ML20099E309

Accident Source Terms for LIGHT-WATER Nuclear Power Plants. Draft Report for Comment

ML20099E309
Person / Time
Issue date:	06/30/1992
From:	Burson S, Ferrell C, Richard Lee, Ridgely J, Soffer L NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
To:
References
NUREG-1465, NUREG-1465-DRFT, NUREG-1465-DRFT-FC, NUDOCS 9208100123
Download: ML20099E309 (46)

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NUREG-1465

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Accicient Source Terms for Lign~:-Water Nue:Lear Power Plan:s prart neport for comment U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research L Soffer, S. B. Hurson, C. M. Ferrell,' R. Y. Lee, J N. Rkigely

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AVAILABILITY NOTICE Availability of Reference Materials Cited in NRC Publications

- Most documents cited in NPC publications will be available from one of the following sources:

1.

The NRC Pubhc Document Room 2120 L Street, NW., Lower Level, Washington, DC 20555 2.

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a NUREG-1465 1

JAccident Source Terms for Light-Water Nnclear Power Plants

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. Draft Report for Comment Manuscript Cornpleted: June 1992 -

Date Published: June 1992 L Soffer, S. B. Ilurson, C. M. Ferrell, R. Y. Lee, J. N. Ridgely Division of Safity Issue Resolution

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i Abstract in 1962 the U.S. Atomic Energy Commission pubished.

product releases has been developed based on significant

- '11D-14844, " Calculation of Distance Factors for Power severe accident research. This document utilizes this re-

' and Test Reactorf which specified a release of fission search by providing more realistic estimates _ of the

- products from the core to the reador containment in the

" source term" release into containment, in terms of tim-event of a postulated accident involving " substantial ing, nuclide types, quantities, and chemical form, given a meltdown of the core." This " source term," the basis for severe core. melt accident, This revised " source term" is the NRC's Regulatory Guides 1.3 and 1.4, has been used to be applied to the design of future Light Water Reactors to determine compliance with the NRC's reactor site (LWRs). Current LWR licensees may voluntarily pro-criteria,10 CFR Part 100, and to evaluate other impor-pose applications based u pon it.These will be reviewed by

- tant plant performance requirements. During the past 30 the NRC staff.

years substantial additional information_ on fission 4

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iii NUREG-1465 s

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l Contents Page Abstract iii Preface........

vii Foreword bc 1 Introduction and Background........

I 1.1 Regulatory Use of Source Terms..........

I 1.2 Research insights Since TID-14844..................................

2 2 Objectives and Scope....

5 2.1 General...

5 2.2 Accidents Considered....

5 2.3 Limi t at ion s,................................

5 3 Accident Source Terms 7

3.1 Accident Sequences Reviewed..

7 3.2 Onset of Fission Product Release..........,.

7 3.3 Duration of Release Phases..............

9 3.4 Fission Product Composition and Magnitude...

12 3.5 Chemical Form................

14 3.6 Proposed Accident Source Terms...............

14 3.7 Nonradioactive Aerosols 16 4 Margina and Uncertainties....

17 4.1 Accident Severity and Type....

17 4.2 Onset of Fission Product Release....

17 4.3 R el eas e P ha se D u ra t io n s.........................................

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4.4 Coaposition and Magnitude of Releases........

17 4.5 kxtine Chemical Form....

18 5 in-Containment Removal Mechanisms....

19 5.1 Containment Sprays..........

19 20 5.2 BWR Suppression Pobis 5.3 Filtration Systems..

21 5.4 Water Overlying Core Debris 22 5.5 Aerosot Deposition.........

22 6 References....

26 Appendix A: Uncertainty Dist-utions...

29 33 Appendix B: STCP Bounding Value Releases.

v NUREG-1465

m Contents (continued)

Page Tables 1.1 Release Phases of a Severe Accident..................................

2 3.1 - BWR Source Term Contributing Sequences................................................

7 3.2 PWR Source Term' Contributing Sequences............................................

8 3.3 ' Contribution of LOCAs to Core Damage Frequency (CDF}-Internal Events.....................

10 3.4 In. Vessel Release Duration for PWR Sequences............

11 3.5 In. Vessel Release Duration for BWR Sequenc:s.........................................

11 3.6 Release Phase Durations for PWRs and BWRs....................,.........................

12 3.7 STCP Radior aclide G roups.............................................................

12 3.8 Fraction of mean core damage frequency with high, intermediate, and low pressure sequences......

'13 3.9 Mean Values of Radionuclides into Containment for BWRs,Imw RCS Pressure, High 2irconium Oxidation......................................................

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3.10 Mean Values of Radionuclide Releases Into Containment for PWRs, Low RCS Pressure, High Zirconium Oxidation.......................................... 4

-14 3.11 BWR Releases Into Containment.....

15-3.12 PWR Releases Into Containment...............

15 3.13 Non-Fission Product Aerosol Releases into Containment...............

16 5.1 Distributions for Spray Decontamination of in. Vessel Releases....

20 5.2 ' Distributions for Spray Decontamination of Core-concrete Releases...

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^ 5.3 Distribution for Suppression Pool Decontamination Factors...............................

21 5.4 Median and Mean Decontamination Factors for Water Pools Overlying Core Debris 24 5,5 - Containment Atmosphere Aerosol Removal Rates...

25 5.6_ Containment Atmosphere Aerosol Removal Rates..

25 NUREG-1465 vi

Preface In 1962, the Atomic linet gy Commission issued Technical In the past 30 years, substantial information has been Infonnation Document (11D)l4844," Calculation of Dis-developed updating our knowledge about sesere LWR tance Factors for Power and Test Reactors." In this docu-accidents and the resulting behavior of the released fis-ment, a release of fission products from the core of a light-sian products.The purpose of this document is to provide water reactor (LWR)into the containment atmosphere a postulated fission pnxtuct source term released into

(" source term") was postulated for the purpose of calcu-containment that is based on current understanding of lating off-site doses m accordance with 10 CFR Part 100, LWR accidents and itssion pmduct behavior, ne infor-

" Reactor Site Criteria." De source term postulated was mation contained in this document is applicable to LWR considered to be representative of an accident that re-designs and is intended to form the basis for the develop-sulted in substantial meltdown of the core, and the fission ment of regulatory guidance, primarily for future LWRs.

pnxtucts assumed released into the containment were in addition, when final, this report will be made available based on an understanding at that time of fission product to existing reactor licensees and c4m serve as a basis tor behavior. In addition to site suitability, the regulatory possib!c changes to current requirements.110 wever, ac-applications of this source term affect the design of a wide ceptance of any proposed changes will be on a case-by-range of plant systems.

case basis.

t ii NU RiiG-1465 I

Foreword

'lhe information in this report will be considered by the date published in the Federal Register Notices. Com-U.S. Nuclear Regulatory Commission staffin the formu-lation of updated accident source terms for light water ments received after the duc date will be considered to reactors to replace those given in TID-14S44, calculation the extent practical. Comments should be sent to the of distance factors for power and test reactor sites. These Chief, Rules and Directives Review 13 ranch, Division of source terms are used in the heensmg of nuclear power 3:reedom of Irtformation and Publications Services, Mail plants to assure adequate protection of the public health Stop P-223, U.S. Nuclear Regulatory Commission, an sa ety.

Washington, DC 20555. Further technical information can be obtained from Mr. Leonard Soffer. Office of Nu-Any interested party may submit comments on this report clear Regulatory Research, Mail Stop NUS-324, U.S.

for consideration by the staff. To be certain of considera.

Nuclear Regulatory Commission. Washington, DC tion, comments on this report must be received by the dee 20555. Telephone (301) 492-3916.

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NURIiG -1465

1. INTRODUCTION AND IIACKGitOUND 1,1 llegulatory Use of Source Terms neered safuy fcatuns, w hich have iargely been optimized to remove elemental ioJme. Containment isolation valve The use of postulated accidental relca es of radioactive closure times have also been affected by these assump-materials is deeply embedded in the regulatory policy and tions.

practices of the U.S. Nuclear Regulatory Commission (NRC). l'or almosi 30 years, the NitC's reactor site crite-Use of the Tilh14M4 release has not been confined to an ria in 10 O '" N' '.00 (Rc f.1) have required, for ncens-evaluation of site suitability and plant mitigation leatures ing purposes, that an acc: dental fission pnsluct release such as sprays and filtration systems. The regulatmy ap.

resulting from "aunstantial racitdown" of the core into plications of this release are wide, including the basis for the containmi.nt be portulated to occur and thal its poten-(1) the post-a ident ra-Jiation environment for which tial radiological consequences be evaluated assu ming that safety related.iuipment should be qualified, (2) post-the containment remains intact but leaks at its maximum accident habitability requirements for the control room, allowable leak rate. RaJioactive material escaping from and (3) post-accident sampling systems and accessibility.

the containment is often icierred to as the "radiologica' release to the entiionment." The radiological release is in contrast to the TID-14S44 source term and contain-obtained from the containment Mak rate and a knowledge ment leakage release used for desien basis accidents, se-of the airberne radioactive i,- m "y in the containment vere accident releases to the envir'onment first arose in atmosphere. The radioacti,c ory within contain-probabilistic risk assessments (e.g., Reactor Safety Study, ment is referred to as the "insmitamment accident WASil-1400 (Ref. 5)) m examining accident sequences source term.

that involved core melt and containments that could fail.

Severe acciJent releases represent mechanistically deter-The expression "in. containment accident source term,"

mined "best estimate" releases to the environment, in-as used m this document, denotes the radioactive material cluding estimates of failures of containment integrity.

composition and magmtude, as well as the chemical and

'lhis is very different from the combination of the non-ph'ysical properties of the material within the contain-mechanistic release to containment postulated liy ment that are available for leakage from the reactor to the TID-14844 coupled with the assumption of very limited environment, The "in-containment accident source term" contamment leakage used for Part 100 siting calculations will normally be a function of time and willinvolve consid-for design basis accidents. The worst severe accident re-eration of fission products being icleased from the core leases resulting from containment failure or containment into the contamment as well as removal of fission prod-bypass can lead to consequences that are much prcater ucts by plant features intended to do so (e.g., spray sys-than those associated with a TID-14844 source term re-

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tems) or by natural removal pmeesses.

leased into containment where the containment is as-sumed to be leaking at its maximum leak rate for its design l'or currently licensed plants, the characteristics of the conditions. Indeed.some of the most severe releases arise fission product release from the core mM the contain.

from some containment bypass events, such ns rupture of ment (which must be distinguished from a release to the multiple steam generator tubes.

environment) are set forth in Regulatory Guides 13 and 1.4 (Refs. 23)and have been denved from the 1%2 re-Ahhough sescre accident source terms have not been port.TlD44844 (Ref.4).This release consists of 100% of used in individual plant licensing safety evaluations, they the core inventory of noble gases and 50% of the iodines have had significant regulatory apphcations. Sou rce terms (ha!f of which are assumed to deposit on mterior Wurfaces f rom severe accidents (beyond-design-basis accidents) wry upidly). These values were based largely on expen-

ame into regulatory consideration and usage shortly af-ments performed in the late 1950s involving heated irra-ter the issuance of W ASIl-1400 in 1975, and their apph-diated UO pel! cts. TID-14S44 also included IG of the cation was acederated after the Three Mile Island acci-2 remaining " solid" fission products, but it was dropped dent in March 1979. Cunent applications rely to a large fmm consideration in Regulatory Gu des 13 and 1.4.

extent on the results of W ASIl-1400 and include (1) part of the basis for the slees of emergency planning zones for Regulatog Guides 1J and 1.4 (Rets. 2 and 3)specify that all plants, (2) the basis for staft assessments of severe the source term withm contamment is assumed to be acciJent nsk in plant environmental impact statements, instantaneously available for release and that the iodme and (3) part of the basis for stalf priontuation and resolu-chemical form is assumed to be predominantly (91G)in tion of generic safety issues, umesobed safety issues, and elemental (19 form. with 50 assumcJ to be particulate other regulatory analyses. Source term assessments baseJ iodme and 49 assumed to be in organic form. These on W ASil-1400 methodolop appear m many probabilis-assumptions have significantly affected the design of engi-tie risk assev tent stuJies performed to date.

1 NURI G-1465

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1.2 Rcscarch Insights Since duratimt fission product releases. In addition, an exam.

ination and ass ssment f the chemical form of iodine TID 4 W likely to be found within containment as a result of a Source term estimates under severe accident conditions severe accident has also been carried out (Ref.18).

became of great interest shortly after the Three-Mile in contrast to the m.stantaneous rel cases (nat wer e postu-Island (FMi) accident when it was observed that only lated in Regulatory Guides 1.3 and 1 A, analyses of severe relatively small amounts of iodine were released to the accident sequences have shown that, despite differences environment compared with the amount predicted to be released in licensing calculations. This led a number of in plant design and accident sequence, such releases can be generally categorized m terms of phenomenological observers to claim that severe accident reicases were much lower than previously estimated.

phases associated with the degree of fuel melting and relocation, reactor pressure vessel integrity and, as apph.

ca, ana up n can'te below the reactor cavity by The NRC began a major research effort about 1981 to molten core materials, the general phases, or progres-obtain a be tcr understanding of fission-pmduct transport a are LWR accident are shown m l'able 1.1.

n, and release mechanisms in I WRs under severe accident conditions. This research effort has included extensive Table L1 Release Phases of a Sewre Accident NRC staff and contractor efforts mvolving a number of riational laboratories as well as nucicar industry groups.

These cooperative research activities resulted in the de.

RELEASE PilASES velopment and application of a group of computer codes known as the Source Term Code Package (STCp)(Ref. 6)

Coolant Activity Release to examine core-mcit progression and fission product re-Gap Activity Releast lease and transport in LWRs. The NRC staff has also sponsored significant review efforts by peer reviewers, liarly in-Vessel Release foreign partners in NRC research programs, industry 1.x-VesscI Release groups and the pencral public, The Sl'CP methodology I ate in-Vessel R, fase for severe accident source terms has also been reflected i; NUREG-il50(Ref. 7),which proWdesan updated risk assessment for five U.S. nuclear power plants.

Initially there is a release of coolant activity associated with a break or leak in the reactor coolant swtcm Assum-As a result of the NRC's research effort to obtain a better ng that the coolant loss cannot be accomtriodated by the understanding of fission pnxtuct transport and release reactor coolant makeup systems or the emergency core mechanisms in LWRs under seveic accident conditions, cooling systems, fuel cladding failure would occur'with a the STCP emerged as an integral tool for analysis of release o'f the activity located in the gap between the fuel fission product transport in the reactor coolant system pellet and the fuel cladding.

(RCS) and containment. The STCP models release from the fuel with CORSOR (Ref. 8)and fis.non product reten-As the accident prorcsses, fuel degradation begins, re-tion and tmnsport in the RCS with TR APMl!I P(Ref.9)-

sulting in a loss of fuct geometry accompanied by gradual Releases from core-concrete interactions are modeled melting and slumping of core materials to the bottom of using the VANIiSA and CORCON (Ref.10) codes. De-the reactor pressure vessel. During this period, the early pending upon the containment type, SPARC or ICliDF in-vessel release phase, virtually all the noble gases and (Refs 11,12) are used in conjunction with NAUA significant fractions of the volatile nuclides such asitx!ine, (Ref.13) to model the transport and retention of fission ecsium, and tellurium are released into containment.The product releases from the RCS and from core-concrete amounts of volatile nuclides released into containment interactions into the containment, with subsequent re-during the early in-vessel phase are strongly influenced by lease of fission pmducts to the envtronment consistent the residence time of the radioactive material within the with (Se state of the containment.

RCS during core degradation. High pressure sequences result in long residence times and significant retention improved modeling of severe accident phenomena, in-and platcout of volatile nuclides within the RCS, while ciudmg fission pnxtuct transport, has been provided by low pressure sequences result in relatively short resi-

. the recently developed NIELCOR (Ref.14) code. At this dence times and little retention within the RCS and con-time, however, an insufficient body of calculations is sequenth higher releases into containment.

available to provide detailed insights from this model.

If falure of the bottom head of the reactor pressure Using analyses based on the SI CP and NIEI COR codes vessel occurs. two additional release phases may occur.

and NUREG-1150. the NRC has sponsored studies Nlotten core debris re, cased from the reactor pressure (Refs.15-17) that analy7ed the timing, magnituJe, and vessel into the contamment will micract with the concrete N URl!G-145 2

l structural rnaterials of the cavity below the reactor (ex-Two other phenomena that affect the release of fision vessel release phase). As a result of these interactions, products into containment could also occur, as discussed quantitics of the less volatile nuchdes may be released in lieference 7. '! he first of these is referred to as "hich into containment, lix-vessel releases are influenced pressure melt ejection" tilPMii) If the llCS is at high somewhat by the 13pc of concrete in the reactor cavity.

pressure at the time of failure of the bottom head of the 1.imestone conerete decomposes to produce greater reactor pressure '.essel, quantities of mohen core maten-quantities of CO and CO gases than basaltic concrete.

als could be injected into the containment at hh;h veloci-2 These gases may, in turn, sparge some of the less volatile tW. In addition to a potentially rapid rise in con'tainment nuclides, such as barium and strontium, and small frac-tenmerature, a sig.'ificant amount of radioactive materi.d t;ons of the lanthanides into the containment atmo'

.:ouid also be added to the containment atmosphere, pri-sphere, l_arge quantities of non-radioactive acrowls may mail.,, the form of cerosok.The occurrence of 1IPME also be released as a result of core concrete interactions.

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s prtcludcJ at low 1(CS pressures. A second phenome-the presence of water in the reactor cavity overlying any core debris can sigmficantly reduce the ex vessel releases non that could alfect ti.e retease of fission products into (both radioactive and non radioactisc) into the contain_

mnta nment s a pode steam exploin as a rey q interactions between rno! ten core debris and water, t his ment, either by coohng the core debris, or at least by scrubbing the releases and retaining a large fraction m th'e could lead to fine fragmentation of some portion of ae water. The degree of scrubbing will depend, of course, molten core debris with an increase in the arnount 01, tipon the depth and temperature of any water overlying mrborne fission products. While small scale steam exnlo-the core debris. Simultaneously, and gencrally with a sions are considered quite likely to occur, they wdl not longer duration, late in vessel releases of some of the result in significant increases m the airborne activity -

volatile nuclides, w hich had deposited in the reactor cool-ready within containment. large scale steam explosions, ant system during the in-vessel phase, will also occur and on the other hand, could result in sigmficant iner cases in be teleased into containment.

airborne activity, but are much less likely to occur.

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2. OlUECTlWS AND SCOl'E 2j Gegjernl isel damare but without acactor vessel f ailure or core.

concrete interacuens (simdat in severny to the i Mi aco-

^1he ptimaoj ohettivc oi this rt part is to define a revised dent), or (h complete core-mcit es ents wah core-actident ioutte term for reputatory appbeatica for lutur e concrete interacthms 'lhese stenarios ate not equally 1 WRL The mient is to capture or inajor releiant m.

probable. Since many reactm systems must fad for corc sights available f rom recent sescrt accident rewatch on derradation with reattor sessel Imluie to occur and core-the phem.menolory of fiuion poduct release and trans.

umcrete interactions to occus one or rnos e systems may port behavior. Ihe levised source term ts espiested m be returned to an opciah!c status bclote co!c melt com-terms of tilnes and rates of appearante of taduu;tne mences. llence, moJern phmt designs, topcther with a fo.sion products into the containment, the types and Oporous program amied at duchipmp acudent manare.

quantitit of the species released, and other important tuent proctdures, are espccted to show that complete atinhotts suth as the themical forms of iodme. In aJJi.

cote-melt eve.ts are considerably leis hkel) to occur than tion,irr.portant removalinethanisms te p.. sprays natura' those invoh mg slight fuct damage.

deposition) that tenJ to reduce the hwon pmJucts m the u.ntainment almtn phere available for relece to the envi.

Since scu te accident sequences h:ac a large unccitmnty rontnent are diseuwed. l'his mechanistic appt v ih will regardmg the behartot ol molien fuel, hydrogen pencia.

therciore present, f or regulatory purposes, a moic f(ahs.

t:en, timing, cilectmnew of rcuwec, attions, anJ other DC portrayal of the afnount of liMiori prodult% presellt in phenomena nnd IMm es w hKa ma) mCW, no strMglW.

.he ontainment frotn a postulited severe accident.

for ewludmg comp'ete cote-mcits from consideration could be f mnd. In Om plants studied in NURiiG-ll5tl and for the 1 aSa" f ant about.10 to 90 percent of the 2.2 Aceittelits Corisidereti core dama, los.

s resulted in breachmc the reaaor pr emte vessel. Nnct t eactor pr eoute s cucl f ailur e may in mder to deime an accident source ter m for regulator) be a hkcly outcome of core danure sequences, although purposts, the staf f has exanuned a rance of severe acci tlus is qmte unteitem, the stafI rccommends that an ap.

dents that hase been analy/ed for IFR plants. besete pror iate accident soutte term f or regulatory use should accidents weie considered appropriate smee such acci-be one that is based oc a complete toie-melt, ne., one dents have been shown to dominate risk. I urther, Foot-involving remtor pressure vessellanu.e and subsequent note I to 10 Cl R Part 100 (Ref.1), in refernng to the core-toncrete mteraction. Therefore, the stall has se-postulated fioion product release to be used foi evalual-tected an accident souice te.m that is repicsentatne of ing sites, notes that "Suth accWents have pencially been the mean or average release hmunns awoeiated with a assumed' tesult in substant.almeltd wn of the core with ytoup of accident sequenccs typilymp a complete core subsequent telcase of apprecuble quantities of insion melt. The stafI also n,tends to allow cred!t for removal or pt oducts "

rednetion of fisdon products winun containment sia engi.

nected iratutes prosided for fission product reduction l! valuation of a tv.ge af severe retdent sequences was such as sprays or filters as well as by natural processes based upon work done in support of NURFG-1150 such as acrosol deposition. Thcx are dncoucJ in Sec-(Ref 7) and other nsk studies. This wock is documented tion 5.

in NURl!G/CR-5747 (Ref.17) and employ ed.he inte-grated Source Term Code PacLape (SICP) computer

""3 ggjg codes together with insights from the Miil COR code, which weic used to analy/e specific occident sequences ol The accident souice terms dehned m tius repon hac interest to provide the actident chronology as v.e!! as beco derived from cumination of a set of sescre accident detailed esumates of fission product behavior withm the sequences for 1 WRs of cuncnt design. llecause of gen.

reactor coolant 9 stem and the other pettment parts of cul Wimlanties in plant aad core design parameters, the plant. The sequences studied progreurd to a com-these results are also conddeied to be appheable to evo-plete core melt,imotving fadute of the reactor pressure lutionan 1.WR designs such as the Ashancsd lloihng vnsel and incle.Ng core concie'e mteractions as wclt Waer licactor ( AllWR) and Combustion Engmcenng's A key decision to oc rede in defming an accident source ter m is the ses enty of the accident or group ol accidents to Cunently, the NIC stalT n reucw mt s cactor dngns for be considered Possible choices range from (l)shp'il fuel desetal smal!ct I.WRc employing some pawn e leaturo damage accidents ins oh mg releases into containment of n for cos e coohnp and contamment heat r emovah WLdc the small flaction of the volatile nuchdes such as the noble "pa%ive" planw are eencrally sinnlar to p!csent I WRs.

gases. G) sescie core damage accideals mvoh mg ma or the) are espcc :d to bas e somewhat L nver cor e power i

5 NU RI (i-l 165

densities than thosc of cericos LWits.11cnce, an accident dation. llowever becausc of the lack of specific accident for the passive plants sirnk.r to those used in this study sv3 - nce mic.< nation for these designs, as well as the would hkely extend over a longer time span. I!or this general sirmtarity of the "pastne" plants to present reason, the tirmng and duration values provided in the 1-. Wits, the in containment acudent source terms pro-telease tables given in Section 3.3 are probably shorter vided below are considered generally apphcable to the j

than those applicable to the pmive plants. *lhe iclease

' passive" desagos as well.

fractions 14hown rnay also be overcitimated somewhat for high pressute sequences associated with the passive The accident source ter ms provided in this report ute not j

plants, since longer times for accident progiession would conswieted applicable to teactor designs that are very also allow for enhanced rete, tion of fauion prodxts in di": rent frorn 1. Wits, such as high. temperature gas-the primary coolant mtern dnting core beatop and degra-cooled reactors or liquid rncial reactors.

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3. ACCIDENT SOURCE TEllMS l

1

'Ihe expression "in<ontainment source terms "as used in Package (S'l C P) c2dcolations wcre performed. The domi-this report, denotes the finion product inventory present nant sequences wluch are considered to signdicandy im-in the wntainment atmospher e at any given time during a pact the source teun are summ:uved in Table 3.1 for sevene accident. To evaluate the in-containment source ll'.Vitt and Tuole 3 2 for PWik term dming the course of an accident, the time-history of the fission prm!uct release from tbc core into the contain.

The details of the specific accident sequences ate docu.

rnent must be known as well as the clicci of finion mented in NUltliO/CR-574i. "!! stimate of Itadet i

prmlurt ternosal mechanisms, both natural anc. er:pi.

clide Release Characteristics into Containment Under nected. to remow radioactive rnaternis from the contam.

Severe Accident Conthtions"(Rel.17).

j ment atrnosphercJlhis section discusses the tune.histry of the fission product releases into the containment. Re-M b elOfI;'lSSIM h'O M'l Nel m luoval mcchanisms are docussed m Section $.

This section discusses the anumptions r i in selecting the scenario appropriate f or defining the,arly phases of 3.1 ACCldell! SU(ltlellCCS HOVICWed the source term (coolant activity and ran releate riases).

It was considered appropriate to base these early telease i

All the accident sequences identified in NUltI:0-ll50 phases on th" design basis initiation that could lead to were reviewed and some additional Source Tertn Code cadicst fuellalnres.

Table 3.1 flWil Sous(c Team Cnntiihuting Sequemt.i I

l*lant

' Sequence Description Peach llottom TCf AlWS with reactor dept s rived TC2

' A'lWS with reactor prenurned l

TC3

'102 with wetweil venting Till Silo with lxit tery depletion I

Til2

't ill with conhinment fadure at venet failure S2131 LOCA (2"), no !!CCS and no ADS i

S2112 S21il with basaltic concrete V

RilR pipe failure outside containment TilUX Silo with loss of all DC power laSalle Til Silo with late containment failure

- Grand Gulf TC KlWS carly containment failure faus !!CCS Till Silo with hittery deplet:on T112 Till with 1I burn fails containment f

TilS

- Silo, no l!CCS but reactor deprenurited l

(.

lim TilS with AC recovery after vessel failure SilC Station tilackout iDCA low of Coolant Accident-l-

ItCP l(cactor Caohmt Pump HilR Residual lleat Removal ADS Automatic Depressurization System Al%S Anticipated Tiansient Without Scram l

1 7

NURIGl465

-n....ww.-.n,r

,,,,,-.,,,,,,n.,,,na,e,+--m

.~,-.+_.g,....w,-,w,,-

agenwr,-,,,,ne,

,,,m.w.,,

__,,-.-i-r-

--,-,--,.-,,,,,,ye m-,-..w--

w-,r 3,

,m,.

-y,y

Tahic 3.2 PWit Source Term Contributing Sequentes Plant Sequence Description Surry AO llX'A (but leg), no containment heat removal systems TMIF LOC-ao PCS and no A1 WS V

Interiacing system LOCA S 311 S110 with itCP scal i OCA S2D-B Si1LOCA, no IICCS and 11, combustion S2D fi SillDCA with E' hole in containment Won S2DCit LOCA (2"), no IICCS no CSitS S2DCF)

LOCA ItCP seal, no IICCS, no containment sprays, no coolers-ll, burn or DCII fails containment l

S2DCl:2 S2DCF1 except late 11, or overpressure failure of containment l

TMLU Transient, no PCS, no 1 CCS, no AFWS-DCl! fails containment Oconce 3 TMI.ll' Silo, no active 11SF systems SIDCF LOCA (3"), no !!Sl? systems

.Sequoyah S311F1 LOCA ItCP, no liCCS, no CSitS with reactor cavity flooded

$3llF2 S311F1 with hot leg induced LOC AS

~

311F3

- S311F1 with dry reactor cavity S 311 LOCA (M'? with Silo TilA Silo induces hot leg LOCA-hydrogen burn fails containment ACD LOCA (hot leg), no liCCS no CS S 3161

' Silo delayed 4 llCP seal failures, only steam driven AFW opcrates I

' S311F LOCA (IICP scal), no I!CCS, no CSitS S311 1.OCA (ItCP scal) no 13CC recirculation Silo Station tilackout LOCA Loss-of Coolant Accident

' ItCP lleactor Coolant Pump DCil Direct Containment licating

-PCS Power Conversion System I!SF I!ngineered Safety Feature CS Containment Spray

'CSRS CS liccirculation System ATWS: Anticipated Transient Without Scram -

N UltliG-1465' 8

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

S

'l l

A review of curreni plant final saMy analyrds reports A comparison c:deulation was done using theTit AC-PFI (USAlls) was made to identify all dedgn twsis accidents in MOD 3 code, version 14.3USQ.10 on the.W plantJihis which the licensee had identified fuel failure. I!or all analysis indicated that the first fuel rmt failure would accidents with the potential for release of radioactivity occur 34.9 seconds after pipe rupture, in contrast to the into the environment, the clan of accident that had the value of 24.6 seconds calculated using SCDAP/III! LAP.

shortest time unt!1 the first fuel tml failed was the design The reasons for ll.e dif ference between the SCDAP/

Imus 1.OCA As might be expected, the time until clad-Iti!1.AP5 MOD 3.0 and 'IR AC-PFI MOD i are dis-ding failure is very renetive to the design of the reactor, cuased in 1(eference 15.

the type of accident assumed, and the fuel rod design. In i

- particular, the maumum litur heat generation rate, the The review of the FSARs for ilWits indicates that fuel internal fuel rod pressure, and the stored energy in the failures rnay occur r:ignificantly later, on the order of fuel 1od are signifiennt considerations, several minutes or more. No calculations have been per.

formed using the aforementioned suite of codes.

To deter mine whether a design basis LOCA was a reason For determining the time of appearance of gap activity in able scenario upon which to base the timing of initial the containment (i.e., initial luct failure), whkh corte-fismon product clease into the containment, various sponds to the duration of the coolant activity phase and p

- PRAs wete reviewed to determine the contribution to the beginning of the pap activity phase,it would be appro-core damage frequency (CDF) resulting from LOCAs.

priate to perform a plant specific calculation using the This information is shown in 'l"able 3.3. As can be seen cales described above. Ilowen r, if rm plant specific cal-from this table, LOCAs are a small contnbutor to Cl)F culations are performed, the minimum times discussed for llWits, but can be a substantialcontributor for PWRs.

atiove may be used to provide au estimate of the earliest Therefore, for PWl(s a large LOCA is considered a rea, time to fuel rod failure.

Sonable imtiator.to assume for rmKieling the earliest ap.

pearance of the gap activity if the plant has not been 3,3 gg.dm M Helease Phsses approved for leak before break (Lilll) operation.1 or plantr, tLat have received 1.1111 epproval, a small i OCA Section 1.2 provided a qualitative discussion of Ibc re-(6"line btcak) would more appropnate:y model the tim *-

lease phases of an accident. 'lhis section provides esti-ing. For llWits, large LOCAs may not be an appropnate mated durations for these release phases.

scenario for gap activity timing. Ilowever, since the time

- to initial fuel rod failure is long for llWits, even for large The coolant activity phase herins with a postulated pipe l.OCAs, use of the larpe 1 OCA scenano should not un-rupture and ends when the first fuel md has been esti-duly penalize llWits and will maintain consistency with taaled to fail During this phase, the activity icleased to tl'c anumptions for the PWR, As with the PWR, for an the containment atmosphere is that associated with very 1.1111 approved plant, the timing associated with a small small amounts of mdioactivity dissolved in the coolant i OCA (6" line break) would be more approptiate.

itself. As discussed in Section 3.2 above, this phase is estimated to last about 25 seconds for Westinghouse PWits, and about 13 seconds for 11& W PWRs, assuming a in order to provide a tealistic estimate of the Mortest time large break LOC A. For a smaller LOCA (e.g., a 6-inch for fuel rmi failure for the l_OCA, calculations were per-Ime break), such as would be considered for a plant t'ial fonned utng the Fit APCON2, SCDAP/Rl!!.APS MOD has recebed Lilli approval, the coolant atthity phase 33),and FR APlh compulcr endes for two plants 'lhe two -

duration would be expected to be at least 10 minutes.

plants were a liabcock and Wilcox (likW) plant with a 15 Although not specifically evaluated at this ime, Combus-by b fuel iod arrty and a Westinghouse 4. loop (W) plant tion lingineering (Cli) PWRs would be expected to hme with a 17 by 17 fuel rod array. For each planh a sensitivity coolant activity durations similar to Westinghouse plants.

studywas performed toidentify the sire of the ! OCA that For llWits, the coolant activity phase would be expected resulted in the shortest fuel rod failure time (Ref,15). In to last longer; however, dess plant specific calculations both casesi the accident was a double <nded guillotme are made, the duratmos discussed above are considered supture of the cold leg pipeJihe minimum time from the applicable.

~

l-time of accident initiation until the first fuel red fails was l

calculated to be 13 and 24.6 seconds for the ll&W and K

'the gap activity rclease phase begins when fuct cladding L

plants, respecinely. A sensitivity study was perforrnN m r&re commences. 'lhis phase involves the release of L

determine the effect of tiipping or not trippio; the reae.

that radioactivity that has collected in the gap between tor coolant pumps 'Ih tesults indicated that trippmp of the fuel pellet and cladJmg. *lhis process releases to I

the icactor contmt pumps had no apprecable irupact on containment a few percent of the total inventory of the limingiFoc a 6-inch line btcak, the time until the first fuel more volatile radionuclides, particularly noble gases. io-tod fails is expected to be greater than b.5 and 10 minutes.

dine, and cesium. During this phase, the bulk of the fis.

respectively, sion products continue to be retained 'm the fuct itself.

9 NURl!G-1465

'lable.4.3 Conteihution of 1.OCAs to Core Damage Facquency (CDF)-Internallhents Pemnt of CDP Pertent of CDF caused by large I.GCAs lloiling Water th arton, caused by lA)CAs

( > 6" line break)

Peath llottom (NUltliG-1150) 3.5 1.0 Grand Gulf (NUlt!!G-1150) 0.1 0.03 Millstore 1 (Utdity) 23

'3 l'acssusired Water Itcattors Surry (NUltliG-1150) 15 4.3 Sequoy'.h (N UllIiG-1150) 63 4.6 Zion (NUl(FG-il 50) 87 1.4 Cniver t Cliils (Ilti!P) 21

<1 konce 3 (l! Pill /NSAC) 43 3.0

'lhe gap activity phase ends when the fuel pellet bulk for llWit plants than for PWit plants,'l his is laigely due tempciature has been raised sufficiently that significant m the lower core power density in ll~ Wit phnts that cv amounts of fission products can no longer be retained in tends the tune for complete core melt. Itepres atative the fuel. As noted in iteference 16, a leview of STCP times for the duration of the in vessel release phase have calculated tesults for six reference plants, PWits as well been selected to be 1.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> and 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />, for PWit and as llWits, indicated that significant fission product re-IlWit plants respectively, as tecommenced by lteference leases from the bulk of the fuel itself were estimated to 17, commence no entlier than about 30 ininutes and 60 min-utes for PWits and DWits, respectively, after the onset of the accident. On this basis, the duration of the gap activity release phase has been selected to be 0.5 and 1.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />, The ex vessel iclease phase begms when rnoiten core for PWits and llWits respectively.

debns exits the reactor pressure vessel and ends w hen the debns has cooled sufficiently that significant quantities of During the early in vessel release phase, the fuel as well fission pmducts are no longer being released, During this as other stiuttural materials in th; core reach sufficiently phase, significant quantities of the volatdc radionuclides high temperatures that the reactor core geometry is no not already released during the early inacssel phase as longer maintained and fuel and other matenals melt and well as lesser quantities of non volatde radionuelides are relocate to the bottom of the reactor pressure vessel.

released into containment, Ah igh releases from core-During this phase, significant quantities of the volatile concrete interactions are prediced to take place ovtr a nuclides in the core inventory as well as small fractions of number of hours after vessel breach, Iteference 16 indi-the less volatile nuclides are estimat ed to be released into cates that the bulk of the fission products (about 90%),

contamment,'lhis retcase phase ends when the bottom with the exception of tellurium and ruthenium, are ex.

head of the reactor ptessure vessel fails, nilowing molten pected to be released over a 2 hout period for PW"ts and core debris to fall onto the concrete belo,v the reactor a 3-hour period for IlWits For tellurium and ruthenium, pressure vessel. Itelease durations for this phace vary evvessel releases extend over 5 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, respectively, dependmg on both the scactor typt and the. eident for PWits and llWits. 'lhe difference in duration of the sequence. Tables 3.4 and 3,5, based on results from itef-evvessel phase Setween PWits and llWits is largely at-crence 16, show the estimated dumtion times for PWits tributable to the larger amount cf zirconium in llWits, and llWits, respectis ely.

which provides additional chemial energy of oxidation.

3 Ilased on lieference 17, the enve 'sel t elease phase dura-Ila.cd on the information in these tables, the staff con-tion is taken to be 2 and 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.. espectively, for PWits cludes that the in-vesreli clease phase is somewhat longer and llWits.

NUlti!G-1465 10

.~. -._-

1 Table 3.4 In-Yessel Release Duration for l'WR Sequences

.j Release Durallon Plant Attident Sequente*

(Min)

Surry TMLil' (11) 41 Surry S31!

(!!)

36 Surry AG (I.)

215 Suny V

(L) 104 Zion TMLU (If) 41 Zion S2DCit/S2DCF (11) 39 i

Seqturyah S3111t/S3tl

-(II) 46 Sequoyah S3111 (11) 75 Sequoyah TMI.ll' (11) 37 Sequoyah TilA (L) 195 Sequoyah ACD (L) 73 Oconce TMI.ll' -

(11) 35 Oconce SIDCit (L) 84'

• (11 or L) Denotes whtther the accident occurs at high or low pressure.

Talle 3.5 In-W5sel Helease Duration for llWit Sequences I'lant Accident Sequence

• llelease Duration (Min)

Peach llottorn TC2' (1i)

-66 Peach llowom TC3 68 Peach llottom TL1 (1.)

97

= l'ench 30ttom Till/nl2 (11) 91 I

- Peach llottom V

(L) 69 Peach llottom

- S21!.

(11) 81-l'each 1101 tom TilVX (11) 67 g

12Salle Til -

(II) 81-Grand Gulf '

Til (11) 122 Gmnd Gulf _

TCt (I.)

130

- Granti Gulf TilS/I'l!R (1,)

96

- (11 or L) denot:s whether the accident occurs at high or Io., pressure.

11 NURIIG-1465

'the late in vessel release phase cornmences at vencl Table 3.7 STCP Itadmnuclide Groups breach and proceeds simultaneously with the occurterice of the ex veuel phase, llowever, the duration is not the GitOUP 1;lfMENTS same for both phases. Durmg this release phase, some of the volatile ntielides delmsited within the reactor coolant 1

Xc Kr system earlier during core degradation and melting may re volatilire hnd be released into containment. Iteference 2

I, lir 17, after a review of the source term uncertainty method-3 Cs, Itb ology used in NUREG-1150 (Ref. 7), estimates this 4

Te,Sb,Se phase to have a duration of 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> 'Ihis value has been 5

Sr selected for this report, 6

Ru, Rh, Pd, Mo, 'l.e 7

1.a. Zr, Nd, !!u, Nb, Pm, Pr, Sm, Y 8

Ce, Pe, Np

- A summary of the telease phases and the selected dona-9 U"

tion times for PWRs and ilWRs is shown for reference purposes in Table 3.6.

Source term rcleases into the containment were evalu-Table 3.6 ated by reactor type, i.e., IlWR or PWR, from the se-Release Phase I) orations for PWRs and DWits quences in NUlt!!O-1150 and the supplemental ilCP i

calculations discussed in Section 3.1.

Duration, 1)uration, Releases into containment d> ring the early in vessel I WRs dWRs phase, prior to reactor prevure vessel failure, are mark-Release Phase -

(llours)

(llours) edly affected by retention in the RCS, which is a function of the residence time in the RCS during core degradation.

~'

~^ ~ ~ ~~

Coolant Actsvity 10 to 30 30 seconds, liigh pressure in the RCS during core degradation allows seconds

• for longer residence time of acrosols released from the core.This,in turn, permits increased retention of aero-Ws within the RCS and lower releases frorn the core into Gap Activny 0.5 1.0 the containme nt. Simitatly. low pressure requences cause acrosols generated within the RCS to be swept out rapidly Early in Vessel 1.3 1$

without significant retention within the RCS, thereby re-sulting in higher releasc fractions from the core into con.

tasment Ex Vessel 2

3

'the relative frequency of occurrence of high vs. low pres-litte In Vessel 10 10 -

sure sequences were examined for both HWRs and PWRs. The results of this survey are shown in Table 3.8,

'Without approval for leak-before break. Coolant ae.

and they indicate that a significant fraction of the se-tivity phase 6uration is assumed *.o be 10 minutes with quences examined in terms of frequency, occurred at low leak before-break approval.

pressure, in adoition, advanced PWR designs are increas-mgly inceporating safety-prade depressunzation sys-tems, primarily to minimlie shc likelihood of high pres-sure melt ejectian (llPME) with its associated high containment utnosphere heat hiads and large amounts of 3.4 FiSsmn Product Compos tion mRI atmospheric acrosols.

i Magnitude For these reasons, the composition and magnitude of the simrce term has been chosen to be representativc of con-In considering severe accidents in which the containment.

ditions associated with low pressure in the RCS at the might fail, WASil-100 (Ref. 5) examined the spectrum time of reactor wre degradation and pressure vessel fail-of fission products and grouped 54 radionuclides into 7 ure. Reference 17 pro, ides estimates of the snean core majo;;troups on the basis of similarity in chemical tshav-fractions released into containment, as estimated by ior The effort associated with the STCP further a naly7ed NU Rl!G-1150 (Ref. 7), for accident sequences occurring these groupings and exp;mded the 7 fission product under low RCS pressure and high rirconium oxidation groups into 91troups. These are shown in Table 3.7.

conditions. These are shown in Tables 3.9 and 3.10.

' NUREG-1465 12

Tubic 3.3 l'rattion of mean tore darnage fraquency with high,interrnediale, and low pressure sequtnces (internal egents only unless othen.ise noted) low I'sessure at flolling Water liigh Pressure at inter med, prest Vessel lirtath Iterclors Vessel liteach at vessel breach

(< 200 psi)

No sessel breath IaSalle external events only 0 27 N/A 0.67 0.06 1aSalle. internal 3ents only 0.19 N/A 0/,2 0.19 Grand Gulf 0.28 N/A 0.51 0.21 Peach llottorn 0.51 N/A 0.41 0.08 Pressurued Water liigh Pressure Interined. press.

low pressure at No vessel Reactors at ve:,scl breach at vessel breach vessel breath breach

(< 200 psi)

Surry

' d6 0.07 0.37 0.50 Scquoyah 0.14 0.21 0.24 0.41 Zion 0.03 0.1 $

0.72 0.10 Table 3.9 Meun Values of Itadionutildes into Containment for PWRs, law ItCS Pressure, liigh 7,irconium Oxidation NUCl,lDE Ear ly in. Vessel ikVeuel

1. ate inacssel N. O, 4.0 0

0 1

0.27 0.37 0.07 Cs 0.2 0.45 0.03 Te 0.11 0.38 0.01 Sr 0.03 0.24 0

lla 0.03 0.21 0

11u O OC7 0.004 0

1a 0.002 0.01 0

Ce 0.009 0.01 0

13 N UltliG-14$

t___-_-___

Table 3.10 Mean Values of Itadionutilde lleleases into Containment for PWits, l.ow itCS Pressure, liigh Zire mium Osid400ti NUCLIDl; f.aily in.Venei l'a Veshi Late Ina rssel N.G.

1.0 0

0 1

0.4 0.29 0.07 Cs 03 0.39 0.06 Te 0.15 0.29 0.025 Sr 0.03 0.12 0

Ha 0.04 0.1 O

llu 0.00R 0 004 0

Ia 0.002 0.015 0

Ce 0.01 0.02 0

3.5 Cheinical Fmcm he expected within containment w here there is no control of the pil.

A study has been completed recently on the chemical form of iodine and its subsequent behavior af ter entering All other fission products, except for the noble gases and containment from the reactor coolant system. NUltliG/

iodine, discussc3 above, are expected to be in particulate Cit-5732," lodine Chemicall'orms in I Wit Severe Acci.

fot m.

dents"(ltef.18) documents the sesults of this study.

The results moicate that iodine entering the containment 3J I'roliosed Accident Source Terms is at least 95% Csl with the remaining 5% as I plus lil, with not less tha n 1 % of each as I and iII. Once the iodme

'Ibe propesed accident source terms, including their tim-entets containment, however, additional icactions are ing as well as duration, are listed in Tables 3.11 for llWits likely to occur. In an aqueous environment, as 1 pected and 3.12 for PWits. It should also be noted that the rate of for 1. Wits, iodine is espected to dissolve in water pools or relcase of fission products into the containment is as-plate out on wet surfaces in ionic form as 1 Subsc.

sumed to be constant during the duration time shown.

quently, iodine behavior within containment depends on the time and pil of the water solutions. ilecause of the

'the information for these tables was derived from the presence of othei dissolved fission products, radiolysis iF simplification of the NUltliG-1150(Iter.7) source terrns expected to occur and lower the pil of the water pools, documented in NUltliO/ Cit-5747 (ltef.17).

Without any pll control, the results indicate that large fractions of the dissolved iodine will be converted to ele.

With regard to the ex vessel releases associated with mental k(hne and be teleased to the containment atmo-core-concrete interactions, according to iteference 17, sphere ( rganic iodine will thcu be produced slowly over there were only slight differences in the fission products time from the elemental iodine available. Otherwise, if retrased into containment between limestone

s. basaltic pli conttol is available and the pil is maintained at a value conctetc. Ilence, the tab!c shows the releases only for a of 7 or greater, very little (less than 1%) of the dissolved limestone concrete.17urther, the releases shown fo. the todine will be convertri to elemental iodine.

ex-vessel phase are assumed to be for a dry reactor cavity having no water overlying any coce debris. Where water is These resultsindicate that only foi cases in which the pl!

covering t he core debris, act osol scrubbing will take place is controlled.o levels of 7 or greater would the elemental and reduce the quant.ty of acrosols entering the contain-iodine inrentainment be a small fraction of the total.

ment atmospher e. See Section 5.4 for f ur ther information Significantly greater quantities of elemental iodine may in this regard.

NUltliG-1465 14

l Talile 3.11 ItWit Ittleases into Containment

• Gap Ittlease I:atly in.bsel EoVessel I. ate in&sul 1.5 3J) 10.0 1O Dur ation (f lours) 0.05 0.95 0

0 Noble Gases 0.22 0 37 0.07 0 05 faline 0.03 0.15 0.45 0.05 Cesium 0.01 0.11 0.38 0

Tellurium 0.03 6.24 0

0 Strofitium 0

0.03 0.21 0

liarium 0.007 J.004 0

0 Ituthenium 0.009 0.01 0

0 Coriurn 0 002 0.0 '

O 0

t imthariurn

' Values shown are fractions of cote inventory.

Table 3.12 l' Wit iteleases into Containment

• Early in Yesul I N Vessel Late in. Vessel Gap Itcicase 1.3 2.0 10.0 0.5 Duration (i fours) 0 0,95 0

0 03 Noble Gases 0 07 0.35 0.29 0.05 Iodme 0.25 0.39 0m 0.05 Cesium 0.025 0 15 0 29 0

Tellurium 0.03 0.12 0

0 Strontium 0.04 0.10 0

0 llatium 0.006 0.004 0

0 Ituthenium 0

0,01 0.02 0

Cerium 0.002 0.015 0

0 1 anthanum

' Values shown are fractions of core inventory.

NU) LEG-1465 15

3.7 Nonratlicactive Aerosols considered typicai, may have an cfrect upon plant equip-In addition to the huion product fractional releases into ment i.uch as finter hadmp, as discussed in Sectica 5.

containment show n in Tahics 3.11 and 3.t h large quanti-ties of nonraditw tive or relatively low activity aerosols Table 3.13 Nnn-l'isi, lor. Product Aerosol Ittleases will also be released into containment. 'these aerosols into Containtnent*

arise liom core structural and control rod materials re-leased dunng the in-vessel phase and from concrete de-H\\YR H)VR P\\VR P\\YR composition products during the ex vessel phase. A de-tailed analysis of the quantity of nonhssion product In Vevel fix Vessel in-Vessel I!x-Vessel actomb scleased into containtnent was not undertaken.

lloacsce, the qt antities released mto containment for

)

one PWit (Sequoyah) sequence and fot one llWR (Peach

--y.

ilottorn) sequcnce, evaluated in Reference 19, ate shown

~~

in Table 3.13. The acrosol masses shown in Table 3.13,

• Values shown are in kilogranis. Generation rate is assumed to be conuant over the release phase.

f

\\

l 1

l l

N URiiG-14ti$

16

1 l

4. MAllGINS AND UNClfitTAINTIES This section dncusses some of tSe snore sir icant con-cred appropnate since such f olore times are hkely to be r

servatisms and margins in the proposed accident source used pomardy in consideration of the necessary closure tetrn given m Section 3. linefly, the proposed source tet m t:me for <.ertam containment isobtion 3ahes. Smcc it is has been developed from a complete core melt accident, important that closure of such valves be vnsured before that is, assuming core melt with reactor pressure vessel the releasc of significant radioadnity to the environment, fadute and uth the assumptmn of cort-concrete interac-a conservative estimate of fuel fadute time and conse-tions. 'the timme aspects at e based primardy upon those quent onset of fission proJuct appearance is dectned ap-associated with a 1 OCA, and the rnag!.it ude of Ihe fission propriate, l'or plant s w:th leak.before-br cak appros al for products released into containment is evtimated to be then reaaor coolant system piping, a loager duration that asviciated with mean values for a typical low-before fuel clad failure is expected. Ilowever other con-pressuec cmc. melt scenario.

str,iints may become the limiting factor m containment isolation valve closure time.

4.1 Aceitlent Severity aiul Type 4.3 Itelease Phase thirations As to severity, the accident chosen for determming an 1he durations of the various telease phases have been accident source term is a cornplete core-mcit accident, selected primardv by examination of the values available mcludmg the consequences of teactor pressure vessel for the group of'sevcre accident scenarios considered in failure with evolution of fission products arismg horn Section 3. The duration oi the carly in vessel and en core concrete mleractions. the accident selccted is one vessel release phases difier for ilWit's sersus pWits and in which core melt occurs at low pressure conditions. A reflect the differing core heatup rates as well as the differ-low pwssure core melt scenano results in a relatn cly low mg amounts of /irconium available to supply chemical level of fission product retention within the reactor rool-energy after core. melt. Whde the selected duratior. if ant system, and a consequently high level of release of the release phases are realistic, some conservatisms fission products from the care m, to containment durml should be noted.1he duration of the caily in-vessel re-the early in venel release phase. Since the bulk of the lease phase for llWRs and pWRs is shott and does not fission products entering containment do so during the represent a probabihstically weighted average or mean early in-venel reicase phasa, selection of a low pressure value for the accident sequences considered. 'lhis will core melt scenarie provides a high estimate of the total intnulure a given qu:mtity of fission pnW s mto ccn-quantity of fission products released into contamment, as tainment in a sha ter time than nught be expected for a well as that dming the early in vessel release phase.

gAsiad t.cs1uence.

Sin Harly, the duranon of the cove: sel telease phase.

4.2 Onset of 14ission Protluct llelease whde considered retlistic for the bulk of the fission prod-ucts being released,is short for releases of tellurium and

,the onset, or earliest time of appearance of finion ptod-ruthenium since, as noted m Section 3.3, release of these ucts within containment, has been selected on the basis of nuchdes occurs over a longer time.

the earlic3t time to f adure of a fuel rod, given a design basis 1.OCA. Ths is estimated to be fu.m about 13 to 25 The selected release duration times have been chosen seconds for plants that do not have leak before-break primanh on the basis of simplicity, since an accurate approval for their reactor coolant system piping, and it is determination of the duration of the release phases de-expected to v ay dependmg on the reactor as wcil as the pends not only on the reactor type but alsoon the applica-fuel rod (liom the awumption of mstantr 'ous appear-ble acciJent sequence, which vatics for each reactor de-lesign. *lhis vahie, while representing some re'

~

laation g n.

ance,is nevet theless conservatn e. As noted in Reference 15, these estimates are valid for a double-ended rupture of the largest pipe, assume that the fuct rOJ is being 4,4 Con 1 position ani(1 Magnittitle of opeinted at the maumum peaking factor permitted by the l{cleases phnt Technical Specifications and at the highen burnup levels anticipated, and assume that the emergenev coie The composition of the hssion pn> ducts is based on the cooling system diC"S) is not operatmg. Use of more groupmg developed with the S'l CP and drseussed m Sec-realistie auumptions for any of these parameters would Don 3 A. The magnitude of the hsuon products released mercase e>ttmated times to fuel twl faduce by factors of into containment for the accident source term was se two or more. Nevcitheless, the use of conservative as-lected to be the mean values, using Nlh(IRI150 mcth-sumptions m estimating l'uct rod failure times are consid-odology for itWit and pWR loeprewuie venanos D

N U Rl:G-14h5

l L.

involving hich estimates of /irct nium oxidation. 'thest volatile nuclides such as iodinc and tesium, with tellunum values are shown it; Tables 3.11 and 3.12.'lhe uncertair ty to a somewhat lesser degree. The uncettainly distribu-distributions for the m vessel release and total rettase tions for this group of radionucli' es is also the smallest, a J

into containnient nie displayed graphicedly in Appendix shown in Ihe graphical tabulations of Appendix A. Hence, A, llounding estirnates for the releases into containtnent our ability to predict the behavior and releases for this taken from Reference 17, using the STCP rnethodology.

group of nuclides is sigmficamtly better than for other are shown in Appendix II. Upper bound estimates, tabu-fission product groupings.

lated in App"r. dix 11, indicate that vittually all the iodine and ecsium could enter the containment Similarly, for "Ihe mean value estimates selected for the in-

- tellurium, upper bound eshmates indicate that as much as containtnent accident source term provide reasonable es, about two-thuds of the core invernory of tellurium could timates for the important nuclides consisting of iodme, be relem ed into containment. llence, for this important cesium, and tellurium.These estimates show a telativcly group of radionuclides (iudine, resiun., and tellurium),

low degree of uncertainty and are unlikely to be exceeded the upper boutid estimates of total telease mto contain-by more than 50% Uncettainty increases in estimating ment aie approximately 1.5 times the me m value esti-releases for the remaining nuclides, and for tl.c refraciory rnaler..

nuclides, the mean value uuld be low by as much as a factor of five.

- For the lower volatility radionudides such as barium and strontium, upper bound estimates range frotn about 50 to 70% of the core inventury released into containment.

4.5 Iodiile Chemical l<orm 4

Almost all of this is estonated to be releascd as a iesult of core concrete interactions. In contr:,st, mean value esti.

The chemical form of iodine entering containment was mates range from 15 to 25% lierre, in this case, the inv':stigated in Reference 18. On the basis of tL3 work, upper bound estimates are about twa to three times the the staff concludes that iodme entering containment from incan values.

the teanot coolant system is composed of at least 95%

cesium iodide (Cs!), with no more than 5% I pim: 111.

l'inally, for the refractory nuclides such as lanthanum and Once within containment, highly colubic cesiurn iodide cerium, the upper Imund estimates indicate that about will readily dissolve in water pools and plate out on wet 5% of the inventory of these nuclides coukt ippear within surfaces in ionic form. Radiation induced conversion of contt inment, whereas the mean ialue estimate indicates Ihe ionic form to elemental iodme will ]vtentially be an ordy about 1% released.

impor tant mechanism. lf the pil is controlled to a level of 7 or greater, such conversion to elemental iodine will be l'R As have indicated ihat, considering the magnitudes of mir: mal,if the pilis not controlled, however, a relatively the radioactive species estimated to be teleased to the large fraction (pleater for PWRs than llWRs)of the io-entitontnent fer severe reactor accidents, the radionu-dme dissolved in containment pools in ionic form will be ehde:s basing the greatest impact on risk are typically the converted to elemental iodine.

)

NU R F.G-145

.18 s

5. IN-CONTAINMENT ltEMOVAL MECilANISh1S Since radioactive fission proJocts within containment aie mo.al becoine, /ero after some reduction has been in the form of gases and finely div;Jed airborne particu-achiesed,or changing to a much smaller value of lambda lates (a nosols), the principal mechanisn; by w hich fission to tenect the decreased removal chcetiveness of the products find their way from the reactor to the environ-spray when airborne concentrations are low.

rnent with an intact containment is via leakage f om the containtnent atmosphere.'The specific Iission proJuct in-g, &cuon 6.5.2 (itef. 20) provides expr essions for cal-ventory present m the contananent atrnosphere at any culatmg spray lambdas, dependmg on plant parameters as time depends on two factors: (1) the " source, ie..the well as t he ty pc of species removed. In addition, Sit P 6.5.2 rate at which fission products aie bemg introduced into uMy requires that the containment sump solution be the contain sent atmosphere, and (2) the " sink, the rate maintamed at edues at or abm e plIlevels of 7,commenc-at which they ate bem.g removed. Aspects of the release ing with spray recirculation, to nunimve revolatili/ation and trimsport of fission produ-ts from the core into ine of odine m the st mp v ate r. Cerrent guidance states ttat containment atmosphere were presented m hection 3.

nntainment spra r splems be initiat ed automatically, be-cause of the instanDmeous appearance of the source term Mechanisms that removt lission prodacts from the at.

within containment, and that the spray duration not be mosphere with consequent mitigation of the in-contam-less than 2 hows. In contrast, the revised source term ment source term fall into two classes: (1) engineered mformation rivei. in Section 3 suppests tha' spray system safety fea tures (liSFs) and (2) natural processes. liSI s to actuation mipht he somewhat uclayed for radiological remove or r educe fission products within the containment purposes, but that the spray system duration should be a are presently required (Criterion 41 m Appendix A of 10 longer period of about 10 or more hours.

CFR Part 50) and include such systems..a containment atmospw sprays, itWR suppression pools, and fdtra-

't he sprm removal coefhcient for particelates appears tion sptems utilving both particulate filters and chari oal particularly impottant in view of the information pre-

~

adsorption beds f or the removal of : 'dme, particularly m sented in 'Section 3, which indicates that most fission elemental for m. Natural removal meludes such processes products are expected to be in particulate f orm.The spray as aerosol deposttion and the sorption of vapors on equip.

iemoval ccefficient (A) is denved from the followmg

~

ment a.d structural surfaces. lhese at e discussed in more equMion fmm Surdard Resiew Plan Section 6.5.2 detail below.

A ;)hifi 2m SJ Containment Sprays 7

h& Fall heicht of spiav drops Containment spreys, covered in Standard Review Plan (SRP) Section 6.5.2 (Ref. 20) ate used in many PWR V = Containment buddmg volume designs to provide post. accident containment cooling as l, m S. pray Gow well as to remove released radioactive acrosols. Sprays ate effective in reducing the airborne concentratbm of IUD n the ratio of a dimensionless collection effi-elemental and particulate iodines as w ell as other partico.

cienev li to the average spray drop Diameter D.

lates, suca as cesium, but are not effective in removmg F./D is conservativeh assun$ed to be equal to nobic gases or organic forms of iodine. 'the reduction in 10/ meter for sprav drops 1 mm in diameter chang-airborne radioactivity within containment by a spray sp-mg to Umeter Jhen the actosol mass has been tem as a function of time is expressed as an exponential depleted by a factor of 50.

reduction process, where the spray removal coefficient, lambda, is taken to be constant over a large part of the Using values typical for PWRs, the formulation given in regime. 'lypical PWR containment spray systems are ca.

SRP 6,5.2 estimates particulate remm al rates to be on the puble of rapidly reducing the concentration of firbotne order of 5 per hour 'the staffis tently evaluating more activity (by about 2 orders of magnitude within aoout 30 ieahstic s, ay remov. models, Nourbakhsh (Iter. 21)has minutes, where both spray trains are operable). Once the eumined the ef fectreeness of contamment sprays, as bulk of the aetnity has been remosed, however, the spray evaluated in NURIE1150 (Ref. 7L in decontam;nating becomes significantly less effective in reducing the re-both in.vesee! and envessel releases. 'lhe overall effec-maining fission productsflhis is usually accounted for by tn eness of spray systems. gnen in terms of a decontami-cither employing a spray cut-off, vherem the spray ie-nation factor (DF). are shown m Tables 5,1 anJ 5.2.

lu NURIE1465

Q P

T lt 1-q.

1 Table fi,1 Distributions for Spray Decontamination ofin-Yeuel Relca-es

• ~*

i CONDITIONS Sth percentile DF Mediac DF Mean DF 95th percent DF

PWRs-liigh pressure 1.6 1.8 1.8 2.2 with containment failutc

at vessel breach

.pWRs-ll;gh pressure 3.0 40.0 12.7 1800 with no containment "lailure '

11WRs '

1.3 11.0 5.2 78 Table 5.2 Distributions for Spray Decontamination of Core-cnnerete Releases CONDITIONS 5'h percentile DF Median DF hican DF 95th percent UF PWRs 7.7 -

28, 23.

2000 HWRs '

1.5 17, 7.

480 these tables present the range of DFs associated with.

ranged from 1.2 to about 4000 with a median value of containment spray systems. Tabic 5.l shows, for example, about 80. The suppression'peol has been shown to oc

- that the mean DF value for PWR in vessel sprays are effective in scrubbing some of the most imitrtant radio-atmut 12 for low pressure sequences. Table 5.2 indicates nuclides such t's iodine, cesium, and tellurium, as these that sprays are also effective in reducing acrosols evolved are released in the early in. vessel phase Resulte from from ex. vessel releases.-

Reference 21 shown in Table 5.3, present the range of DFs as evaluated in NUREO-1150. It should be noted from this tab!e that in vessel releases are more highly 5.*

G R ylpprcSS10D pools scrubbed (larger DFs) than ex. vessel releases.

Ture suppression pools to condense steam if not bypassed, the suppresnion pool will also be effective ~

IlWx a

resulting ha a loss-of. coolant accident. Prior to the in scrubbing ex vessel releases. Suopression pool bypass L release to the reactor building, these pools also scrub is an important aspect that places an upper limit on ti.e radioactive fission produck that accompany the steam.

overall performance of the suppressien poolin scrubbing -

Regulatory Guide 1.3 (Ret,2) does not allow credit for -

fission producte' For exampi:, if as little as 1% of the -

4

. fission product scrubbing by IlWR suppression pools, but

. fission products bypass the suppt ession pool, the effective SRP Section 6.5.5 (Ref. 22) has recently been revised to :

DF, taking bypass into account, will be less than 100,:

allow such credit. The pool water will retain soluble,-

regardless of the pool's ability to scrub fission products.

gaseous, and solid fission products such as iodines and

cesium but provide no attenuation of the noble gases Although decontar ination factors for;th' suppression -

l released from~ the core. 'lhe Reactor Safety Study pool are significant, a key question is the potential for (WASit-1400, Ref. 5) assumed a decontamination factor '

- iodine re evolution. Re evuution of iodine was judged to H

~ (DF) of 100 for subcooled suppression pools at:d 1.0 for bc important in accident sequencas wl.cre the contain.

steam saturated pools.S'nce 1975 when WASil-1400 was ment had failed and the suppression pool'was boiling.

_ published, several detailed models have been developed There is presently no requirement for pli control m BWR for the removal of radioactive ne, sc!s during steam flow suppression pools. Ifence it is possible that suppression

through suppression pools.

pools would scrub substantial amounts of iodine in the

~,

early phases of an accident, only to re. evolve it later as Calculations fer a BWR with a Mark I containment (Ref.

elementaliodine it may well be that addidonal materials

( 23) used in NUREG-1150 (Ref. 7) indicate that DFs.

likely to be in the suppression pool as a result of a severe 3

.- N UR110-1465 -

20

..a

l l

Table 5,3 Distribution for Suppression Pool Decontamination Fartors 3

CONDITIONS 5th perctutile DF Median DF Mean DF 95th percent DF l

3 Duringinesselreleases j

r Peach bottom 2.3 81 14.5 1200 laSalle & Grand Gulf 1.8 56 10.5 2500 During Et wsselreleases l

1 Peach flottom 1.2 9.5 5.1

- 50 12Salle & Grand Gulf 1.2 6A 4.0 72

]

accident, such as cesium borate or cesium hydroxide and Revised insights on accident source terms, given m Sec-core conerete decom}msition products, would counteract tion 3, may have several implications for llSF filtration any reduction in p!! from radiolysis and would ensure that systems. Present ESF filtration systen s are not sized to the pil level was sufficiently high to preclude re-handle the mass loadings of non-radioactive acrosols that evolutioc of elementalicjinc.Therefore if eredit is to be might be released as a result of the ex vessel release given forlong term retention ofiodin. b the suppression phue, which could produce releases of significant quanti.

l

' " ' * " ' " ' * " " " ' * " ' ' " " " " ' " ' " ' * " " ' " " * ' " ' ' * " ' ' '"'"""'""'"""""'"'"""di"""""'"

Fic demonstrated. It is important to nu. however, that llowever, if ESF filtration systems are employed in con-this is not a matter of concern for present plants since all junction with HWR suppression pools or if signiFcant i

llWRs employ safety related filtratica systems (see Sec-quantitics of water arc overlaying molten core debris (sec tion S.3) designed to cope with large quantitics of clemen.

Scetion 5.4), large quantitics of nonradioactive (as weil as tal lodine, llence, even if the supp.ession pool were to radioactive) acroso's will be scrubbed and retained by re-evolve significant amounts of elemental iodine, it these water sources, thereby reducing the acrosol mass would be retained by the existing downstream filtration loads upon the filter system.

system.

A second implication of revised source tenn insights for ESP filtration systems is the impact of revised under-5 3 Filtration Systems standing of the chemical form of iodine within contain-(Tient. Present ESF filtration systems presume that the USF filtration systems are di.aussed in Regulatory Guide chemical form of iodine is primarily elemental iodine, and

'l.52 (Ref. 24) and are used to reduce the radioactive these systems include charcoal adsorber beds to trap and acmsols and iodine released during postulated accident retain elemental iodine. Assuming that pil control is

. conditions.

maintained v.ithin the containment, a key question is whether charcoal beds au necessa.j, Two questiorm p.

- A typical USF filtration system consists of redundant pear to han a bearing or, this issue and must bv ad.

dressed, evenssuming pH control.These arc (1) to what trainsihat cach have demisters to remove steam and degree will C si retained on particulate filters decompose water droplets from the ait entering the filter bank, heat-to evok &meraal iodine? and (2) what effect would ers to reduce the relative humidity of 'he air, high effi, hydrogen burns have on the chemical form of the iodine ciency particulate air (HEPA) filters to remove particu-withm containment? llased on preliminary information, lates, charcoal adsorbers to remove iodine in elemental Csl retamed on particulate filtets as an nerosol appears to and organic form, followed finally by addnional ifEPA be chemically stable provided that it is not exposed to filters to remove any charcoal finet released.

moisture. Exposure to moisture, however, would lead to Csl decomposition and production of iodine in ionic form Charcoal adsorber be6 can be designed, as indicated in (1-), which in turn would lead io re-evolutior of elemcntal Regulatory Guide 1.52, to remove from 9010 99% of the iodine. Although ESF filtration rystems are equipped -

elemental lodine and from 30 to 99% of the orgar'ic io-with demisters and heaters to remove significant moisture dide, depending upon the rpecific filter train design, before it reaches the charcoal adsor ber bed, an additional 21 NUREG -145

l l

i conceru is that the demisters themselves may trap some is discussed in some detail. Aerosols are usually thought C51 aemsol.

of as sohd particulates, but in general, the term also includes finely divided liquid droplets such as water, i.e.,

in conclusion, pr esent I.SI' filtration systems, w hile opo.

fog. 'lhe two major sources of aerosols are condensation mired to remove iodine, particularly in elemental form, and entrainment. Condensation aerosols form when a have llEPA filters that are effective in the removal of vapor originating from some high temperature source moves into a cooler region w here the vapor falls below its particulates ns well, Although such fibration Systems arc saturation temperature and nucleation begins. lintrain-not designed to handle the large mass hiad ngs expected ment aerosols form when gas bubbles break through a as a result of ex vessel releases, when they are used in hquid surface at d dr9 dioplets of the liqmd phase into conjunction with large water sourtcs such as llWit sup.

the wake of the bubble asit leaves the surface. In general, pression Iwls or significant w ater depths overlaying wre condensation particles are smaller in site (submicron to a debris, the water sources will reduce the acrosol mass few microns), while entrainment particles are usually loading on the filter system significantly, making such larger (LO mu microns) Once airborne, both types of filter systems cifective in mitigation of a large spectrum of acrosols behave in a similar manner with respect to both accident sequences, natural and engineered removal processes.

5A -Water Overlying Core Debris There are four natural processes that remove aerosols from the containment atmosphere over a period of time:

INperimental measurements (ltef. 25) have shown that (1) gravitational settling, (2) diffusiophoresis, (3) ther-signif cant depths of water overlying any molten core mophoresis, and (4) particle diffusion. (Particle diffusion dcbris after reactor pressure vessel failure will scrub and is less important than the first three pr ocesses and will not

- retain particulate Insion products. The question of be discussed further.) All particles fall naturally under cr.alability of the inolten debris as a result of water overly-the force of gravity and collect on any available surface ing it is still under investigation. At the present time, the that terminates the fall, e.g.. the floor or upper surfaces of degree of scrubbing as a f unction of water depth is under equipment. Iloth diffusiophoresis and thermophoresis investigation by the staff, A major factor that rnay affect cause the deposition of netosol particles on all surfaces the degree of scrubbing is whether the water layer in regardless of their orientation, i.e., walls and ceiling as contact with the molten debris is boiling or not, wc!! at the floor. Dlffusmphoresis is the process by w hich water vapor in the atmosphere ' drags' acrosol particles Results from 1(cf. 25, shown in Figure 'c indicate that with it as it migrates (diffuses) toward a relatively cold both subcooled as well as boiling water layers having a surface on which condensation is teking place. Ther-depth of about 3 meters had measured DFs of about 10, mophoresis also causes acrosol particles to move toward Calculations made using the SPARC code agreed well and deposit on colder surfaces but not as a result of mass with meavured values for a subcooled pool, but the calcu-mot"n. Rather, the decreasing average velocity of the lated DF for a boiling pool drupped from a value of 10 to sunounding gas molecules tends to drive the particle about 2 *the staff intends to give cretht for fission product down the temperature gradient until it traverses the in.

and other aerosol scrubbing where significant depths of terface layer and comes into contact with the surface water are shown to overlie any molten core debris, liased where it sitcks.

on the results of Ref. 25, a DF of 10 for a water depth of about 3 meters is considered appropriate. Lesser depths will be assigned lower DF values. '

Acrosol agglomeration is another natural phenomer.an that has an influence on the rates at which the removal Results from lieference 21, shown in Table SA, present processes described above will proceed. Agglomeration the mean and tnedian DF values, as evaluated by results fmm the random inelastic collisions of particles NURl!G-ll50 (Ref. 7), for an overlying pool of water, with each other. The process brings about a gradual in-Results are shown for several PWRs and UWRs. These crease m average particle sue resultmg in more rapid indicate that, w here the reactor eavity can be flooded, DF gravitational settling. Three phenomena contribute to -

values of about a factor of 10 or more are estimated. For particle growth by agglomeration: (1) Brownian motion, geometries where deep flooding is not feasible, howeser, (2) gravitational fall,- and (3) turbulence. Brownian +,g-i i'

lower DFs ranging from 2 to 5 have been evaluated.

clomeration is caused by particle collisions resulting from random ' buffeting' by high-energy gas molecules. Gravi-tational agglomeration results from the fact that some 5.5 Aerosol Depositioit particles fall faster than others and therefore tend to collide with and stiC to other slower falling particles on Since the principal pathway for transport of fission prod-their way down. Finally, rapid vanations in gas vehicity ucts is via airbonic particulates, i.e., aeromis, this subject and flow direction in the atmosphere, i.e., turbulence, NURiiG-l#5 22

--s-ne e,

,ms,

.-w w-,

w em-s we

- -~ - - - - m,y w<,

~,m e

m -<

sw

-m.

e,-e,-e,we m wn nrv,-

w

,----...w*~

-, -~,-- ---- -

-L------

DF 10' LATEX diameter (# a) 01.o9 Pool sater temperature O o. 5 9 Open symbol 80 t (subcool)

Closed symbol:110 t (bolling) jga d o. 2 s O

O o

10' Calculation U

(Subcooled) p O

O 105 t>

A Calculation O

CB III"8)

O y o h

IK-

"--~~~~~~"~

10' 1

2 3

4 Scrubbinz depth (meter)

I'.gure 1 i

Scrubbing depth and pool water effect on DF compared with prototype SPOARC calculation for 1 inicro meter IXfEX 23 NUlt!!G-1465

Table 5.4 hiedlan and Mean Detontamination l'attors for Water Pools Overlying Core Debris

• CONDITIONS I

Cs Te St Itu La Ce lia PWiw Zion & Surry-5.5 5.5 2.1 5.5 2.1 5.5 5.3 5.5

- slmtlow water (4.3)

(4.3)

(2.4)

(4.3)

(2.4)

(4.3)

(4.3)

(43)

Zion & Surry-30 30 15 30 15 30 30 30 ihded Cavity (13)

(13)

(5.8)

(13)

(5.8)

(13)

(13)

(13)

Sequoyah-5.

5.

2.

5.

2.

5.

5.

5.

shallow water

.(4.2)

(4.2)

(2.1)

(4.2)

(2.1)

(4.2)

(4.2)

(4.2)

Sequoyah--

25 25 13 25 13 25 25 25

~

flooded cavity (2?)

(21)

(11)

(21)

(11)

(21)

(21)

(21)

IlWits :

Peach llottom 4.4 4.4 4.4 4.4 4.4 4.4 4.4 4.4 --

A I ASalle (3.1)

(3.1)

(3.1)

(3.1)

(;.1)

(3.1)

(3.1)

-(3.1)=

Grand Gulf--

4.4 4.4 4.4 4.4 4.4 4.4 4.4 4.4 shallow vater (3.1)

(3.1)

(3.1)

(3.1)

(3.1)

(3.1)

(3.1)

(3.1)

Grand Gulf-6.

6.

6.

6.

6.

6.

6.

6.

ikioded cavity (3.9)

(3.9)

(3.9)

(3.9)

(3.9)

(3.9)

(3.9)

(3.9)

'Mean values are shown in parentheses tend to increase the rate at'which particle co!!isions occur havior in the reactor containment is found in CO! TAIN, and thus increase the average particle sire. It is to be a computer code developed at Sandia Nationall.aborato-expected that, as agglomeration advances, the size of the ries under NltC sponsorship for the analysis of contain-

' particle will increase, and its shape can be expected to ment behavior under severe accident conditions. The chan;c as w ell. These latter factors have a strong influ-acrosol models in the NAUA code are _very similar to ence on the removal processes.

those used in CONTAIN; NAUA was developed at the Kernforschungszentrum, Karlsthue, F.11.0.. and was used for aerosol treatment in the NltC STCP, Rete are a

%e agglomeration and acrosol removal processes t11-

. depend critically upon the thermodynamle state anu

'1 umber of other well-known aerosol behavior computer thermal-hydraulic conditions of the containment atmo-codes, but these two are the most widely used and ac-sphere. For example, the condensation onto and evapora-cepted throughout the mternational nuclear safety com-E tion of water fiom the aerosol particles themselves have

ruunhy, strong effects on all of the agglomeration and removal The rate at which gravitational settling occurs depends-.

processes; Water condensed on aerosol particles in' crec.ses their mass and makes them more spherical: both upon the degree of agglomeration at any particular time of these effects tend to increase the rate of gravi *ational

(.c., the average particle size) as well as the total particle settling. Some aerosols, such as Csl and Csol l, are 1.ygro-density rn (mass per unit volume). Thus, as in most cases scopic had absorb water vapor even when the contain-where the decrement of a variable is proportional to the

- ment atmosphere is below saturation. As with condensa-variable itself, one can expect an exponential behwlor.

tion, hygroscopicity also inercases the rate of deposition.

The gravitational settling process is quite cornplex and depends upon a large number of physical quantities, e.g.,

collision shape factor, particle settling shape factor, gas

' liceause of its impoitance to fields such as weather and viscosity, effective settling height, density correction fac- -

atmosphere pollution, the behavior of aerosols has been -

tor, normalized llrownian collision cocificient, gravita-under study for _many decades. A number of computer tional acceleratiot, and particle material density. De codes have been develor.ed to specifically consider aeros only variable in this list that is independent of the plant,-

sol behavior as it relates to nuclear accident conditions.

the accident scenario, and the atmospheric thermal.

De most complete mechanistic treatment of acrosol be-hydraulic conditions is the constant of gravitation. It

- NUREG-1465 24 -

follows that no smgle Di can be awnbed to cover the fashion, the removal rate coef fiaents, lamidt, obtained entire ranpc of plant designs, accident scenarios, and from the survey for each plant are shown m Table 5.5.

rs.urce snaterials. An cifort is under way to establish a set of umphfied algorithins that can be used to provide a set Containment atmosphere acrosol iemoval rates were of specific ranges of atmosphere conditions also evaluated for the NUlRG-1150 plants plus I aSalle (itel. 21) and are show n in Table 5.6. I he major factor to be noted in natural acrosol deposition is its relatively slow The staff has, however, made a prehminary survey of rate wmpared to engineered features, such as sprays.

acrosol removal rates frorn the containment atmmphere Spray systems are capable of reduemp the containment based on review of a selut:d number of severe a'

.cnt atmmpheric concentration by about an order of snagni-sequences. Usmg data froin 1(eferences 26 and 11,.ud tude in about 15 to 20 minutes. but natural aerosol depo-with the smu.nption that the acrosol eoncentration within ntion processes will require 4 to 10 hour:,, depenJmg on contamment decreases with time in a simple exponential the plant and sequence, for the same reduction factor.

A Table 5.5 Containment Atmmphtse Aerusol itemmal Itates Acrosol itemoval Plant flate (per hour )

Zion 0.18 Surry 0.2 Peach llottorn 06 Grand Gulf 0.5 Sequoyah UP 1 rom NUlti.G-0772 0.33

'Ite beds partially operating durmg sequence.

'Iable 5.6 Containintnt Atmmpheic Aciosol lleounal ltates Pi ANT Altl% tm?)

VOI 1311'. On')

lll310 val IIATl; (per Imut )

Zian 1.44 X (()

7.6S X 104 0.18 Su: ry 1.25 X 10' 5.1 X 104 0.23 Sequoyah K36 X 1(F (1 C)*

1.1 X 104 (l.C) 0 73 5.7S X 10/ (UC) 2.54 X 104 (UC) 0.21 Peach Itottom 152 X 10'(DW) 4.5 X 10)(1)W) 0.32 I aSalle 4.99 X IF (DW) 6.51 N 10)(DW) 0/13 Grand Gulf 3.71 X 102 (DW) 7.65 X 10' (l)W) 0.46

• 1f refers to 1ower Comf.artment UC tefers to Upper Compartmtnt DW refers to Drywell 25 N1'Id1 b1465

6.11EFEllENCES 1.

U.S. Nuclear llegulatery Comminion: *lteactor moval C(de," NU RiiG/CR-3317 (PNi e 4742). pre-Site Coter m." Title 10, Code of l'ederal Regulations pareJ for NRC by llattelle Pacific Northwest (C Ht). Pat t 100.

I ahoratories, hiay 1985.

2.

U S. Nuclear Regulatory Commission: " Assu mp-

12. WK Winegardner, AK Postma, and hi.W.

tions Used for livaluating the Potential Radiological Jankowski, " Stud;es of l'ission Prmluct Scrubbing Consequences of a loss of Coolant Accident for within lec Compartments," NURIIG/CR-3248 thuling Water Reactors," Regulatory Guide 1.3, (PNI.-4691) prepared for NRC by llattelle Pactfic itevision 2, June 1974.

Northw est laboratories, hiay 19d3.

3.

U.S. Nuclear Regulatory Commission: " Assu mP-

13. I I. lloni hi. Kavr o, and W. Schock,"N AU A-hhx14:

tior's lhed for livaluating the Potential Itadiological A Code for Calculating Acrosol llehavior in LWP C3nsequences of a Ixss of Coolant Accident for Core hielt AcciJents," KlK-3554 Kernforschung-Pressurized Water Reactors," llegulatory Guide sientrum. Karlsruhe Germany,1983.

1.4, Revision 2, June 1974.

14. R.hi Summers, et al., "hiliLCOR 1.8.0: A Com-4.

JJ.1%unnoctal.,* Calculation of Di tanee l: actors puter Code for Nuclear Reactor Sescre Accident for Power and Test Reactor Sites," Tcchnical Infor.

Source Term and Itisk Assessment Analysis,"

v rnation Document (11D)-14844, U.S. Atomic lin.

NURl;G/CR-5531 (SAND 90-0364), prepared for ergy Conmussion,1962.

NRC by Sandia National laboratotics, January 5.

U.S. NuJear llegulatory Commission: "Itcactor Safety Study: An Assessment of Accident llisks in

15. K A Jones. ct al," Timing Atialysis of PWR 1 uel Pm U.S. C ommerctal Nuclear Power Plants " W ASil-IWures," Draft NURiiG/CR-5787 (l!GG-2657),

1400 (NURl!G-75/014), December 1975.

prepared for NRC hy Idaho National !!ngineering I aboratory, h1 arch 1992.

6.

J. A. b,leseke et al,"h,ource.I,crm Code Package: A User's Guide," NURiiG/CR-4587 (RN11-213S),

16, ll.P. Nourbakhsh, hi. Khatib-Itahbar, and R.1:.

prepated for NRC by llattelle hiemorial Institute, 1

Daus " Fission Product Release L haractensticsinto July 19%'

Containment Under Design liasis and Sescre Acci-dent Conditions," NURiiG /CR-48SI (ilNic 7.

U.S. Nuclear Regulatory Commission: " Severe At.

cident ltisks: An Assessmen'. for Five U.S. Nuclear NURIIG-52059), prepared for NRL by llrookhaven Power Plants," NUltliG-ll50. December 1990.

National 1 aboratory, h1 arch 1988.

17. II.P. Nourbakhsh: listimates of Radionuclide Re-8.

NiR KuMman, DJ. l.chmicke, and R.O. hicyer, "CORSOR User's hianual/ NURl!G/CR-4173 lease Charactenstics into Containment Under Se-(IIN1I-2122), prepared for NRC by llattelle hiemo.

vere Accidents" Draft NURiiG/CR-5747 (llNic rial Iahoratory, h1 arch 1985.

N U RiiG-52189), prepared for N RC by llrookhaven National l aboratory, January 1992.

9.

11. Jordan, and ht A Kuhlman, "TR AP-hililll'2 User's h1anual" NURl!G/CR-4205 (llN11-2124),

, b.

ILC. licahm, L.F. Weber, and L S. Kress, " lodine prepated for NitC by llattelle hiemorial L abo a.

Chemical I orms in 1.W R Sescre AcciJents "

tory, hiay 1985.

NUREG/CR-5372 (ORNIKht-llb61), prepared for NRC by Oak Ridge National laboratory, April 10.

D. A. Powers, J.l!. Prockmann, and A.W. Shiver, 1992.

"VAMliSA: A hicchanistic hhxlet of RadionucliJe Release m:J Aerosel Generation During Core De.

19. h1. Silberberg et al," Reassessment of the Technical bris interactions with Concr ete,"

NUREGI liases for listimatmg Source Terms," NUREG-CR-4308 (S AND 85-13701 prepared for NRC by 0956, July 1986.

Sandia National Iahoratories July 14S6.

20.

U.S. Nuclear Regulatory Commission: "Contam-

11. IT. Owc /arski, AA Postma, and R.I. Schreck, ment Spray as a 1ission Product Cleanup System,"

"l'echnical llases and User's hianual for the Proto-Standard Review Plan, Section 6.5.2, Revmon 2, type of SPARC-A Suppression Pool Aerosol Re-NURPG-OStXL Deccmber 19SS.

N UR EG-1465 26

1 l

21. f l.P, Nourbakhsh: "In-Containment itemoval

!!ngineered Safety.I'eature Atmosphere Cleanup h.echanisms " Presentation to NitC itaff january 3, System Air 17iltration and Adsorption Units of 1992, llrookhaven National lalxitatory, January light-Water-Cooled Nuclear Power Plants," llegu-1992.

latory Guide 1.52, llevision 2, March 1978.

22.

U.S. Nuc! car llegulatory Commission: " Pressure

25. J. llakii et al.,"lixperimental Study on Acrosol lle.

Suppression Pool as a Fission Product Cleanup Sys.

movat lifficiency for Pool Scrubbing Under lhgh t e m,"

Standard Iteview Plan, Sectien 6.5.5, Temperature Steam Atmosphere," Proceedmgs of NUlti!G-0800, December 1988.

the 21st IX)li/NitC Nuclear Air Cleaning Confer.

23 it.S. Denning, et al., "lladionuclide Itclease Calco-lations for Selected Severe Accident Scenanos:

26.

J.A. Gieseke: "Itadionuclide llelease Under Spe-11Wl( Mark i Design," NUltliG/ Cit-4624, Vol.1, cifu:I.WR Accident Conditions," IIMI-2104, Vok.

prepared for NitC by llattelle Memorial Institute, 1-Vil,11attttle Memorial 1 aboratories,1983.

27.

U.S. Nuclear llegulatary Commission: " Technical

24. U.S. Nuclear llegulatory Conunission: " Design, liases for I!stimating Fission Product flehavior Dur-Testing, and Maintenance Criteria for Postaccident ing 1. Wit Accidents," NUlt!!G-0772, June 1981.

p 27 NUlt!!G-1465

APPENDIX A UNCERTAINTY DISTRillUTIONS 1

1 i

l 1

29 Null 'G-1465

- - ~.

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10 r sis 3

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(6) Low Zirconium Oxidation (High Zr Content in the Melt)

Uncertainty Distributions for Total Re'ases into Coat inment PWR, Low RCS Pressure. Lime-stone Concrete, Dry Cnity, Two Openings Aner VB. 3. ART = 1 s,-

NtilEG-1465 30

IC[ 7:

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Uncertainty Distributions for Total Retcases into Containment PWK Low ItCS Pressure. Basaltic Conente, Dry Cavity, Two Openingt After VB, FPART = 1.

31 NURl!G-1465

L 10 J]

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(b) 1.aw Zirconium Oxidation (Iligh Zr Content in the Meh) 11ncertainiy Distributions for Total Releases into Containment BWR, law Prusuit fast Sistion Blackout Umestone Concrete, Dry Pedestal, Low Drywell Temperature, FPART a 1.

'.NUREG-1465 32 1

AI'PENDIX 11 STCI' IlOUNDING VALUE llEl. EASES s

4 33 NUlWG-1465

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E 35 NUREG-1465

-~

,~

.~

NRCPORM P6 U.8. M.JCLEAT REGui.ATORY COMM:S$10N

1. REPORT NUMBE4 P-89) -

(Assigned by NRC, Add Voi.

54tCM 1102.

$upp, nov. and A4 der.de N an.

3 N 3232 BIBLIOGRAPHIC DATA SHEET b- * - I' *"vl (so. instructma on vos reverso>

NUREG-1465 2, UTLE AND SUullTLE 3 DATE AEM.)HT PutsuSHELI Accident Source Terms for Light Water Nuclear Power Plants y3y7g yg,g June 1992 Draft P.cport for Comment

4. NN OR GRANT NUMBER

6. AUTHOR [56 6, TYPL OF HE K)M1 L Soffer, S. f. flurson, C. M. Ferrell, R. Y. Lee, J. N. Ridgely

7. PERIOO COVERFO 't.clusive Dates) i

$, PUlFOHMiNG CMGANIZ AllON - NAME AND ADDRESS (if NRC. prov6de Devision, Office or Aegm. U S. NuA Itegi.1 story Cornmission, and mcHing modress; if contractor, prov'de name and malling aMress )

Division of Safety Ime Resolution Office of Nuc! car Regulatory Research

'l U.S. Nuclear Regulatory Commission Washington, DC 20555

9. 6PONSOHNQ 04GAN12ATION - NAME AND ADDRESS (rr NHC, tyim hame as alsove'; if contractce, provide NRC Division. Off,co or Ang cn, U.S. Nuclear Regu: story Commissico, and malling adoress ?

Sarie as 8. above

10. SUPPLLMENT AHY N')TES L

11. AnstaAct (200 woros or sessi In a62 the U.S. Atornic Energy Commission published TID-14844," Calculation oi >istance Factors for Power and Test Reactors' which specified a relea:a. of fission products fam the core to the reactor containment in the event of a postulated accident invo'virg "substant'ai meltdown of the core". His " source term", the basis for the NRC's Regulatory Guides 1.3 a..,.i i.4, has been ised to determine compliar:ce with the NRC's reactor site criteria 10 CFR

. Part 100, and to evaluate other importr -' plant performance requirements. During the past 30 years substantial additionalinformation on fission prodo. releases ha been developed based on significant severe accident research.

This document utilizes this_ research by providing more realistic estimates of the " source tenn" release into contain-

- ment,in terms of tirring, nuclide types, quantitics, and chemical form, given a severe core-melt accident. This re-vised " source term"is to be applied to the design of future fight water reactors (LWRs). Current LWR licensces may voluntarily pmpose applications based upon it. These will be reviewed by the NRC staff.

12, KEY WO el/OESc4!PTORS (Ust words or phrases that witi assist researchers Iri locating the report.)

13. AVALABLITV s r ATEMENT Unlimited

"'S'" " CS* *" "

Severe Accident Source Term

" " "* 8 * )

Core Meltdowm.

Unclassified Design llasis Acciden*

U$a 5'Pa")

TID-14844 Replacement

- Core Fission Product Releases Unclassified 15 NUMBER OF PAGES

16. PRICL e-NRC FORM 335 (2-49)

.. +.

f t,

g l,..

l.:.

l THIS DOCUMENT WAS TRINTED USING RECCLED PAPER' i

i-I:

l.

-n

+

i

. NUREG-1465

' ACCIDENT SOURCE TERMS FOR LIGHT-WATER NUCLEAR POWER F1 ANTS JUNE 1992 -

2

. DRAFT -

1

-i UNITED STATES FIRST CLASS MAIL.

.- NUCLEAR REGULATORY COMMISSION

POSTAGE AND FEES PAID

WASHINGTON, D.C. 20555-0001 usnac.

PERMIT NO. G-67

- OFACIAL E5USINESG PENALTY FOR PRIVATE USE $300 i

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