The other cases involved iodine corcen-trations of 5 to 10 ppm in the pool water.
Approximately 200 lbs of steam were discharged to the pool in each test. Steam mass flow rates were from 79,000 to 238,000 lb hr'I
~
ft . In the first few experiments, a relativeiy large number of samples were taken at various elevations in the tank to determine the extent of possible concentration gradients. The data indicated, however, that gradients were negligible. In the subsequent experi-ments fewer suples were taken and these were primarily at the horizontal midplane of the tank.
Table 7.1.6 lists the quantity of iodine removed normalized to nicro-grams of iodine per pound of injected steam for each of the various test conditions. The data are grouped according to the quantity of iodine injected into the system and subgrouped accreding to the steam flow rates and steam-air e.ixture ratios. These data
- vere usec to calculate decontamination factors for each test by determining the ratio cf total iodine i.ijected to the total iodine removed. OFs ranged from 57 to 1445. Under all conditions tested the quantity of iodine released was a small fraction of the total, the largest value (snallest DF = 57) was $1.6%.
7.1.6 Scale Model Tests of Methyl ledide Removal in Suceression Pools (7.7)
In a General Electric Company test, a 1/10,000 scale model of a BWR pressure suppression system (Mark I design) was constructed to study the absorption of methyl iodide in suppression pools under loss-of-coolant accident conditions. The major components of the scale model contair.,ent (pressure vessel, drywell and wetwell) and associated piping are shown in figure 7.1.9. Table 7.1.7 1 its data relative to the design and actual size of the facility.
7-8 ,
HE00-?5420 Results are presented in Table 7.1.8 as a function of quantity of methyl iodide injected, (C); pH of the pool; depth of submergence of downcomer, (S); time of methyl iodide release, (R); and pool temperature. The effectiveness of the pool in removing methyl iodide is presented as percentage absorption in tne pool. Absorption '
(A), is related to the decontamination f actor OF by the relations, hip: t A= 100% - 100%
DF The absorption varied from 11% to 69% resulting in OFs ranging from 1.1 - 3.0. The results showed that absorption of methyl iodide in the model suppression pool was not affected by increasing the
=
containment inventory of CH3 1 from 0.2 to 2 mg. No change in absorption was observed due to changes in the radial cosition of the downcomer, or changes in the pH of the pool from 7 to 10. Pool absorption for a downcomer submersion depth of I foot was about 20%
lewer tnan for a 4 toot su::morsion; the absorption for an initial '"
pool temperature of 150 F was about 28% lower than that for a pool te:rperature of 90 F. An increase in the delay time hetween the r start of the blowdown and the addition of CH3 I to the drywell
- initially increased the pool a5 sorption. For delays greater than 2 -
seconds, the pool absorption decrcased with increasing delay until the absorption for a 6 second delay was about 24% less than for a zero delay.
(
7.1.7 The Marvilren Full Scale Centaiteent Exaeriments (7.8) ,
A series of 16 blowdown tests was perfortred in the full scale containment of the Marviken oower station. Ine behavior of iodine ?
in the containment during the blowdown and post blowdown period was examined so that the transport behavior of iodine in a multi-compartment, pressure-suppressien type containment could be more h fully understood. The main objectives of the tests were to l
investigate: (1) the removal of methyl iodide (CH3 I) DY " I"#81 b'
7-9 k;n Y
NEDO-25420 processes from the containment atmosphere during a lit.ited time; (2) the removal of elemental iodine in Kg ouantities, (3) the effects of spray cooling on the removal of iodine from the contain-
. ment at..iosphere; (4) the trapping of iodine in the wetwell pool; and (5) the leakage of iodine free the containment.
The test facility consisted of a pressure vessel and the containment system of the shutdown Marviken nuclear plant. A cross sectional view of the containment is shown in Figure 7.1.10. The containment is divided into two compartments, the drywell which surrounds the pressure vessel, and the wetwell wtich contains a condensation water pool connected to the drywell by a vent system. The drywell, ..
consisting of several corr.0artments (shown as encircled numbers) in Figure 7.1.11 has a total air volume of 1934 m3 including the vent system to the normal water level inside the vent pipes. 'The wetwell lies below the drywell; the normal depth of the wetwell water pool is 4.5 m, giving a pool volume of 560 m3 (105 gallons) and a wetwell air space of 1584 m3 . The vent system consists of four channels (with a 1.2 m 10) connected to a header located in the wetwell air space.
From this header SS vent pipes (with a 0.3 m ID) lead vertically downward to the wetwell pool. The normal submergence depth is 2.8 m.
The pressure vessel has an inside diameter of 5.22 m and is 25.6 meters high. The net volume of the vessel is 414 m3 The vessel is designed for a pressure of 834 psia and a tercerature of 272 C.
The simulation of a pipe rupture was possible at three locations:
l in the upper crywell (Room 124); and in the lower drywell (Room 122) where simulated breaks of the feedwater system and main steam line system were possible. The equipr.ent for injecting iodine was located in Room 121 just behind a door cpening to Room 111 (See Figure 7.1.11). During the injection, iodine was carried by air I into Room 111. Gas sampling of iodine occurred at five sample l points in the drywell and two in the wetwell atmosphere (sample l
l points are shown as a, b, and c in Figure 7.1.11). Water samples
! from the pool could be obtained from the center of the pool and at the periphery of the pool.
7-10
- j NE00-25420
)
i i
l A series of 16 blowdown tests were conducted. A brief description ,
of each test ar.d initial conditions is summarized in Table 7.1.9. "
The events and experimental conditions for each test, including the quantity of iodine injected and the time of injection, are summarized 7 in Table 7.1.10. A more detailed description of each test can be found in Reference 7.0.
The transport behavior during the blowdown period was investigated -
J first using methyl iodide. CH I3 was injected prior to the blowdown to investigate the transport, ges removal and trapping in the water
}
pool ty natural processes. When CH 1 3 I was injected prior to the t
blowdown, it was completely removed from the drywell atmosphere f
during the blowdown phase. In the wetwell, CH 3 I was partially G present in the pool water and in the wetwell atmosphere. Test
- V times for these experiments were 3-4 hours and these experiments Z M
were too short for equilibrium conditions to be reached. Following j the blowdown, CH 3 1 was retransferred into the drywell through we j wetwell pressure relief valves. Results from two typical test-
]
(Run 4 and Run 8) are presented in Tables 7.1.11 and 7.1.12. From gj these data decontaminatioa factors can be calculated for various -
times. These DFs along with theoretical equilibrium DFs are presented in Table 7.1.15. The experimental DFs vary between 1.2 to 4.9 and Y.
are larger than the predicted equilibrium values. j.j The transport behavior of elemental iodine during the post-blowdown N period was investigated in Runs 15 and 16. In Run 15 sprcy cooling v f
of the los .r drywell occurred, while spray cooling in the lower i drywell did not occur in Run 16. Two kg of elemental iodine was (;(
n released durirg the post-browdown period. Injection, for example, 7 c' '
i occurred at 1 and 0.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />, however, the blowdown period was over p at 0.6 and 0.08 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> respectively for runs 15 and 16 (Refer to i Table 7.1.10), therefore very little steam was present to transport
{,
the I 2through the pool. In both runs the concentration levels in the upper drywell and wetwell gas spaces were extremely low and at t times the concentrations were below the detection limit. The 'h i y I
?
7-11 1
O g
(
NE00-25420 ,", i
);$
ET concentration levels in the lower drywell were also Icw and of the ik same order in both runs. Due to the high rate of oeposition of d
elemental iodine, the advantages of spray cooling were concluded to
}
4 be only marginal. Results from runs 15 and 16 are presentea in Tables 7.1.13, 7.1.14, and 7.1.16. OFs of 225 and 265 were reported fi for runs 15 and 16 respectively (7.8) af ter 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.
Tne results indicate that a DF of at least 200 can be expected for '
12 duri.1g the post blowdown period due to natural deposition processes.
One would have expected a much larger DF if sufficient steam had 3 been present to transport the 1 2 through the pool during the post blowdown period. Unfortunately this case was not investigated.
)e p
l(
7.1.8 Simulation of Container Ventino Under Sea Water (7.9) O r
- }:.
[ In this Mine Safety Appliances Co. experiment the under water 0,1 release of radioactive contaminants from a reactor compartment was h simulateo oy injecting contaminated steam and air from an open end 9 2 inch pipe in 10 feet of sea water. A schematic of the test ,
apparatus is shown in Figure 7.1.12. Two 40 gal, autoclaves were f
used to bring contaminated coolant to a temperature of 550*F and a C pressure of 1065 psig. Diping was arranged so that water from both .!
autoclaves could be discharged simultaneously into an empty 10,000 ha gallon tank. A 2-inch pipe led frcm the top of the tank to the sea l]
water container which was constructed from five SS gallon barrels )
welded end to end. Radioactive Na2 C0 2 , RbC1,21 , and p3Y 0 wre hI added to the reactor solution in concentrations of 2, 7.5, 1, and Q w
10 ppm respectively. The contaminated coolant was disti.arged to the 10,000 gallon tant. The gases passed through the sea water and [A were collected and analyzed. $
W
~. )
Results from the tests are sowtarized in Tables 7.1.17 and 7.1.18. ' !)
The major portion of the contaminants were observed to be retained N;)
in the 10,000 gallon tank, thus hindering the evaluation of the sea a
water column as a decontaminating agent. The area in the 10,000 g i h) c V) 7-12 h
- _- . M ..
' ~
- NED0-25420 9
m) gallon tink where the unflashed water and condensed water collected hI was found to contain the highest activity. Decontamination factors [d;p for plateout and water removal mechanisms were calculated from the % <d data in Table 7.1.17 and results are summarized in Table 7.1.18. j/
For elemental iodine (12 ) plateout DFs were on the order of 104 , fj and a DF=6 was calculated between the water and the air space above y,i the water. Plateout DFs were on the order of 105 for Na* and Rb*,
and were approximately 10 for Y 0 . DFs between the water and the h
2 3* []
air were 30 and 50 for Y23 0 and Rb respectively. Plateout phenomena 4
in the tank occurreo before pool scrubbing occurred.
{,
7.1.9 Scrubbing of Iodine in a Column of Water (7.10) [h )
A series of small scale tests was performed by Westinghouse to obtain a quantitative measure of the iodine and carbon dioxide 9 absorption from gas bubbles by a surrounding liquid. The test ij' d
apparatus was set up to provide release of gas bubbles of con-trollea diameter at the bottom of a water column. Sampling devices h
above the colutn allowed measurement of the iodine or carbon dioxide !'[
remaining in the gas bubbles reaching the water column surface.
Decontamination factors were computed as the ratio of total trace g
Q component injected in the bubbles to that found by the gas space .
samplers. M N
Tne test assembly shown in Figure 7.1.13 consisted of ar eight foot W
glass column having a cominal incide diameter of nine inches.
i i,:
a .-
Gaseous iodine (Iy ) in helium was released at the botton of a y column of liquid. The iodine vapor carried by helium was injected 'S into the column via a heat traced stainless steel line to prevent iodine deposition in the injection line. The solution i.. the h
$a column was maintained at 120*F. The gas space above the ligesc column was purged with helium at a rate of 2 liters / min?
- to Q
remove the injected helium plus whatever iodine emerged frcm the Q
column of liquid. The purr;e flow was initiated before the iodine }h Jr "t
bubble injection began and continued for ten minutes af ter the C p ',0 injection had been completed. The results shown in Table 7.1.19
$f H.',
A
?t 7-13 ([;
h{
. 9 NE00-25420
$ I ih include all the pertinent inforttation on test conditions and iodine 1.;
inventory. Decontamination factors are calculated from the data in ,j this table as the ratio of iodine injected, to the sum of iodine collected above tne test solution. Results for different depths of solution and varying bubble diameters are also presented. OFs are /
~
seen to range from 88-1980. For depths of release greater than 7 7 feet the OFs were measured to be greater than 200. f 1
7.1.10 Scrubbing of Particles by Water (7.11) ;.
.H.
Particulate scrubbicg by a pool of water was investigated a; the Tq Hanford Engineering Development Laboratory. One pound of soaium $
iodide (Nal) was vaporized at 1700*F and carried by 0.5 CFM N2 'j through a foot deep pool of water containing a gravel bed. The . ,3 I
resultant scrubber efficiency of the pool was 99% or a DF=100. In other tests a DF=20 was cbserved for 4p sodium oxide (Na2 0) particles - "l through 2 fcct of atcr.
et 7.2 Iodine Scrubbin.1 by Boiling Pools .
7.2.1 Trapoing of Iodine in Water Pools at 100*C (7.12) I q
Unlike tne earlier references reviewed, this Swedish reference reports data on trapping of 1 2by water pools under satu ated j)'e) conditions, i.e., the pool is boiling. In addition the pool pH was e/[
varied between 1 and 12 and pool iodine concentration was varied ,4 ii between (0.0001-1 ppm) to assess the overall effectiveness of the ,1 pool to trap 1 . Two types of experiments were conducted.
2 In the [
first, a laboratory study evaluated iodine release from boiling ,
water to which elemental iodine was added. In the second, large-scale jq experiments investigated which condi'.icns were needed for direct
- 1 breakthrough of iodine without reaction with the pool water. The 3 1
latter tests are the most relevant to this study.
Q The large-scale system is shewn in Figure 7.2.1. The pool water ,
was preheated by means of a heat exchsnger in the poolbottom. An U
- e
'l
- )
7-14 .
u
NED0-25420 P b -
- 9 d
accident was simulated by inserting a 17 mm diameter tube vertically b in the pool. Through this tube, superheated steam at 175-320*C and '
x at flow rates of 0.1-0.4 kg/sec (800-3000 lbs/hr) was injected into the water just above the pool bottom. Elemental lodine, dissolved 9 in water, was injected into the steam, upstream of the pool. 1 Samples were taken from bcth the pool water and the exhausted steam. Two experiments were performed in the large scale system.
In the first experimer.t iodine was added to the steam for 20 minutes, @
followed by a period with pure steam injection. The second Lk experiment cone.isted of simple boiling runs as in the smaller i laboratory study, and involved an instantaneous addition of iodine.
d l: '
g The total amount of iodine released in the boiling runs is shown in .,5 Figure 7.2.2. These results were consistent with the small scale laboratory study and indicated a strong dependency on the pool pH.
Note that higher retention factors were observed under basic conditior.s.
]
[
Y m.
In the experiment in which iodine at a constant rate of 1.3 mg/sec EM was injected with steam, the concentration of icdinc in the exhausted Ich 4..
steam increased with time as shown in Figure 7.2.3. A rapid decrea:e 1 of iodine concentration in the exhausted steam occurred as sot .s [i]
the addition of (1 )2was stopped, although the steam continued to .h be injected into the pool. Differential decontamination factors, M L
defined as the ratio of the con entration of iodir.e in th* injected M Ur steam to the concentration of iodine in the exhausted steam, were $51 l
t detennined for the first 10 minutes and the direct breakthrough of aq'j f ..
! iodine was "of regligible significance." The increase in iodine in f.y l
the exhausted steam during the period of addition (refer to Figure G 7.2.3) appeared to depend on the decreased degrec of hydrolysis as b the total concentration in the pool became higher. [<
[
M Decontamination factors during the period of iodine additica are h -
shown in Figure 7.2.4 as a fraction of iodine concentration in the %j pool. The DFs ranged from 3 to over 100. Major conclusions from ,9 xJ the tests were that no dependence of DF on water height or steam
}'
flow rate was observeu, however, a rather strong dependence on pool (12) concentration and pool pH was observed. 6; 7
a 7-15 ;72I Yx
'. 5,,
Qj NE00-25420 g
q
' .'.1 .
. 7.2.2 Transfor si lodine From Annecus solutio s to Saturateo Vapor
_ g (7.13) m
,s
@ lhis lussian tes* .< s similar to tha previous test in that transpor*.
} of element i iodine (!j) f rom a saturated liquid was investigatG f r.ing s;all scale latalatory equipent. A schematic cf the test 3
Z spparr'.us i, shown in Figure 7.".5. Stean from a steam generator g was-fed through a bubbling column. The column was fitteo with a
'.hermosteam jacke; so that steem condensation on the walls would
] not occur. Experiments were conducted at steam flow rates of a.t 0.2-1.0 kg/hr. with vapor spaces of greater tnan 70 cm above the 7 water phase. Very low (10 -5%) carryover of roisture drops by steam
(? occurred. Investigations were ;arried out at three press"res: 27, 57, 142 psia (with boiling points at 113, 143, and 179'C, respec-
-a pl tively). The experiments were conducted over a broad ange of '
5.._; iodine concentrations (10-103 ppm) and pH values of the solution (5.5-10.5) in the calumn. Samples of stealt were taken for arilysis r
i af ter it was allowec' ta pass thrcu;'h the icdir.e solution in the
,1 babble column, and the apparent discribution coefficient, K, defined
('?
,d w below was evalrated during these steady state conditinns,
?-
K = [HIO + I ] vapor L..[ -
[HIO + HI + 123 soi.
The epparent DF = 1/K q Results for the three different pressure temperature combinations m ; are sb.,wn in Figures 7.2.6, 7.2.7, and 7.2.8. The app. rent distri-() h tioa coef ficient is seen to vary f eem 0.01% at pH = 7 to as low a; J, 0.004% for pH:10. The cpparent decontamination factor thus varied M from J0' to 2.5 i 104 -
- ]
I The sbrupt decline in K as the pH of the solution increases is rel.ited to the fraction of the unhydrnlyzed molecules of elemental iodine in solution. If B is the fraction of $ydrolyzed 1 2, then
!] (1-B) is the unhydrolyzed fraction. Th.: depe ider.ce of the unhyc ot;* zed l
o
.i 7-16 Edi
.W
NED0-25420 y.
m' j fraction ca t'.e pH is shown in Figure 7.2.9 for 1 c ncentrations -
2 bsiween 2.5 - 250 ppe. This figure shows that elemental iodine is more hydrated in basic solutions and tnerefore less would be expected p .l to trans(er into the vapor phase.
Li
'. This. relationship between the DF and pH of the pool water is censistent d
i with the hydrolysis of iodine as discussed in Section 5.0, i e.,
j elemental iodine tends to remain hydrolyzed in a basic solution t[
m resulting in a mi..imal release frota *.he pool.
$.4 .
In BWR core damage ar.cident scenarios the pH of the pool is expected
- a. to be approximately 10.5 because of the large quantities of Cs that
( would be released (as Cs ions, elemer tal Cs, Cs oxides, or CsI)
~
p from the fLe1. Under these conditions iodine in the pool (as I or as I p) would be hydrolyzed and would stay in the water even under
~
}) saturated conditions.
- 7. 3 Sue., ry .f suppression Pool scrubbino Tests
]
I
, A summary of the data found in the literriure on iodine and partic'-late scrubbing by a pool of water is presented in Table 7.3.1. Relevant data includes the experimental test cc.iditions, chemical species
- tested, carrier gas, pool volume, depth of release, orifice size, and transport rate. The magnitude of tne attenuation factor (DF) l w2 is a strong function of the chemical form of iodine, particle size t of particulates, presence of sleam, depth of release, and tiie transport rate.
o q 1he applicability of these data to BilR transport mechanisms is J addressed in Section. 3.0.
IW Y
l)
Ye a
j s 7-17 4
,a~*
k
. T)
!'J NED0-25420 TABLE 7.1.1
.a; r' '
DECONTAMINATION FACTORS FROM FISSION PRODUCT ENTRAINMEhT STUDIES i
3 J
'y PERCENT RELEASED 1
't INITIAL AMOUNT IN TO SUPPRESSION DECONTAMINATION ELEMENT PRESSURE VESSEL CHAMBER AIR SPACE FACTOR "Df"
?.
- p Kr 250 cc (std) 25 - 50 4-2
'l Xe 250 cc (std) < 45 2 6 X 10-5 6
. 1 2
1 lb. 2 X 10 7 X 10 0 5 hrs. 1 X 10 5 X 10 5 Nal 110 ga 2 X 10'4 ZnS (2pm) 2 X 10
-6 5 X 10 7
2 110 cm 4
1 1
- [
^l t !
'.1 l
,f CF is defined as the ratio of mass in wetwell air space to the mass released in the reactor vessel. This is an overall DF and includes the effect of plateout in the drywell.
'i 1
7-18 YY
- . . - - _ y - .m a s - m m_2-.._c a .m.a u..m.uri - m m us.m _- -_ -.__ r w n m TAhtE 7.1.2 IOD!kt DICONIAMINAi!0N FAC10R$ AGA!:;5T STEAM FLOW AND PROPDRTION OF AIR
!cdine D.F. = I in pond + 1 in chimney T In chiency Rig Eig Proportion lodine Xel33 Test steam air cf air release recovery Total less No, flow tiow. .t.%
mg 1 of TotJ1 large RIMAR(5 lb/sec. IL/src. release droplets 1(a) 8.8 0 0 3 -
> 34 > 35 High Xel33 count on charcoal y reduced sensitivity m y (b) -
c.33 100 177 192 8
s Total D.F., 1(a) + 4 G 2 8.) 2.4 21.6 8 94 li' 1(b): >?S us 3
R o
- 7. 7 7. 3 48.7 1 102 20 28 4(a) 6.0 0 0 5 -
(b) 0.33 100 321 349 Total D.F., 4(a) +
144 150 4(b): 99 5 6.2 1. 5 19.5 5 15 280 316 6 6.0 1. 3 55 5 -
20 52 7 6.0 7. 3 55 5 107 34 40 8(a) 2. 6 i) 0 4 -
139 (b) -
3.33 100 144 Total 0.F.
."1 229 8(a)+8(b): 85 Elution -
- 0. 7 10(, -
146 159 9 3.1 0.7 19 1 85 70 19 Elution 0.7 100 -
100 106 10 3. 0 7. 3 73 -
33 20 41
(,eg,7,3)
. e e
n
. e g ,m8.Qe o m u 0 1 2 9 5 0 h t2
)
t2
) ) )
in s1) )
t2 t2 t 2 3 s s s s m sd e1 e1 e1 e1 en1 7
- T T T T 1 T a
- 7 7 7 7 7 c,
S e e e e n e1 f.
K R
r e re r e r e o r e e 2 al al al al a .l R 4 pb pb pb pb it y p6b (
. P ea sa ma ma lul n e a
. I R
ul oT oT oT o .T C( C( C( C( Ee C3(
Y r Ih Cy c e n s a I" o t
s e t s
a Ih l ee re gl
.
- l rp 3 5 4 9 4 2 n aao 6 8 4 4 8 2 d tl r
= "
i o d 1 2 2
. " I T I"
v-I
=
l a
t 9 6 3 6 8 4 F. o 2 6 1 2 3 1
- 0 T 1 1 2
- e n
id o
I S
E I
C A
]
R A
y r e ef s voa V eo e 5 m R
[
N c5l e
r r e
9 -
1 8
9 A
8 8 e
. H w.
T O
T 3 S t
1 n
l A e 7 G es A naeg 0 2 6 7 f
t B
R idl oe m 1
4 1
A T i o l r r C A
I N n
- C o
- l r 2
l A itcai %. 1 7 N n t 7 9 8 4 0 0
. I pf w 6 1 1 2 3 0 M oo 1 1 A r I P
- N y O C
-r 8 l' .
c ,c 5
[
N we 3 7 7 7 0 3 s l iilgros / 7 1 0 1 1 7
- 0 Rafb 0 I 1
t gsws e .ce 6 5 3 2 ieo/ n e, Rtlb sfl 3 7 3 5 f
r o e dd v n na
_ o oy oe) t
_ l ptC n d isn s* e e ro lail- no s u, e s
a les d e
mt imf itC0 i*
e r
p
)
, l en I n0l t y b ro f i6 a ol r
- la r
ee rs c
o:
hs r( m e .o r
nn m
o i t e h pn ar V iow ce pt eu ge rf ei t a iht o ma If Dl 5(
n
_ i o
t . t so u eh 1 2 3 a l 5 T 1 1 1 l E 1 y4O
TTi!LE 7.1.5 DEC0tiTAt'IrlATIT4 TA_CTORS_ f0R JLFig,N,TA!._!0DifiE IN A CO C/RRIER GAS
- 2 11 - Level of water in the tank Cm - Concentration of iodine ir. the 022 a t injec t. ion level Att Duration of the injection Ata - Duration of the first exhaust atove the water
- Ac - Activity measured in the water Aa - Activity measured te the air Of the first exhaust (first filter + cocoon film) k - Transfe factor for iodine through water, calculated after the first exhcust, frcm:
Aa 3
' Ae + AI DF =g
- 1) FLOW RATE: 20 g/s Exp. 11(m) Cm (pg/s) Ati (s) Ata (mn) Ae ("g) Aa (pg) k UF 1 1.7 2.7 10-7 73 (,0 3.5 10-4 4.8 10-0 1.4 10-2 + 75% 71 2 1.7 2.6 10-0 180 70 7.d 10-5 *4. 10-7 5.1 10-3 30% 196 3 1.7 1.1 10-7 35 60 6.7 10-5 *6.1 10-I 9.0 10-3 30% 111 y 4 1,7 4.7 10-2 45 25 11.9 4.8 4.0 10 -3k10% 250 ,
b 5 2.7 3. 10-7 90 10 3.7.10-4 8.5 10-7 2.3 10-3 60% 435 6 2.7 1.5 10-3 30 100 3.0 10 5.7 10-6 1.6 10 -2 35% 6? 5 h 7 2.7 1.5 10-2 75 2** 13.f 1.9 10-2 10-3 + 10% 1000 5 8 3.7 4.3 10-2 105 165 Cl 1.3 10 -I 2.2 10-3 + 10% 455 9 4.7 1.3Id'I 20 60 37.5 1.7 10 -2 4.5 40-4 30% 2222 10 5.7 1.4 10 -5 60 115 1.4 10-2 6.2 10 -5 4.4 10-3 + 25% 227 11 6.4 8.7 10-2 30 30 113 7 10-2 6.2 10-4 10% 1613
- 2) t. 9W RATE: 2 g/s 12 1.7 3.7 10 -1 Sb 111 19 7.7 10-2 2. 10-3 1 10% 500 13 3.7 5 10 100 15 106 6.1 10-2 5.0 10-4 + 15% 1724 14 5.7 7.5 10 -I 60 130 90 9.0 10-3 10-4 35% 10 4
The quantitles of todine deposited on the cocoon flim could not be determincd because the activities involved were small The trensfer factor is an uniere-timate because the duration of the first exhaust from above the water was very short.
t Added for this study.
Ref. 7.5 -
=
TA8tf 7.1.4 DICONI AMINAllCN F ACIORS FOR THE 3.06pm Ni/Cr ALR050L Aerosol D.f. s Aerosal in pend + Aerosol in CP.1,myy Aesosol In (Timney Rig Rig Propor t h a Test steam a6r of air Total less No. f!0w. flow. wt.% Iotal dropletl 10/sec. Ib/sec.
16 2.5 O.037 3. 7 171 241 17 2.5 0.375 13.1 120 241 18 2.5 1.8 42 381 > 751 r%
19 o
- 2. 5 7. 5 y
E
?$ 531 > 221 ?
F%3
% 20 2.5 7. 5 75 1681 > $41 E 21 9.0 1.8 16.7 15 49 22 90 7. 4 45 60 > 181 (Ref. 7.3)
NED0-25420 Table 7.1.6 (Ref. 7.6)
Iodine Removal Data for Various Blowdcwn Rates, Blowdawn Mixsu es, and Initial Iodine Concentrations in the Sucpression Pool Water Iodine Steam a
Air in injected Run Flow Rate Steam Into Water Air Sample Data d
No. (Ib/hr) (Wt%) (g) Iodine Removal pg/lb Steam DF 16 200 0 0.5 3.3 753 19 200 2 0.5 18.9 132 17 400 0 0.5 3.5 714 20 400 2 0.5 8.9 281 18 600 0 0.5 3.5 714 21 600 2 0.5 8.0 313 4 200 0 5.0 22.8 1096 7 200 2 5.0 126.2 198 3 400 0 5.0 17.3 1853 8 400 2 5.0 75.5 331 4
6 630 0 5.0 25.5 980 9 600 2 5.0 128.2 195 i
10 200 0 10.0 217.7 230 l 13 200 2 10.0 880 57 11 400 0 10.0 160.4 312 14 400 10.0 I
2 521.0 96 F 12 600 0 10.0 91.3 548 l 15 600 2 10.0 499.C 100
(
400 2 5.0 22.8 1096
- 22) b
- 23) 400 2 10.9 231.6 216
- 25) 600 2 5.0 333.6 75
- 24) c 600 2 10.0 748.0 67 i
Foctnotes: :
- a. Nor. ally, each run involved, the injection of 200 lb of steam
- b. Water initially contained 1000 ppm ha,Cr0
- c. Ocublesteamdowncenersof0.680inchesIfD.wereuseo
- d. Calculated for this stuay.
7-23
NE00-25420 l
TABLE 7.1.7 j Methyl Iodide Test Facility Parameters l Comparison of Measurement with Design Specifications 1/10,000 Prototype Desigr. Actual Pressure Vessel Total Volume, ft 3 1.58 1.6 Simulated Dreak Area, ft 2 0.00055 0.00063 Drywell Total Volume, ft 15.0 14.6 Wetwell Total Volurae, f t 3 21.6 19.8 2
Downcomer Area, ft 0.0289 0.0295 (Ref. 7.7) 1-24
i
, NED0-25420 l
Table 7.1.8 1
l N sorption of liethyl lodide in the 1/10,000 Scale Model Suppression Pool i TeseCec e.es %os Tat C. eg cH L S. f t R, sec T,
- F Absorpreon, %
1 2 D1 7 Center 4 0 90 68 6 2 0 20 77 , 4 25 90 62 5 3 0 72 7 4 35 120 523 4 2 07 72 4 54 12C 51.7 5 0 72 7 4 0 150 45 1 6 0 20 7.1 4 64 150 35.2 7 207 99 4 0 90 77.0 8 020 10 4 46 90 68 6 9 0 72 9.7 4 0 120 50 5 10 2 07 10 4 63 120 23.4 11 0 72 10 2 4 0 150 47 8 12 0 20 10 Cearer 4 5.7 150 22 9 13 0 72 72 $4e 4 6J 90 51 S 14 0 20 7 d' 4 0 116 59 3 15 2 07 7 4 32 150 55 6 16 0 72 to 4 63 90 57 0 1' 0 20 to 4 0 120 58 4
'8 2 01 10 Los 4 34 120 54 5 19 0 20 7 Center 1 0 90 49 8 2C 207 7 1 6 93 44 D 21 0 72 7 1 0 124 46 4 22 2 07 7 1 35 120 $2 6 23 0 20 74 1 35 150 48 2 23 0 72 74 1 6 150 402 25 0 20 to 1 0 90 43 6 2E !Ci 10 1 6 90 37 2 27 C 72 to 1 0 124 463 2E 2 C7 to 1 35 120 50 7 29 0 23 10 1 35 150 40 0 2 0 72 10 1 C e " *e 1 6 150 38 6 31 0 72 81 See 1 25 95 53 5 22 0 23 7 6 120 1
37 5 33 2 01 78 1 0 14C 10 8 T M 0 72 13 1 35 90 l SE 3
- 5 0 23 to t 1 6 120 35 5 36 2 07 10 2 SJe 1 0 152 22 9 37 207 81 Cev 2.$ 33 92 83 $
24 0 20 76 i 25 6 90 476 (
39 2 0e 81 25 0 120 62 1 43 0 '2 80 ?$ 6 1 20 2G $
41 0 23 79 25 0 153 40 0 42 ;
0 72 78 25 25 150 51 6 '
43 2 11 101 25 35 90 63 5 44 0 20 10 25 5 9] 23 1 45 2 37 10 2 25 0 123 62e 46 0 72 10 3 25 6 120 55 4 47 0 23 90 2
25 C 153 43 3 A s3 0 72 134 Ce, vee 25 35 150 55 5 ;
47 0F8 75 SJe 25 0 93 61 6 v' 0 02 67 25 35 123 43 8 i 51 21' 72 ?S ( 146 18 3 52 GC8 10 25 0 $3 61 2 s 53 022 98 '
25 35 123 28 7 9 211 t ri 5 oe 25 b 150 37 4
)
7-25 (Ref.7.7)
)
I ARI { 7. l.9 IN!!!AL CONDITIONS (OR fHL MAWIKf N 8t0WUOWN TEST 5 ET4n7551now v1 NET 5DTTIMs c6NIAINMENT COG IIIONS Olow- Type of positions of break orifice pressure man of depth of wet- volt.w of vent pipe teercrature of vent pSe Date of down be*ak diaeeter w4ter well pool wetwell pool submergence wetwell pool flow area per f ormance room pipe mm bar ton a m"* a 'C at Remarks I stes= 124 top
- 125 51 110 4.51 561 2.61,2.71,2.81 19 4.03 Orifice area Aug 13, 1972 reduced 40% z 2 stens 121 top 250 50 110 4.51 561 2.61,2.71,2.81 22 4.03 due to damays Sept I rn 3 steam 124 top 100 50 120 4.51 561 2.81 19 4 03 of measurement equhwnt Sept 21 8
4
@4 stese 124 top 100 49 121 4.60 572 2.90 44 4.03 Oct 20 v
a 5 ste.m 174 top 200 50 119 3.38 422 1.68 21 4.03 8
Oct 26 6 ste4m 124 top 200 50 119 4.99 618 3.29 18 4.03 Nov 2 7 water 122 feed water 150 50 117 4.50 560 2.80 17 4.03 No. 21 8 w4ter 172 feed wite.- ISO 50 119 4.4G $55 2.76 58 4.03 Orifice area Nov 30 reduced St ty l 9 steam 172 steam 310 50 105 4.51 561 2.81 17 4.03 the rupture Dec 19 I
disc
, 10 water 122 ste4e 330 51 293 4.50 560 2.80 17 4.03 Jan 23, 1973 water 122 feed water 150 Orifice area (330 m) re.
11 water 122 steam 313 51 293 4.54 561 2.84 18 2.68 cace.t Iftt by Feb 9 water 127 feed w4ter 150 the rupture
! disc for 23 seconds 12 water 17? s t e.o 330 52 297 4.59 553 2.80 19 1.84 Orifice area Feb 20 water 1/2 feed witer 150 .( 330 m) re-duced 2/% by tte rupture disc
'7h* soccIa7 pape conoc< tWlo the top-cupola of tt.e pressure vessel is ref erred to as the top pipe.
The wohne of the wetw*ll : out se(lwies the water standiwg in Lte upen vent pipe, but not in Lt-e Llocked went pipes.
(Ref. 7.8)
L % v ,u c:;1 w x<,.: a u :. c,rw .n . n 'na '.'W -' u n ' a 6 - t A L'nM MBmEmuMMM =^L- 3 M'" " A "
IARll 1.1.9 (Lont'd)
INiilAl COND1110 % (OR THE MARVIK[N Bt000WN !!il5 BEKI C;6"ITIGW STTsti cotcIT10NS va AIktimi whoI610%
Blow- lyre of posittois of break orifice pressure mass of depth of wet- valuee of vent pipe temperature of vent pipe down break diamet<r water well pool Date of hu. wetwell pool submergence wetwell pool flow area performance rcos pipe re bar ton a m"* m 'C at Remarks 13 water 124 top 90 48 308 4.49 551 2.19 11 1.84 9 90 only Mar 8,1913 water 122 steam 330 leakage 5 330-a ca re- K y water 122 fe-d water 150 duced 32%
y 9 150 only N leabaqe 14 water 122 steam 330 48 y
water 296 4.52 555 2.82 46 1.84 a
122 feed water 150 Mar 22 g 15 water 124 top 90 50 4.58 563 2. B8 18 1.e4 9 90 opened P.ar 29 wa'er 122 steas 330 281 and closed 8 330 never water 122 feed water 150 opened 16 water 124 tcp 90 3.34 412 1,64 14 1.84 8 90 only Apr 5 water 122 steam 330 50 321 leakage water 122 feed ater 150
- fhe special pipe connected in the top-cuenta of the pressure vessel is referred to as the top pipe.
The volume nf t v wetwell water pool includes the water st uding in the ope 1 vent pipe, but not in the blocked went pipes.
(Ref 7.8)
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7-30 ,
NE00-25420 j TABLE 7.1.11 J
' ,I ] 5patia' distribution of methyl lodide in the contair. ment at
(< various times relative to the t,1cwdown start. Methyl iodide was injected prior *.o the binwdown. Run No. 4.
Volume ccncentration of CH 3 I-131 (pCi/1)
Start of In wetwell In lower drywell In room 124
] sampling secs a b a b c a b
,7,$ -360 128 84 2047 1896 1952 2261 2298 e <
(" 540 853 865 78 37 0 0 --
r['j 1500 692 611 417 375 303 63 45 2760 532 532 541 4 72 423 193 179 i .'l IA .y H
-< The total and non volatile iodine contents of tha I,ool water samles.
d Run No. 4.
. Pool iodine contents Start of Sample Off gi.s Sample r- sam 5'.ing Sae.oling volume activity activity
'q ;. secs point (1) (pCi) (pCi/1) Total Mon volatile 1 540 A 5 5570 1500 2714 720 8 6 5520 1840 2760 1130 q
~. c 7' A - -
1200 -
3 0 - -
1400 - -
l
'A f
A: At the center of the pool.
.]
l B: At the pcr.phery of the pool.
I. .ia j .Ref. 7.8)
, f, 7-31 w-
,7 NEDO-25420 i
ii td TABLE 7.1.12
'L' The atmospheric distributic.i of I-131 in the containment at various times reistive to the start of t.ie blowdown. Run No. 8.
Volume concentration of CH3 1-131 (pCi/1)
G Start of In wetwell In lower drywell In rcom 124
'fl' sa vling secs a b a b c a b
-240 67 o7 1550 2470 2150 516 470 L ;. 480 1130 1230 22 37 38 168 119 l 3600 701 --
446 393 379 275 268 7200 554 605 366 344 312 296 270 10800 485 460 378 390 352 270 269
.g
..q <
.1 a
.li .'
The total and non-volatile I '.31 concentratiens in the pool water f/ I semples. Run No. 8.
i-
,, Pool iodine cont?nts 4 (pCi/1)
Start of Sample Off gas Sample s arapli,3 Sempling volume activit" activity j secs point (1) (pCi) (pCi/1) Total Non-volatile I
j 480 8 5.76 3730 1040 17 0 247 3600 A 4.25 2330 1140 1688 528 q
j 7200 A 3.93 1410 1140 1610 10S00 B 3.20 124] 1170 1558
.yl ,
b~ ~
A: At the centar of the pool.
6: At +he periphery of the pool.
[ (Ref. 7 8)
- . 7-32
. I NED0-25420 i
TABLE 7.1.13 ne spatial distribution of elemental iodine in the containment.
. Times ere refe. red to the start of the blowdown. Run No. 15.
j Volume concentration of I2 (pCi/1)
Start of In wetwell in lower drywell in room 124 sampling a secs a b a b c a b
.. E 3360 <6 3 94 93 9 <3 1
] 7080 2 1 287 144 6 1 1 9660 49
<4 2 71 10 2 1 q
i; 12420 <4 2 30 20 8 <2 <1 16020 4 2 46 44 14 <2 1
.i 1
1 The iodine contents of the sco) water samples. fiscs are referred
.i to the stTrt of the blowconn. Run No. 15.
Start of Sat.ple Off-gas Sample
[.l sampling Sampling vol uT.e activity activity Pool iodine contents sect poit.t ( .) (pCi) (pCi/1) (pCi/1)
'f,. ,]
? 7080 A 5.2 </S 4920 4920
- l Wa 9660 B 3.3 (18 3200 3200 1
~'
12420 B 4.1 <29 4370 4370 2
1 16C20 A 3.4 24 8553 8557 n.
I; F-i 1 A: A; the center of the pool, 7) h.
8: At the peri;hery of the pon1.
J 1
1 5.
(Ref. 7.8)
,5 1-33 .
d
,@ NE00-25420 O
F TABLE 's.l.14 The spatial distributiu. of elemental iodine in the containment.
I.I f,
1 Start of blowdown at 216 sets. Hun No. 16.
s.
.l Volun.e concentration of 12 (pCi/1)
Start of In wetwell In lower drywell In room 124 sampling
.. sr:s a b a b c a b 2100 <S 8 38 97 33 5 1 V 3600 <6 3 425 3810 79 7 2 1 7200 <5 3 27 74 19 <a 2 c, 9000 <4 2 tOS 142 26 6 3 h
10800 <6 3 22 79 23 5 10
'I di
, The total and non volatile iodine contents of the pool water sampics.
I Start of blowdown at 216 secs. Rui No. 16.
2 Pool iodine contents l
(pCi/1)
!? Start cf Sample Off gas Sample
!. sa pling Samp?ing volv e activity activity secs paint (1) (pC1) (pCi/1) Tots) Non volatile 1
3600 A 4.73 <28 1930 1930 y
b 7200 B 5.22 s24 7280 '280
/
9000 2 5.04 68') S150 8153 8150 10800 A 3.7 <32 7620 7620 b,
li '
A: At the center of the pool.
' B: At the periphery of the pool.
- 0) St.mple was we.ted by water.
{'
- (Ref. 7.8) 3 7-34 b .
NED0-25420 TABLE 7.1.15 DECONTAMINATIC.i FACTORS ,'0F) 0F METHYL IODIDE BETVEE*1 GAS AND WATER
- CURING THE BLOVD0%'N PERIOD e
BJ1owdownNo. DF Experimental 0F Equilibrig Wa'er Temperature *C 3 1.2 0.53 73 4 2.1 1 57 5 4.9 1. 5 46
- transport through pool 1
l t
t l
~
l l
1 7-35
NEDO-?5CO Table 7.1.16 DFs For Elerental locine During vost Blowdoan PerioG**
Blowdewn No. DF Experimental 15 225 16 265
- DFs due to natural removal processess (not by pool) i ,
1 l
ll I
I 1
, 7-36 l .
l j -
1 1
T ABL E 1.1.17 fl5510N N00tsri s p jvAt gy 5tAwaitg Ac tivity Distribution Outside
~ Of 10.000 cat. Tant Percent Removal By lattial Activity In Water Sea Water Calculated Inittet Arrountable - t Ac't iv ity[In Escafn. XT See Escape free Charge IIs tnN Gal. Eg7 AdjT(7,} wt * (B) Sag (C) B Run Autoclave Total I se tep* fec) Tv4 1pg (p_c] (pc)8 (pc): E.M g N 24 50 97.5
- 9 /. 5 0.112 i
0.215 n1' n1 5 -
1-131 6 91.4 -
'JI . 4 0.015 0.105 no ad - -
i-99 36 H6 9 -
86.9 24 4.5 nd nd - -
2 Rb-86 50 91.0 1. 7 99 7 0.317 0.285 0. .*51 <0.005 >98 >$1 1-131 20 'J I . / 1. 0 92.2 0. le-8 0.691 0. 6(20 0.094 86 C6 Y 's0 30 11.0 10.4 81.4 3.300 2.970 3.000 0.110 97 96 e f ft intels 4.385 3.946 3.851 0.204 95 95 f
N
?
w Y
-a N
- 1. O Normalised for air volta
- and laotope charge. Run 2 u ed as the basis for normall'ation.
- 2. Activity correctec to time out of reactoe. Actual activ;ty of Na-24, 1 131 and Y-90 measured after approulmately 4, 1 and 3 balf-lises cecay, respe(ttvelv f.i Pun 1. Actual activity of Rb-c6,1 131 and Y-9'l measured ef ter e proatmately 0.2,1 and I half-life respectively in Run 2.
- 3. Niet f! ejected. Minimum detectable limits 10- the sawales analysed are: I a 10 *I pC for Na-24; I a 10 f or I-131; Is 10
- f or V '30 and 5 x 10 8 pc for Rb-e6. *he sea i. ate. could ne' be cor.(entrated because of the high salt
( ntent, hence the large e tnisus detectable. limit val 6*s.
- 4. Not analyted.
- 5. N;t applicable. Activity escaping through se.s water (cluen t,* low alniaeva detectable limit.
(Ref. 7.9' i
! __m - _
I
NE00-25420 l Table 7.1.18 Decontarinaticn Factors for Container Venting Ur. der Sea Water Run Species Platecut Dr(D) Water Sr. rubbing DF 1 Na 2x10 5 .(c) 4 1 6x10 -
2 3
Y0 23 8x10 .
5 2 RbC) 2x10 50 4
1 3x10 6 7
4 Y0 23 1x10 27 OveraII DFsID) = Platecut DF x Water Scrubbing DF 7
RbC) 1x10 5
I, 1.8x10 5
Y0 23 2.7x10 (a) concentrations belcw detectable limit l
(b) in these tests the plateout rnechanisims occurred before the water scrubbing process I
i l
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I Table 7.3.1 .
- SU!'".ARY ()F pr,OL SCRCO!?;G TESTS 1 _
ILST COT 4DIT10iis '
f OlEftICAL FORf1 Or I j TEST TEST DESCRIPTI0ft SIftulATED FISSIO!1 DF
. POOL VOLUttE DLPTil 0F RELLASE r)RI FICE SIZE r A!:d!ER TO P0OL MA N T RAM J REF E X PERIMlliTAL [9000Cf5 pl (PAP.TICil SIZE) Tff!? F 7.2 51:rulated LOCA Xe, Kr 2-4(b) Fuci Vol 1000 gal Flashirig tlater Rapid Blowdown 010wdm.n to Dry- I (a) 105 - 6 Depth 18" 1" Air-Stecm
's well T. Pool fl$1 5x10g0 i'll 7 !!!xture 2 Steam & Sat Water ?n57 (2p m) 5x107 Temp -
f, 0 100 p:1 9 /500"F (L,e j
g 7.9 Bl owtowtj to a I (a) 2x105 Pool Vol 300 nal 2" Flashing llater Rapid Blowdown g- 1300 ft empty Y.;03 3x105 Depth 10-20' Air-Steam drum connected to pil -
Mixture a tank of water PbCl 107 Temp -
7.7 Simulated LOCA C11 1 3
1.1-3.2 Pool Vol 150 gal 2" Flashing flater Rapid Blowdown
] 1/10,000 scale Depth 1-4' Air-Steam
] eodel o f B'IR I4K I- pil 7-10 itixture g temp 90-150 h 7.3 Stean-Air mixtures 1 14-320 ; Pcol Vol 2000 gal - Steam-Air Steady 25-100 sec
- carried sirr.ulated 2 2' filxtures
. Depth Steam 3-9 lb/sec fi', sic,n produc ts pil 6.8 Air 0.1-7. Ib/sec through a water fli-Cr (0.06pm) 15-1680 Temp 10-60"C Air liti, 0-100 tank j 7.8 Simula ted LOCA blo.r-Cil)I 1.2-4.9 Pool Vol 105 gal 12" Flashing tiater Rapid Blowdown
,g j down to drywell and pool Depth pil 9'
Air-Steam Te:::p. 20-400C g 7.13 Steam pa , sed througt 1 10-750 Fool Vol - - Steam Steady 0.2-1.0 Kg/hr.
2
, a pac 1 of sat . water Lepth -
] containing I.,' 2.5-250 ppa pil 5.5-10 j Temp. 120-186*C j
j (a) Added to Peactor vessel watcr (b) Overall UF includes platcout and pool scrubb i 7-40 5_ . , , . . , , . . . . . . . .. ., . .
g labic /.3.1 (Cont.) *
I-
- I% Sutt"APY Of PCOL SCRUBBING TESTS i.?
N TE5T CON 01TIONS
%j TEST TEST DESCRIPTI0ri CllElllCAL FORT 1 l'00L LU ORIFICE CARRIER Of WUUsRD DT
( TRANSPORT RATE h REF E XPEHittitiTAL- pil SIZE TO POOL st CONDITIONS FISSIOri PRODUCTS TL f1P 'f V
$ 7.1 I2 (dl)10-500 Steam carrying [ la2)10-500 Pool Vol.3 gal 0.12" Steam-Air 0.1 lb/sec system (7 stenolited fission 11$ (b) 10-1000 150 gal 2.0" tilx tures 0.07 lb/sec2 air k' proihtc t s through Cll 1 3 (c) (d) 1.5-S Depth of 2" 4 lb/sec ft steam f] a witer tank Ni-Cr (0.06 f) m 50-100 release 20" 3 lb/sec-ft2 air p 3 11 -
f Temp 50*C
} 7,10 le carrying 17 12 (20 mg/ liter) 88-1500 Pool Vol. 30 gal 3/8-1" lic 5104 through a water g
peol lepth 8; 4-5 0.05 lb/sec.ft 2 g
- }
g,Il g
Temp 120*F A,
?
7 . f> Saturated Steam at 12 (0.5 - 10 ppn) 70-1100 Pool Vol 260 gal . 0.C8" T
- 7) Steam-Air Steady U
- 125 p.lg carried 1 2 Depth 4' flixtures 22-C6 lb/sec-f t 2
- j. through a root of a pil -
water fa.ap -
Air ut% 0-2 i 7, 5 CO, carrying 1 2 32 (15 ci max) 70-10 4
Pool Vol . - CO 2 0.02-0.04 lb/sec CO 2
} th?cugh a p.lol 10,000 gal.
4 of water Depth 6-20'
? CO 2 0 280 psi /400*C 1 11 -
[ Terap -
d
[
If (al) 0.6-40 pga air /stca:a utt 30-1
[] (a 2.'O.UI -0.4 ppn 1 2
alr/ team 5 Wt1 20-1 (c) (air / steam) Wt% 20-1
- ? (b) Air / steam lit 1 30-1 (d) (air /stea.n) Wt% 50-10 b -
U Q 7-41 N
N 5
) 5 M %aLi M b h M V h #Ja 2'
7 TABLE 7.3.l* (Cont.)
N SUI'dARY Of POOL SCRubB!tlG TESTS 5
id 9 IEST C0tIDIT10rir IEST TEST DESCRIPTIO!l CHElll C. L Pfi CARRIER POOL VOLUttE ORIFICE TRANSPORT RATE DF E1FERIMEriTAL U ^
L~
REF. SIZE E0!40!i10t:5 FISSI0ft PRODUCTS T ES. . u t s
J.8 3
P. 7.11 tiaf 0 in n, fla2 0 Pirticle'. ?O Pool Vol - - fi 2 0.5 ft N 2/"I" i*; carried tf. rough I pn Lepth 2'
[y water pti -
p Temp -
H j */.12 5tean carrying 1 2
2-200 Pool Vol 530 gal 0.57" Steam 0.1-0.4 Kg/sec W todine througt. a Depth 3-12'90-360 lb/sec f t 2 Q saturated pool (0.1-2 ppu) 1.H 0-12 (steady-high flew) 2 steam 0 (175 psig/
2 L& 320*C)
Temp 212'F
{
$ U d , R h- BUR OPEitATlNG EXPERIEriCE L.7 Sepa ra t t ori o f l', HIO 1000 Pool Vol - N/A fi/A ., . cam Condensing High flow-S*cady g iodine in UWR Depth - N/A >> 100 l b/s ec .
Q conden;cr IH - 7 b TEhP 100*F
'k' ressure 1-2 l, psia 4
1 6.9 BWR IIPCIS TEST l , HIO 7x10 ' Pool Vol - 0.5" Steam High Flow-Steady b Torus Scrabbing Selob gal 1300 lb/ scc h Test Cata Depth 4' h Steam 0 1000 psig I'"P ~
550*F
- h. ,
d i
i I ' l l 9
7-42 2
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' 96915 I POOL d '
r_-
%s M
9 Finure I.l.l. I r.sm. s en t Paessure Test F.icility (Ref. 1,2) f,Y p
u--- . . . . _ ._.
p'[ "
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$ ' SAMPLING POIPdi 5T[ AM r LOW IOUINE a i N'.JP4 4
CON TF.ot vs Lvf is st t T.O.4
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IPd'TI AL T Fr.9PF ftATURE . IU*C alt Rt, e$ r v';t P T fiUN '4 qvggy GOT RU*d 14 ACOUMUL ATO*45 DEPTH vF 84JEC780N 2 h figure 7.1.2 Scale .k,Jef cf the 5.3.H.W. Reactor Vented Steam Sunpression System (Ref. 7.3)
p
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7 -m 5 FIN AL CONCFNTR ATION OF 6
v 4
IODINE IN POND W4TER 10 -
LUTE DI AV g mos/i.tre t 10 6, Q u
$ 3mm o'
, 0 04 7016 (0 010 4 ret
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- AIR /STE AM tweet %I r
I f'igure 7.1.5. Removal of Clerental lodine Frem a Ster.m/ Air Mix *.ure by
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o NED0-25420 8.0 p plicability of Pool Scrubbing Test Data to BWR Transport Mechanisns Based on the experimental test data summarized la TaDie 7.3.1 and the expected BWR transport mechanisms and transport parameters listed in Table 4.2.1, assessments can le made to determine:
whether the experirental data is representative or non-represer.-
tative of BWR fission product transport conditions and whether the experimental DFs would be conservative or non-conservative when applied to EWR accident scenarios. The results of these assessments are presented in this section along with recommended pool scrubbing factors for each of tne dominant BWR transport mechanisms.
In reviewing the experimental data base, it is clear thit none of the tests exactly d'.' plicated the expected BWR transport conditions.
Several tests, however, are representative of the transport phenomena evpected during a postulated accident. For example, the removal processes for small particles and elemental iodine carried by steam / air mixtures through a pool of water are representative of removal processes which occur in a B4R suppression pool. The majority of the tests used volatile iodine (I ) and thus use of 2
these test results is clearly :onservative for cesium iodide in view of the solubility and volatility of CsI.
8.1 Subcooled Pools Of the BWR transport mechanisms identified in Table 4.2.1., none involve fission product release during rapid depressurization of the reactor in combination with flashing of wate" and subsequent blowdown to a subcooled suppression pool. Recall that fission product transport to the suppression pcal following degraded ccre conditions is expected to be a re k.ively slow and steady process during transients following the post-LOCA blowdewn p.eriod and during the core-concrete vaporization release. In view of these techanisms, the high flow test data reported in References 7.2, 7.7 and 7.9 are rion-representative of expected BhR fission product 8-1
NE00-25420 transport conditions. However,.even during these rather violent depressurization tests through a subcooled pool, the measured overall decontamination factors (plate out and pool scrubb;ng) were 0
extrerely large: 10 and 10 for 1 and 2 2 :.m ZnS, particles (Table
!. 7 1.1). While the blowdown tests are non-representative of expected l B'wR fission product transport conditions, they certainly indicate that in addition to other removal mechanisms t' a suppression pool can be an additional barrier in retaining fission products, even during rapid and energetic transport conditions.
The smaller scale tests conducted under " steady-state" flow condi-tions are more representative of the expected BWR transport condi-tions. Some of the tests reviewed, however, utilized rather shallow pool depths (2 feet) combined with high steam / air velocities.
These conditions undoubtedly led to pool scrubbing DFs which were low relative to what would be expected for BWR conditions' and this raust be considered when assessing the exoerimental tett data.
8.1.1 Retention of Particulates The retention of particu13tes by the suppression pool is estimated by two separ3te rethods. The first method is based on the experimental data using 0.06pm particles and elemental iodine. Since elemental iodine is expected to be less effectively removed than a particulate l (e.g. solids are scrubbed better than gases), usirg the elemental iodine data will provide a conservative estimate for the retention of particulates. The second m2thod is based on an analytical model which considers retention of particulates in rising bubbles. Both i
of these methods indicate that the retention factors for particulates i are expected to be greater than one hundred.
, Several tests (References 7.3 and 7.4) investigated the transport of particles carried by steam / air mixtures through a subcooled pcol l of water. DFs for 0.06 pm Ni-Cr particles in steam were approximately 100 or larger. Particulates released from the fuel in a steam environment may grow into large particles due to condensation 8-2 s
4
. l l
O N -wB NE00-25420 l
effects. WASH-1400 predicted particle sizes on the order cf 5-15 pm based on data from the Containment Systems Experiment (8.1) and .
recommended pool scrubbir.g DFs = 100 for particulates based on '
References 7.3 and 7.4. The actual DFs may be much larger under actual BWR conditions, i.e. 8-19 feet of water with small bubbles and complete condensation, larger particles, and potentially soluble fission produc.ts attached to particles. In view of these !
conditions, the available data can support a DF much greater than l one hundred (DF >> 100) for particulates in steam.
Particles can also be carried by a CO 2- steam mixture through a subcooled pool. For these mechanisms, complete bubble condensation may not occur and particle scrubbing will be determined by processes such as settling, absorption, and retention on the interior bubble surface. DFs for these conditions are more difficult to quantify I because of the lack of representative data. OFs for particulates,
{
however, can be estimated base,d on the measured removal factors for ,
elemental iodine. Since particulates would be removed more effec- d tivelu than a volatile gas, DFs for particulates would be much larger than DFs based on 1
- 2 1
)
H The data that does exist for estimating DFs for particles in a noncondensing gas are based on 0.06 pm Ni-Cr particles in air, 4 pm J Na2 0 in N2 gas, and volatile 12 in air, CO2and He gas. Refer- g ence 7.4 reported a DF of s 50 for 0.06 pm Ni-Cr particles in 30%
air and the data in Figure 7.1.6 can be extrapolated to yield a DF f'
of
- 20 for 100% air. A 0F of s 10 for 1 in 2 20% air was reported ,
and the data in Figure 7.1.5 can be extrapolated to yield a DF of b
/d
$2 for 1 2in 100% air. Both of these tests were based on a release g through 20" of water. In addition Reference 7.11 reported a DF= 20 El for 4 pm Na 2 0 particles in a Np carrier gas released under 2 feet y of water. Thus, DFs for particles are larger than those for volatile y 12 , by approx;mately an order of magnitude. This is in agreement
{
with what cne would intuitively expect. Reference 7.5 reported DFs Q.o of s250 for 12in CO 2released below 6 feet of water and 0Fs of i s2000 when the release was below 15 feet of water (refer to d 14 as 8-3 . 2
, , . . . . .. . ._ , m ,.. ..
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6.
A NED0-25420 l J
H Table 7.1.5). Reference 7.10 reported DFs >200 for 1 i" # h*1i""
2 carrier gas released at a nepth of 7 feet belo. water. Applying $
H the relative relationsnip between Ofs of particles and 1 in a ')
7 noncondensible carrier gas, the DFs for particles would be expected U to be greater than the values for 1 at ec;ual reiease depths.
7 The retention of particulates can also be estimated by examining j particle settling rates and the use of an analytica, model. The j decontamination factors for large particles in a noncondensible d carrier gas (> S pm) will largely depend on the particle settling n
velocity. The DFs for such particulates can be assessed, therefore, j by examining particle settling rates and comparing these retes to the expected bubble rise time through the suppression pool. Particle settling velocities range from 0.25 to 1.0 cm/sec for 5-10 pm g
particles respect.sely based on Stokes La.< calcul'ations and an assumed particle specific gravity of 4.0. This is realistic for
]2 fission product particulates sir.cc the spccific gravities a f cc s ia..; N 3
iodide, tellerium oxide, and zirconium oxide are 4.5, 5.7, and 5.5 Q
respectively. Based on these expected settlirg velocities, it $
would take approximately 0.5 to 2 seconds for 10 to 5 m particles to $
q settle onto the interior surface of a 1 cm bubble. In addition, J JJ turbulence within the bubble r.ay cause the particle to reach the 4 bubble surface even faster. Therefore, particles of this size 9 g
should have ample time to settle onto bubble surfaces and hence be retained by the water, since bubble rise times in the suppression 9 pool are expected to be on the order of 8-19 seconds depending on the accident scenario.
L
s An analytical nodel has been developed to predict the removal of hi particles from sir.gle bubbles (8.2). This model was used to estimate y pool decontamination facters in 1 cm bu::bles having rise times 3 between 8-19 seconds. Results frcM these calculations indicate [
that OFs much larger than 100 are expected for 5-10pm particles 7j having a specific gravity of 4.0. Tnus, based on this model particle g settling within the bubble shculd result in suastantial removal in o
! the pool. c; ;
8-4 Y d
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Thetefore, based on 1) r wervative application of experirrental scrubbing data, or 2) as...ytical estimates which account for the
((
l expected settling rates of particulates in a rising bubble, a CF of ;(
+
100 or greiter is expected for particulates carried through a
[]
subcooled pool by a noncondensing gas, and released at a depth of 8 r:
to 19 feet below the water surface. A s
Vapor release from the molten core / concrete interaction w uld i
condense out i'ito an aerosol as it moves to cooler regions away d t
from the limited area of interaction. These aerosols are expected j to be approximately 2 pm based on the tests at Sandia (7.1) and N should be non-volatile. The aerosol will be carried through the pool in a CO2 / steam / air mixture and released at a depth of S to 13 feet depending on the accident scenario. Due to the size and M cass of the fission product particulates, pool scrubbing is expected ..
,a to be moea effectiva t>1an the cbserved :: .tbing cf ulatilc I . Or }
4 .
the scrubbing of 0.06 pm Ni-Cr particles. Based on the data presented, o
~
a DF of at least 100 or greater is expected for a subcooled pool u daring core-concrete vaporization releases. 'i e
-l 8.1.2 Retentien of Cs1 "-
.3 Data on iodine scrubbing are limited to I ,p CH 1, HI, and HIO. .k 3
There are no reported data on the transport of Cs! in steam through
{'-]
a pool of water. However the behavior of Csl as a vapor or as a particle r~y be inferred from the test results on the behavisc of @) .
12 and soall particles. As discussed in Section 5.2, Csl is the ;:g expected themical form of iodine release from the fuel during a J, !
core damage accident. Cs! would be expected to condense or " plate-cut" h+
when it reiches cntal surfaces at temperatures at or below 400-500*C, and it would hydrolyze i.ito Cs* and I~ fons as soon as water or condensing steam is encountered (8.3). Cs! may also be attached to [}-
g ,_
aerosol particles an.i carried by steam to the supprescion pool.
[1]
For cases where the pool is subcooled, rapid bubble condensation e4 would occur and Cs! would rapidly icnize and remain in solution, 'U M
k 8-5 h
I
. iniiir M sO4 p r r.zmg aan wd-. ;6T;;,u rw um.j ,
I hED0-25420 I it :
1 sir.ce solutioni nf Cs and I~ are nontolati'e and staole. Experi- U mental data from Feferences 6.13. 6.14. and 7.4 snc.ved CFs of s f; 1000 for HIO in the concenser, DFs of s10 for HIO transporte_4 9 A
tr.roug5 4 feet of water, Ois of > 1h for HI in stcan through 20" (j of watcr. and DFs of >102 for 12 in stea:n through 20" of water.
DFs for Cs! in steam would be excetted to be ruch higher than 1000 [!
because Csl is much less volatile than HIO, HI, or 12 . Therefore, a DF >> 1C00 is expected for Cs! in steam relersed at a depth of 8 j to 19 feet below the surface of a subcooled poo!. i 4
h 8.2 Saturried Pools ;j 0
8.2.1 Retention of Cs! and Other Particulates p I
fcr saturated cools, the behavior of Csl in steam is expected to be sirailar to the behavior of particles carried by a non-condense 01e gas througn a subcooled pool. Particle settling and retention on a the bubble surface is expected to limit the ascunt of release from (d
the pool. This would be especially trua for Csl particles since (lj t-they would rapidly ionize upon contact with water. Reference 7.12 o investigated the effects of 1, scrubbing by a saturated pool. DFs ffv ranged bet een 2 and 200 ar.d were a functica cf21 con #"tr#tiO"' b pH of the water and depth of release (Figare 7.2.4). Extrapolating N, this data to a concentration of s 3 ppm and pH of slC (maximura l n
expected 84R conditions) a 0F of approximately 30 is obtained. g This certainly would be a loier bound fer volatile I under saturated 2
pool cceditions. The scrubbing of Csl particles should be much
@d higher in view of settling processes of particles in the bubble, pf; e,<
and ionization of Cs! in water. Indeed, experimental data shows j that particles are reJoved more efficiently than Volatile I . Cs!
2 f.f particles are similar to Nacl particles in that they are both i
,q
( hydroscopic and as such would grcw into larger particles in a hurnid 4 environnent and thus would have a greater chance of being removed.
l Figure 8.1 illustrates how a small Nacl particle can grow into [ >
\
larger droplets in the presence of moist air (8.4). Therefore, DFs ]
of at least 100 or greater for Cs! in stean are expected for saturated ]
pools whether the carrier gas is steam or a CO 2- steam mixture. y i e
,y NEDO-25420 t
- 1 8.3 S* m rv of Minimu? Supportable Pool Screbhing Factors ii J
'4
,a This assessment has identified the slainum suppressinn pool scruboing factors for each doininant SL*R transport mechanism for which tr.c f ry data can provide support. Results ara su::carized in Table 81.
/ 55 The actual 0Fs for se:utbing juring a postulated BWR accident are
- o. i.
, expected to bc such higher because the experimental test paranters
[ were conservative (i.e. shallow posi, small particles, and volatile Q I 2). oter.tially attainable DFs whicie are believed to be suppo-t-j% able by further testing a M alsc presented in Table 6.1.
h*
.'l a The application of t..ese scrubbirg factor, in probabilistic risk evalwations is presented in Section 9.
9
$b J >
6]?A
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m i
1 id k
h a
E, p
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i l
d 2
F 8-7 4 .
L1 smeiera m
Dw_.
~-
- w. . ~~TL, -_ -.: sjEL u >: ~ ih n - J _
I) .ABLE 8.1 HlklMUM SUPPORTABLE Ah3 POTENTIA 8_LY ATTAINABLE S!' PRESSIUN j P0OL OECONIAMINATION FACIO35 FOR IODINE AND PARTICULATES fj _ _
[ Transport Pathway Minimum Supportable Potentially Attainable h and As ociated Event (s) DFs OIS I3)
/, Subcooled Pool (7l Saturated Pool (2) -
g .
, eleactor pressure vessel 103 Csi, I',111 102 particulates (4) Ins-108 Csi, I . HI y to pool via safety rel.ef 102 particulates 30 1 2 103-108 particu.ates y valve and quencher 102 1 2 102 103 :,
E (Trarsients)
$ Reactor cressure vessel to 103 Csi, I', HI 102 particulates (4) 104-108 Cs!, I', Hi I
pool via vents (Transients 102 particulates 30 1 2 IG3 -108 particulates l following RPV depressurizatir.., 102 1 2 102-103 12 plt or LOCA post clowdern perled)
, Aerosol Transport to Pool 102 particulates 102 partict.eates (0) 103-108 particulates ,
c i' Via \'ents (Core-Conciete. 132 1 2 30 1 2 102-103 la
2 V3porization Release) 8 l4 8
(1) Durir.g tin:se conditions, complete condensation is expected when the pool is subcooled.
$ (2) A subcooled po'o1 is at a temperature below the saturation temperature corresponding to the pressure in tN 4 containment, while in a saturated pool steadj state boiling " steaming" is occurring.
J (3) Potentially attainable by further testing (satticated-subcooled pools).
(4) Includes Cs!
P
Mt,UU- 0420 i 100 2% ioc i%
1 802% i
- ',D tow)1oo tois5%)q ) /
-800%
. .w 1 1 1 H
io _
} ao e Cs s. . $
d j
~
i a h b ees Ok e9 7 -
/ l
-i I 5[
sc 6 - ( s '
W5 3E /
[]
I BB 5 -
1 5 .5 i d
r)
.i :t ~'
/ /'
} ' /.
se !
1 !! 3
/ i
, 5e / -
,s a i so a : , so e
g -l s/ essi s .
a ammen n f i , .
i I j g
! c oot o oi o i to j l paAvgTen or cuar = PARTICLE fe'
, j Figure S.I. The Ratio of the Diameter of Nacl Solutaan Droplet to the '
q Dia ster of toe Nacl Particle Tron klich the Droplet Has Been i
I Formed at Varic.us Levels of Relative Humidities (Ref. 8.I.) !
s a
i .
i I, I i i e.
! l j
i l ,
t l
I l I
I J
i i
i 1
, i i
i I
8-9/8-10 t
I
.~
w u w a m m e.,4 ,< . .
n.. v m , .. , s . y , + a .v . ..i>~ .. ,,
NtDU l*b4TJ '
(
- 9. 0 .
APPLICAT!ON OF POC' SCRUBBING FACTORS IN PROBASILISTIC ASSESSMEN_T d
In this section the attenuation factors recommended in section 8.0 are applied to the accident scenarios used in BWR Probabilistic j Risk Assessments. The accident scenarios are classified in terms
} of the initiating event and the presence or absence of ECCS flow.
Init*ati*g events include antuipeted transients, antisipated transients without scram (ATVS) and loss of coolant accidents (LOCA). These events can be rr. presented by four dominant core damage classes which are defined as follows:
[
, Class 1: A transient, a small bre k. or an intermediate break LOCA E
coupled with insufficient makeup water leads to core 4 damage.
/
Class 2: Following a transient or any size LOCA, the core is ,
r' covered with sufficient makeup water. However containmer't d
heat rewval capability is lost, resulting in primary centainment cracking due to overpressure and a possible
( subsequent loss of makeup water 1and cere damage.
'i .
Cla'.s 3: A large break LOCA or an ATWS without sufficient makeup water leads to core damage.
r i
Class 4: Following an ATVS event the reactor does not faimediately e become subcritical. However, the core remains covered with sufficient makeup water. Contis.uous bicwdown to the suppression pool results in boiling of the suppression i pool, cortainment pressurization, and primary containment cracking, which for low probability events may also lead to loss of makeup watert and core daaiage.
1 1 1
l Loss of makeup water is due to loss of Emergency Core Cooling I
System NP5H and not the loss of the suppression pool water.
1 1
9-1 l
i
-) NED0-25420 g
Classes 1 & 3 prstulate that a core damage occurs before primary containment crackfr..; and involves a subcooled pool. Classes 2 & 4 postulate primary containment cracking prior to core damage and
. involve a saturated pool.
Depending on the postulated initiating event, fission product transport to the suppression pool may occur by way of the SRV/qsencher lines, the horizontal vents, or a combination of both pathways.
L Minimum supportable DFs for each pathway are listed in Table 8.1.
i ' In Table 9.1 bFs are applied to each postulated class of accident.
l.
f It should be rs-emphasized that the values in Table 9.1 a.e judged to be minimum values supported by the available literature. It is expected that if experimental programs were conducted wisich include
- anticipated chemical forms of fis-fon products and representative flow rates, bubble sizes, pool depths, etc., appropriate to the hypothetical accident scenario, that the larger DFs would be proven to av.ist. Howevar, until such data are availabic the use of the attenuation factors in Table 9.1 in probabilistic risk assessments is recommended for a realistic estimate of the consequences expected E for such accidents.
I I
i 9-2 m g ,
i - -
a
1 NED0-25420 '
TABLE 9.1 APPLICATION OF POOL DFs FOR EACH POSTULATED ACCIDENT SCENARIO
+
4 i
i SCENARIO POOL CONDITION POOL OF !
Class 1 and 3 Subcooled 1000 CsI
{
100 other particulates l 100 1 7
Class 2 and 4 Satursted 100 Cal and ot5er i t
particulates t
30 I, '
l 3-3/9-4
l NED3-25420 l
10.0 REFE?ENCES l
i 5.1 Lin, C. C., " Chemical Effects of Gamma Radiation on Icdine in
, Aqueous Solutions," J. Inorg. Nucl. Chem., Volume 42, pg. 110.
(1980) 5.2 Ritzman, R. L. et. al. , " Release of Radioactivity in Reactor Accidents,"
Appeedix VI to " Reactor Safety Study," WASH 1400 (NUREG 75/014)
(1975).
5.3 Miller, A. O., " Radiation Source Terms and Shielding at THI-2."
Trans. Am. Nucl. Soc. 34, 633 (1980).
5.4 Postma, A. K. , Zavadoski, R. W. , " Review of Organic Iodide Formation under Accident Conditions in Water-Cooled Reactors," WASH-1233 (UC-80) (Oct 1972).
5.5 Heppolette, R. L. and Robertson, R. E., Proc. Royal Soc. 1959. A 252, 213.
5.6 Adachi, M. et. al., J. Chem. Eng. Japan 7, (5), 364 (1974).
5.7 Adamson, M. G., " Distribution of Fission Products," GEAP-14032-4 pp 1-31, Gen 2ral Electric Co. (1976) 5.8 Lorenz, R. A. et. al., " Fission Product Release From Highly Irradiated LWR Fuel," NUREG/CR-0722, ORNL (Feb. 1980).
5.8 Malinauskas, A. P. , " Iodine Release Frem Fuel," paper presented at the NSAC-14 Workshop on 3 dine Release From Reactor Accidents, EPRI, Palo Alto. CA. (1980).
5.9 Cubicciotti, O. and Sanecki, J. E. " Characterization of Deposits on Inside Surfaces of LWR Ciadding," Journ. Nucl. Materials, 78 pg. 96 (1978).
b.10 J. H. Davies, F. T. Frydenbo and M. G. Adasson, " Determination of the Chemical Activi.y of Fission Iodine in Zircaloy Clad UO2 Fuel Rods," J. Nac1. Mat. 80, 366 (1979).
5.11 J. C. C1:.yton and J. M. Riddle, "Some Studies of the Oxidation States of Fission Froduct 1 aine in Irradiated 002." Bettis (USAEC)
Report WAPO-TM-851 (1969).
5.12 R. Atabek, et. al., "Pelle' Cladding Interaction: Mechanical and Chemic.1 Approach to Modelling." IAEA Specialists Meeting on PCI for Water Reactors, RISO, Denmark, Septemoer 22-26, 1980.
10-1
- use
10.0 REFERENCES
(Cont'd) 5.13 R. A. Lorenz, et.al., " Fission Product Release from LVR Fuel Detected in Steam in the Temperature Range ECO to 1600 C." IAEA Specialists Meeting on the Behavior of Defected feel Rods, Chalk River, Canada, September 17-21, 1979.
5.14 Lutz, R. J. , Jr. , " Iodine Behavior Under Transient Conditions in the Pressurized Water Reactor," WCAP E637 (Nov. 1975).
5.15 Hull, A. P., "A Consideration of the Need for I.vacuation to Protect Public Safety at Three Mile Island, ANS. T;ans. Vol. 34, June 1980, p 91.
5.16 Pelletier. C.A., " Workshop on Iodine heleases in Reactor Accidents,"'
NSAC/14, Ncvember 1960.
5.17 NRC Heeting on Iodine Release From Accidents; and Estimates of Conseque.1ces of Nuclear Accident , November 18, 1980, Presented by D. O. Campbeil of ORNL.
6.1 " Report of the. President's Com.ssion on the Accident at Three Mile Island," John G. Kemeny, Chairman.
6.2 "Thr H"It5 Phy:ics Aspects of +ba St-1 Accident,' Foran, J. R. and Gamill, W. P. , Health Physics 9, 177 (1963) 6.3 " Analysis and Evaluation of Crystal River - Unit 3 Incident,"
INP0/NSAC Crystal River Task Force 1930, NSAC-3/INPO-I.
6.4 Miller, A.D. (NSAC) pc.'sonal phone conversation on 1-15-81.
6.5 Dunster, H.J., et. al., " District Surveys following the Windscale Incidant, Octeber 1957," PROC 2nd UN CONF on Peaceful Uses of Atomic Energy, Geneva, 18 rap te,r.b e r 1958, 296.
6.6 T.J. Tho: pson and J.G. Beckerley, The Technology of Nuclear Reactor Safety, The MIT Press (1965) Chapter II, Accidents and Destructive Tests.
6.7 Transactions of the American Nuclear Society 6 pp.137-141 '1"'3) 6.8 W.E. Kessler, et. al. , Transactions of the American Nuclear Society, 7 pg. 383 (1954) or 100-17D38 and 100-17194.
6.9 Lin, C. C., " Chemical Eehavior of Radioiodina in BWR Systems" J. INORG. NLICl . CHEM. Vol. 42, pp.1093-1099 (1980).
i 6.10 Ccoley, G.C., NUCL. ENERGY,22 153, 1978.
1 10-2 l
b
_..._. _ _ _ .- _. I
1 NFD0-25420
10.0 REFERENCES
(Cont'd) 6.11 Marrero, T.R., " Airborne Releases From BWRs for Environmental Impact Evaluations," Amendment 2 (Iodine-131) NEDO-21159-2, General Electric Compar.y, July 1977.
6.12 Potter, E.C. and Mann, G.M.W. , J. Br. CORR. L 26, 1965.
6.13 Pelletier, C. A. , " Surface Effects in the Transport of Airborne Radioiodine at Light Water Nuclear Power Plants." Science Appli-cations, Inc., Report prepared for the Electric Power Research Institute, Palo Alto, CA (1978).
6.14 Lin, C.C. , " Volatility of Iodine in Aqueous Solutions," to be E published in the J. INORG, NUCL. CHEM.
6.15 Pelletier, C. A. , Docket RM-50-2, " Effluents From Light Water-Cooled aiuclear Power Reactors"; AEC Staff Exhibit No. 22, Results of Independent Measurements of Radioactivity in Process Systems and Effluents at Boiling Water Reactors; USAEC, May 1973.
6.16 Palino, G.F., et. al., " Torus Radiological Measu.w ents During HPCIS Surveillance Test Monticello," March 1976, NEDM-12666. General Electric Company (April 1977).
6.'17'Montice110 Containment Data Sheet; General Electric document 22A5751, Rev. 1.
7.1 Berman, M. , Light Water Reactor Safety Research Program Quarterly Report, January-March 1980, NUREG/CR-1509/LOF4 SAND 60-1304/LOF4.
Berman, M., NUREG/CR-1177, SAND 79-2290, July-Sept. 1979, Vol. 13.
- 7. 2 " Fission Product Entrainment Evaluation Tests for the Pressure Suppression System," GEAP-3206 (1959), General Electric Company.
l l 7. 3 Hillary, J.J., et. al., "Icdine Removal by a Scale Model of the S.G.H.W. Reactor Vented Steam Suppression System," TRG Report 1256 (1966).
1 i
7.4 Diffey, H.R., et. al, " Iodine Cleanup in a Steam Suppression System,"
CONF-650407, 2 (1965) 776.
1 1
- 7. 5 Dadillion, J.; Geisse, G., " Diffusion of Iodine in Water The PIREE Experiment," CEA R-3199, AEC Lib. Trans 623 (April 1967).
l 7.6 " Energy Suppression and Fission Product Transport in Pressure Suppression Ponis," Stanford, L.E. ; Webster, C.C. , ORNL-TM-3448 April 1972.
10-3 l
4 NE DO-25420
10.0 REFERENCES
(Cont'd)
- 7. 7 Siegwarth, D.P., Siegler, M., " Scale Model Fission-Product Removal in Suppression Pools" GEAP-13172, April 1971, General Electric Company.
7.8 " Behavior of Iodine in the Containment During the Blowdown Runs, Discussion of Results" The Marviken Full Scale Containment Experiments,"
Sweden (1974) MXA-3-301; 201; 211; 101.
7.9 McGoff, M.J., Rodgers, S.J., " Sir.ulation of Container Venting under Seawater," Technical Report 59, Contract NOBS-65426, Mine Safety Appliances Co., Gallery, PA (December 1957).
, 7.10 Malinowski, D.D., et. al, "Radiologica' Consequences of a Fuel Hand!" g Accident," WCAP-7828, December 1971, Westinghouse Electric Corporation.
7.11 Hilliard, R. K., Phone Conversation (2/13/81) Hanford Engineering -
Development taboratory, Richland, Washington.
7.12 Devell, L. , (t. al, " Trapping of Iodine in Water Pools at 100*C, Containment and Siting of Nuclear Power Plants Proc. of Symposium IAEA 1967 (Vienna) C0hF 67042.
7.13 Styrikovich, M.A., et. al, " Transfer of Iodine Fron Aqueous Solutions to Saturated Vapor" translated from Atomnaya Energiya, Volune 17, No. 1, pp. 45-49, July 1964.
8.1 Hilliard, R. K. and Coleman, L. F., " Natural Transport Effects on Fission Product Behavior in the Containment Systems Experiment,
BNWL-1457 (Decenber 1970)
[ 8.2 Draft Report on Technical Bases for Estimating Fission Product Behaviar During LWR Accidents, NUREG-0772, 3/6/81, Appeadix E.
8.3 Stratton, W. R. et. al. letter to NRC Chairman John Ahearne on l
August 14, 1980.
l l
8.4 Cinkotai, F. F., "The Behavior of Sodium Chloride Particles in Moist Air," Department of Occupational Health, University of Manchester, England.
10-4
s
. ll NE00-25420 NEDO-25420 l i
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,t NUCLEAR ENERGY OtVISIONS e CENERAL ELECTRIC COM %NY SAN JOSE. CALIFORMA 95125 h
G E N E R A L ',k) E LE CTR I C a,.
TECHNICAL INFORMATION EXCHANGE Ih TsitE FAGE P' L9 bl AUTHCR SU3sEC T TIE NLMSER g i D.M.Kastler 730 g,,,
=
Oct. 30, 1981 7' TITLE 4 GE CLA $$
Suppression Peol Scrubbing I Facters for Postulated Boiling Water f-Reactor Accident Conditions GovEn ENT class 7 l I.j REPHOOUCISt.E COPY Fit ED A T TECHNICAL NUMSER OF PAGE$
SUPPORT SERVICES. R&UO. SAN JOSE.
CALIFORM A 95143 (Mae. Coce 211) 136 d I
$UMM Aft Y
[
The Mark I, II and III containtients are con- .I figured such that for most accident sequences fission f.
products released from the reactor vessel will enter f;-[4 the drywell ar.d be transported to the suppression pool d" where they will be absorbed or scrubbed.
. q-This report focuses on the scrubbing efficiency l,,
of the suppression pool for iodines and particulates.
.0
?
- n is a
fi 9
Y d
n lU x;
hT.D0-2 5420 'i i.
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.f By cu: ting out tNs rectande and folding in t eit, the ebove information can be fited F; into a standard c*d file. 1*
M
.'d ,
DoctNE NT NUMBER !iEDO-25420 )*
,)
INFOAM Ar10N PRE
- APE D F OR I' -
er Plant Desips and Analysis ,d MCTION I lq .
SUILDING AND ROOu NUM3ER 1900 311 M AIL CCCE 165 h
n 414 f6177) '
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