ML20081A046
| ML20081A046 | |
| Person / Time | |
|---|---|
| Issue date: | 02/28/1995 |
| From: | Burson S, Ferrell C, Richard Lee, Ridgely J, Soffer L NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| To: | |
| References | |
| FRN-64FR12117 AG12-1-027, AG12-1-27, FACA, NUREG-1465, NUDOCS 9503140449 | |
| Download: ML20081A046 (40) | |
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l Accident Source Terms for Light-Water Nuclear Power Plants Final Report U.S. Nuclear Regulatory Commission l
Office of Nuclear Regulatory Research L Soffer, S. B. Burson, C. M. Ferrell, R. Y. Lee, J. N. Ridgely ps ancy I
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AVAILABILITY NOTICE Availability of Reference Materials Cited in NRC Publications t
~ Most documents cited in NRC publications will be available from one of the following sources:
1.
The NRC Public Document Room, 2120 L Street, NW., Lower Level, Washington, DC 20555-0001 r
2.
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Standards, from the American National Standards Institute,1430 Broadway, New York, NY 10018-3308.
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NUREG-1465 Accident Source Terms for Light-Water Nuclear Power Plants Final Report Manuscript Completed: February 1995 Date Published: February 1995 L Soffer, S. B. Burson, C. M. Ferrell, R. Y. Lee, J. N. Ridgely Division of Systems Technology Omce of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
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d Abstract In 1%2 The U.S. Atomic Energy Commission published information on fission product releases has been TTD-14844, " Calculation of Distance Factors for Power developed based on significant severe accident and 'Ibst Reactors" which specified a release of fission research. This document utilizes this research by products from the core to the reactor containment in providing more realistic estimates of the " source term" the event of a postulated accident involving a release into containment, in terms of timing, nuclide
" substantial meltdown of the core " This " source term,"
types, quantities, and chemical form, given a severe the basis for the NRC's Regulatory guides 1.3 and 1.4, core-melt accident. This revised " source term" is to be has been used to determine compliance with the NRC's applied to the design of future Light Water Reactors reactor site criteria,10 CFR Part 100, and to evaluate (LWRs). Current LWR licensees may voluntarily other important plant performance requirements.
propose applications based upon it. These will be During the past 30 years substantial additional reviewed by the NRC staff, t
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CONTENTS I
Page Abstract........................................................................................
iii Preface......................................................................................... vii 1 Introduction And Background.................................................................
1 1.1 Regulatory Use of Source Terms.........................................................
1 i
1.2 Research Insights Since TID-14844.......................................................
2 2 Object ives And Scope.........................................................................
3 2.1 General...............................................................................
3 2.2 Accid en ts Considered...................................................................
3 2.3 Limitat ions............................................................................
4 3 Acciden t Source Rrms........................................................................
5 3.1 Accident Sequences Reviewed...........................................................
5 3.2 Onset of Fission Product Release.........................................................
5 3.3 D uration of Release Phases..............................................................
7 3.4 Fission Product Composition and Magnitude..............................................
9 3.5 Ch e mical Form.......................................................................
10 3.6 Proposed Accident Source Erms.........................................................
12 3.7 Nonradioactive Aerosols................................................................
14 4 Margins An d Uncertainties....................................................................
15 4.1 Acciden t Seve rity and Type..............................................................
15 4.2 Onset of Fission Product Release.........................................................
15 4.3 Release Phase D u rations................................................................
15 4.4 Composition and Magnitude of Releases..................................................
16 4.5 I odine Chemical Form..................................................................
17 5 In-Containment Removal Mechanisms...................................................
17 5.1 Containment Sprays..................................................................
18 5.2 BWR Suppression Pools.................................................................
18 5.3 Fil tration Systems......................................................................
19 5.4 Water Overlying Core Debris............................................................
20 5.5 Acrosol Deposit ion.....................................................................
20 6
References................................................................................
21 TABLES 1.1 Release Phases of a Severe Accident......................................................
2 3.1 BWR Soun:e hrm Contributing Sequences...............................................
5 3.2 PWR Source Erm Contributing Sequences................................................
6 3.3 Contribution of LOCAs to Core Damage Frequency (CDF)-Internal Events.................
7 3.4 In. Vessel Release Duration for PWR Sequences...........................................
8 3.5 In-Vessel Release Duration for BWR Sequences...........................................
9 3.6 Release Phase Durations for PWRs and BWRs............................................
9 3.7 S'ICP Radion uclid e G roups.............................................................
10 3.8 Revised Radionuclide Groups............................................................
10 3.9 Fraction of mean core damage frequency with high, intermediate, and low pressure sequences (internal events only unless otherwise noted)..............................................
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i CONTENTS (Cont'd)
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1%t 3.10 Mean % lues of Radionuclides Into Containment for BWRs, Low RCS Pressure, High Zirconium Oxidation.............................................
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-3.11 Mean Values of Radionuclide Releases Into Containment fr r PWRs, Low RCS Pressure, High Zirconiu m Oxidation...............................................................
11 3.12 BWR Releases Into Containment........................................................
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3.13 PWR Releases Into Containment........................................................
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4.1 Measures of Iew Volatile In-Vessel Release Fractions......................................
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t APPENDICES e
r A Uncertainty Dist ributions......................................................................
24 B STCP Bounding Value Releases................................................................
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i NUREG-1465 vi
Preface In 1%2, the Atomic Energy Commission issued provide a postulated fission product source term Tbchnical Information Document (TID) 14844, released into containment that is based on current "CcIculation of Distance Factors for Power and 7bst understanding of LWR accidents and fission product Reactors." In this document, a release of fission behavior. The information contained in this document products from the core of a light-water reactor (LWR) is applicable to LWR designs and is intended to form into the containment atmosphere (" source term") was the basis for the development of regulatory guidance, l
postulated for the purpose of calculating off-site doses primarily for future LWRs. His report will serve as a in accordance with 10 CFR Part 100, " Reactor Site basis for possible changes to regulatory requirements.
Criteria." The source term postulated an accident that However, acceptance of any proposed changes will be resulted in substantial meltdown of the core, and the on a case-by-case basis.
fission products assumed released into the containment were based on an understanding at that time of fission Source terms for future reactors may differ from those product behavior. In addition to site suitability, the presented in this report which are based upon insights regulatory applications of this source term (in derived from current generation light-water reactors.
conjunction with the dose calculation methodology)
An applicant may propose changes in source term rJfect the design of a wide range of plant systems.
parameters (timing, release nagnitude, and chemical form) from those contained in this report, based upon In the past 30 years, substantial information has been developed updating our knowledge about severe LWR accidents and the resulting behavior of the released fission products. The purpose of this document is to i
vii NUREG-1465 l
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1 INTRODUCTION AND BACKGROUND (91%) in elemental (I ) form, with 5% assumed to be 1.1 Regulatory Use of Source Terms 2
particulate iodine and 4% assumed to be in organic He use of postulated accidental releases of radioactive form. These assumptions have significantly affected the i
materials is deeply embedded in the regulatory policy design of engineered safety features. Containment and practices of the U.S. Nuclear Regulatory isolation valve closure times have also been affected by Commission (NRC). For over 30 years, the NRC's these assumptions.
reactor site criteria in 10 CFR Part 100 (Ref.1) have required, for licensing purposes, that an accidental Use of the TID-14844 release has not been confined to f
fission product release resultmg from " substantial an evaluation of site suitability and plant mitigation t
meltdown" of the core mto the containment be features such as sprays and filtration systems. He i
postulated to occur and that its potential radiological regulatory applications of this release are wide, consequences be evaluated assummg that the contam-neluding the basis for (1) the post-accident radiation ment remains intact but leaks at its maximum allowable environment for which safety-related equipment should i
leak rate. Radioactive material escaping from the be qualified, (2) post-accident habitability requirements containment is often referred to as the " radiological for the control room, and (3) post-accident samplirg release to the emironment." He radiological release systems and accessibility.
is obtamed from the containment leak mte and a knowledge of the airborne radioactive inventory in the In contrast to the TID-14844 sourpe term and containment atmosphere. The radioactive inventory c ntamment leakage release used for design basis within containment is referred to as the accidents, severe accident releases to the emironment "in-containment accident source term."
first arose in probabilistic risk assessments (e.g.,
The expression "in-containment accident source term,"
Reactor Safety Study, WASH-1400 (Ref. 5)) in as used in this document, denotes the radioactive exammmg accident sequences that m, yolved core melt and containmems that could fail. Severe accident l
material composition and magnitude, as well as the chemical and physical properties of the material within releases represent mechanistically determined best the containment that are available for leakage from the estimate releases to the environment, mcluding reactor to the environment. The "in-containment estimates of failures of containment integrity. His is accident source term" will normally be a function of very different from the combination of the non-time and will involve consideration of fission products mechanistic release to containment postulated by being released from the core into the containment as TID-14844 coupled with the assumption of very limited well as removal of fission products by plant features containment leakage used for Part 100 siting calcula-intended to do so (e.g., spray systems) or by natural tions for design basis accidents. The worst severe i
accident releases resulting from containment failure or removal processes.
containment bypass can lead to consequences that are For currently licensed plants, the characteristics of the much greater than those associated with a TID-14844 source term released into containment where the fission product release from the core into the containment are set forth in Regulatory Guides 1.3 and containment is assumed to be leaking at its maximum 1.4 (Refs. 2.3) and have been derived from the 1%2 leak rate for its design conditions. Indeed, some of the most severe releases arise from some containment report, TID-14844 (Ref. 4). His release consists of 100% of the core inventory of noble gases and 50% of bypass events, such as rupture of multiple steam l
the iodines (half of which are assumed to deposit on generator tubes.
interior surfaces very rapidly). These values were based largely on experiments performed in the late 1950s Although severe accident sour"e terms have not been invohing heated irradiated UO2 Pellets. TID-14844 used in individual plant licensing safety evaluations, also included 1% of the remaining solid fission they have had significant regulatory applications.
products, but these were dropped from consi-Source terms from severe accidents (beyond-design-deration in Regulatory Guides 1.3 and 1.4. He 1% of basis accidents) came into regulatory consideration and the solid fission products are considercd in certain usage shortly after the issuance of WASH-1400 in 1975, areas such as equipment qualification, and their application was accelerated after the Bree l
Mile Island accident in March 1979. Current Regulatory Guides 1.3 and 1.4 (Refs. 2 and 3) specify applications rely to a large extent on the results of that the source term within containment is assumed to WASH-1400 and inchde (1) part of the basis for the be instantaneously available for release and that the sizes of emergency planning zones for all plants, (2) the 1
iodine chemical form is assumed to be predominantly basis for staff assessments of severe accident risk in f
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plant environmental impact statements, and (3) part of Improved modeling of severe accident phenomena, the basis for staff prioritization and resolution of including fission product transport, has been provided j
generic safety issues, unresolved safety issues, and by the recently developed MELCOR (Ref.14) code. At j
other regulatory analyses. Source term assessments this time, however, an insufficient body of calculations based on WASii-1400 methodology appear in many is available to provide detailed insights from this model.
probabilistic risk assessment studies performed to date.
Using analyses based on the STCP and MELCOR l
codes and NUREG-1150, the NRC has sponsored 1.2 Research Insights Since studies (Refs.15-17) that analyzed the timing, TID-14844 magnitude, and duration of fission product releases. In addition, an examination and assessment of the Source term estimates under severe accident conditions chemical form of iodine likely to be found within became of great interest shortly after the Three-Mile containment as a result of a severe accident has also i
Island (Bil) accident when it was observed that only been carried out (Ref.18).
I relatively small amounts of iodine were released to the environment compared with the amount predicted to In contrast to the instantaneous releases that were be released in licensing calculations. This led a number postulated in Regulatory Guides 1.3 and 1.4, analyses of l
of observers to claim that severe accident releases were severe accident sequences have shown that, despite much lower than previously estimated.
differences in plant design and accident sequence, such releases can be generally categorized in terms of phenomenological phases associated with the degree of Re NRC began a major research effort about 1981 to uel me ng an r cad n, reactor pressure vessel i
obtain a better understanding of fission-product integrity, and, as applicable, attack upon concrete below r
transport and release mechanisms in LWRs under the reactor cavity by molten core materials. The severe accident conditions, his research effort has included extensive NRC staff and contractor efforts general phases, or progression, of a severe LWR accident are shown m *Ihble 1.1.
involving a number of national laboratories as well as nuclear industry groups. Rose cooperative research Table 1.1 Release Phases of a Severe Accident activities resulted in the development and application of a group of computer codes known as the Source Release Phases i
1brm Code Package (STCP)(Ref. 6) to examine core-melt progression and fission product release and Coolant Activit7 Release transport in LWRs. De NRC staff has also sponsored t
significant review efforts by peer reviewers, foreign Gap Activity Release i
partners in NRC research programs, industry groups, Early In-Vessel Release and the general public. He STCP methodology for Ex-Vessel Release severe accident source terms has also been reflected in Late In-Vessel Release NUREG-1150 (Ref. 7), which provides an 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 system.
As a result of the NRC's research effort to obtain a Assuming that the coolant loss cannot be accommo-i better understanding cf fission product transport and dated by the reactor coolant makeup systems or the release mechanisms in LWRs under severe accident emergency core cooling systems, fuel cladding failure conditions, the SKP emerged as an integral tool for would occur with a release of the activity located in the i
analysis of fission product transport in the reactor gap between the fuel pellet and the fuel cladding.
coolant system (RCS) and containment. The STCP i
models release from the fuel with CORSOR (Ref. 8)
As the accident progresses, fuel degradation begins, and fission product retention and transport in the RCS resulting in a loss of fuel geometry accompanied by eith 'IRAPMELT (Ref. 9). Releases from core-concrete gradual melting and slumping of core materials to the interactions are modeled using the VANESA and bottom of the reactor pressure vessel. During this CORCON (Ref.10) codes. Depending upon the period, the early in-vessel release phase, virtually all containment type, SPARC or ICEDF (Refs.11,12) are the noble gases and significant fractions of the volatile used in conjunction with NAUA (Ref.13) to model the nuclides such as iodine and cesium are released into transport and retention of fission product releases from containment. De amounts of volatile nuclides released the RCS and from core-concrete interactions into the into containment during the early in-vessel phase are containment, with subsequent release of fission strongly influenced by the residence time of the products to the environment consistent with the state radioactive material within the RCS during core i
of the containment.
degradation.11igh pressure sequences result in long NUREG-1465 2
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residence times and significant retention and plateout airborne activity already within containment. Large of volatile nuclides within the RCS, while low pressure scale steam explosions, on the other hand, could result sequences result in relatively short residence times and in significant increases in airborne activity, but are little retention within the RCS and consequently higher much less likely to occur. In any event, releases of releases into containment.
particulates or vapors during steam explosions will also be accompanied by large amounts of water droplets, if failure of the bottom head of the reactor pressure which would tend to quickly sweep released material vessel occurs, two additional release phases may occur.
fr m the atmosphere.
Molten core debris released from the reactor pressure vessel into the containment will interact with the 2 OBJECTIVES AND SCOPE concrete structural materials of the cavity below the reactor (ex-vessel release phase). As a result of these 2.1 General interactions, quantitics of the less volatile nuclides may be released into containment. Ex-vessel releases are The primary objective of this report is to define a influenced somewhat by the type of concrete in the revised accident source term for regulatory application reactor cavity. Limestone concrete decomposes to for future LWRs. Ihe intent is to capture the major produce greater quantities of CO and CO gases than relevant insights available from recent severe axident 2
basaltic concrete. These gases may, in turn, sparge research on the phenomenology of fission product some of the less volatile nuclides, such as barium and release and transport behavior. The revised source strontium, and small fractions of the lanthanides into term is expressed in terms of times and rates of the containment atmosphere. Large quantities of appearance of radioactive fission products into the non-radioactive acrosols may also be released as a containment, the types and quantities of the species result of core-concrete interactions. The presence of released, and other important attributes such as the water in the reactor cavity overlying any core debris can chemical forms of iodine. This mechanistic approach significantly reduce the ex-vessel releases (both will therefore present, for regulatory purposes, a more radioactive and non-radioactive)into the containment, realistic portrayal of the amount of fis :.7 products either by cooling the core debris, or at least by present in the containment from a postulated severe scrubbing the releases and retaining a large fraction in accident.
the water. The degree of scrubbing will depend, of course, upon the depth and temperature of any water 2.2 Accidents Considered overlying the core debris. Simultaneously, and In order to determine accident source terms for generally with a longer duration, late in-vessel releases of some of the volatile nuclides, which had deposited in regulatory purposes, a range of severe accidents that the reactor coolant system during the in-vessel phase, have been analyzed for LWR plants was examined.
Evaluation of a range of severe accident sequences was will also occur and be released into containment.
based upon work done m support of NUREG-1150 (Ref. 7). This work is documented in NUREG/CR-5747 Wo other phenomena that affect the release of fission (Ref.17) and employed the integrated Source 7brm products into containment could also occur, as Code Package (STCP) computer codes, together with dircussed in Reference 7. The first of these is referred insights from the MELCOR code, which were used to to as "high pressure melt ejection"(llPME). If the analyze specific accident sequences of interest to RCS is at high pressure at the time of failure of the provide the accident chronology as well as detailed bottom head of the reactor pressure vessel, quantities estimates of fission product behavior within the reactor of molten core materials could be injected into the coolant system and the other pertinent parts of the containment at high vek) cities. In addition to a plant. The sequences studied progressed to a complete potentially rapid rise in containment temperature, a core melt, involving failure of the reactor pressure significant amount of radioactive material could also be vessel and including core-concrete interactions, as well.
added to the containment atmosphere, primarily in the form of aerosols. The occurrence of IIPME is A key decision to be made in defining an accident precluded at low RCS pressures. A second source term is the severity of the accident or group of phenomenon that could affect the release of fission accidents to be considered. Footnote 1 to 10 CFR Part products into containment is a possible steam explosion 100 (Ref.1), in referring to the postulated fission as a result of interactions between molten core debris product release to be used for evaluating sites, notes and water. This could lead to fine fragmentation of that "Such accidents have generally been assumed to some portion of the molten core debris with an increase result in substantial meltdown of the core with in the amount of airborne fission products. While small subsequent release of appreciable quantities of fission scale steam explosions are considered quite likely to products." Possible choices range from (1) slight fuel occur, they will not result in significant increases in the damage accidents invohing releases into containment 3
l of a small fraction of the volatile nuclides such as the 2.3 Limitations noble gases, (2) severe core damage accidents involving major fuel damage but without reactor vesse! failure or ne accident source terms defined in this report have core-concrete interactions (similar in severity to the been derived from examination of a set of severe TM1 accident), or (3) complete core-melt events with accident sequences for LWRs of current design.
i core-concrete interactions. Here outcomes are not Because of general similarities in plant and core design equally probable. Since many reactor systems must fail parameters, these results are also considered to be j
for core degradation with reactor vessel failure to occur applicable to evolutionary LWR designs such as cnd core-concrete interactions to occur, one or more General Electric's Advanced Boiling Water Reactor systems may be returned to an operable status before (ABWR) and Combustion Engineering's (CE) System core melt commences. Hence, past opmtional and 80 +.
accident experience together with information on modern plant designs, together with a vigorous program Currently, the NRC staff is reviewing reactor designs aimed at developing accident management procedures, for several smaller LWRs employing some passive indicate that complete core-melt events resulting in features for core cooling and containment heat reactor pressure vessel failure are considerably less removal. While the " passive" plants are generally likely to occur than those invohing major fuel damage similar to present LWRs, they are expected to have eithout reactor pressure vessel failure, and that these, somewhat lower core power densities than those of in turn, are less likely to occur than those involving current LWRs. Hence, an accident for the passive slight fuel damage, plants similar to those used in this study would likely extend over a longer time span. For this reason, the timing and duration values provided in the release For completeness, this report displays the mean or tables given in Section 3.3 are probably shorter than average release fractions for all the release phases those applicable to the passive plants. The release cssociated with a complete core melt. Ilowever, it is fractions shown may also be overestimated somewhat concluded that any source term selected for a particular for high pressure sequences associated with the passive regulatory application should appropriately reflect the plants, since longer times for accident progression likelihood associated with its occurrence.
would also allow for enhanced retention of fission products in the primary coolant system during core heatup and degradation. Despite the lack of specific it is important to emphasize that the release fractions sequence jn ese p gns, &
a rma n r for the source terms presented in this report are n a nmen accdent surce tem pmMelow intended to be representative or typical, rather than may be considered generally applicable to the " passive,,
conservative or bounding values, of those associated E"'
with a low ; tessure core-melt accident, except for the initial appearance of fission products from failed fuel, The accident source terms provided in this report are which was chosen conservatively. The release fractions not considered applicable to reactor designs that are are not intended to envelope all potential severe very different from LWRs, such as high-temperature aandent sequences, nor to represent any single gas-cooled reactors or liquid-metal reactors.
sequence, smcc accident sequences yielding both higher as well as lower release fractions were examined and Recent information has indicated that high burnup fuel, factored into the final report presented here.
that is, fuel irradiated at levels in excess of about 40 GWD/MTU, may be more prone to failure during Source terms for future reactors may differ from those design basis reactivity insertion accidents (RIA) than
)
presented in this report which are based upon insights previously thought. Preliminary indications are that j
derived from current generation light-water reactors.
high burnup fuel also may be in a highly fragmented or An applicant may propose changes in source term powdered form, so that failure of the cladding could i
parameters (timing, release magnitude, and chemical result in a significant fraction of the fuel itself being form) from those contained in this report, based upon released. In contrast, the source term contained in this end justified by design specific features.
report is based upon fuel behavior results obtained at lower burnup levels where the fuel pellet remains intact upon cladding failure, resulting in a release only De NRC staff also intends to allow credit for removal of those fission product gases residing in the gap or reduction of fission products within containment via between the fuel pellet and the cladding. Because of engineered features provided for fission product this recent information regarding high burnup fuels, the reduction such as sprays or filters, as well as by natural NRC staff cautions that, until further information processes such as aerosol deposition. These are indicates otherwise, the source term in this report t
discussed in Sectiou 5.
(particularly gap activity) may not be applicable for fuel NUREG-1465 4
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1 irradiated to high burnup levels (in excess of about 40 considered to significantly impact the source term are
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GWD/MTU).
summarized in Thble 3.1 for BWRs and Thble 3.2 for j
PWRs.
3 ACCIDENT SOURCE TERMS 3.2 Onset of Fission Product Release I
The expression "in-containment source terms," as used This section discusses the assumptions used in selecting in this report, denotes the fission product inventory the scenario appropriate for defining the early phases present in the containment at any given time dunng an of the source term (coolant activity and gap release accident. Tb cvaluate the m-containment source term phases). It was considered appropriate to base these i
during the course of an accident, the time-history of the early release phases on the design basis initiation that fission product release from the core into the could lead to earliest fuel failures.
containment must be known, as well as the effect of fission product removal mechanisms, both natural and A review of current plant final safety analysis reports engineered, to remove radioactive materials from the (FSARs) was made to identify all design basis accidents l
containment atmosphere.This section discusses the in which the licensee had identified fuel failure. For all time-history of the fission product releases into the accidents with the potential for release of radioactivity containment. Removal mechanisms are discussed in into the environment, the class of accident that had the Section 5.
shortest time until the first fuel rod failed was the I
design basis LOCA. As might be expected, the time l
3.1 Accident Sequences Reviewed until cladding failure is very sensitive to the design of the reactor, the type of accident assumed, and the fuel All the accident sequences identified in NUREG-1150 rod design. In particular, the maximum linear heat were reviewed and some additional Source Term Code generation rate, the internal fuel rod pressure, and the Package (STCP) and MELCOR calculations were stored energy in the fuel rod are significant performed. De dominant sequences which are considerations.
Table 3.1 BWR Source Term Contributing Sequences l
Plant Sequence Description Peach llottom TCl ATWS with reactor depressurized 102 ATWS with reactor pressurized TC3 TC2 with wetwell venting Till SBO with battery depletion j
TB2 TH1 with containment failure at vessel failure
[
S2El LOCA (2"), no ECCS and no ADS l
S2E2 S2El with basaltic concrete V
R11R pipe failure outside containment TBUX SBO with loss of all DC power
[
LaSalle TB SBO with late containment failure Grand Gulf TC ATWS carly containment failure fails ECCS
[
TBl SBO with battery depletion TB2 TB1 with H burn fails containment 2
TBS SBO, no ECCS but reactor depressurized TBR TBS with AC recovery after vessel failure SBO Station Blackout LOCA Loss of Coolant Accident i
RCP Reactor Coolant Pump RIIR Residual IIcat Removal ADS Automatic Depressurization System ATWS Anticipatedihmsient Without Scram i
n p
4 Table 3.2 PWR Source Term Contributing Sequences
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Plant Sequence Dewrlption Surry AG LOCA (hot leg), no containment heat removal systems TMLB' LOOP, no PCS and no AFWS V
Interfacing system LOCA S3B SBO with RCP seal WCA S2D-8 SBLOCA, no ECCS and Ha combustion S2D-p SBLOCA with 6" hole in containment Zion S2DCR LOCA (2"), no ECCS no CSRS S2DCF1 LOCA RCP seal, no ECCS, no containment sprays, no coolers-H burn or DCH fails containment 2
S2DCF1 except late 11 or overpressure failure of S2DCF2 2
containment TMLU Transient, no PCS, no ECCS, no AFWS-DCH fails containment Oconec 3 TMLB' SBO, no active ESF systems SIDCF LOCA (3"), no ESF systems Sequoyah S3HF1 LOCA RCP, no ECCS, no CSRS with reactor cavity flooded S3HF2 S3HF1 with hot leg induced LOCA 3HF3 S3HF1 with dry reactor cavity S3B LOCA (X") with SBO TBA SBO induces hot leg LOCA-hydrogen burn fails containment ACD LOCA (hot leg), no ECCS no CS S3B1 SBO delayed 4 RCP seal failures, only steam driven AFW operates S3HF LOCA (RCP seal), no ECCS, no CSRS S3H LOCA (RCP seal) no ECC recirculation SBO Station Blackout LOCA Loss of Coolant Accident RCP Reactor Coolant Pump DCH Direct Containment Heating PCS Power Conversion System ESF Engineered Safety Feature CS Containment Spmy CSRS CS Recirculation System ATWS Anticipated Transient Without Scram LOOP less of Offsite Power The details of the specific accident seguences are documented in NUREG/CR-5747, Estimate of Radionuclide Release Characteristics mto Containment Under Severe Accident Conditions (Ref.17).
Tb determine whether a design basis LOCA was a LOCA is considered a reasonable initiator to assume reasonable scenario upon which to base the timing of for modeling the earliest appearance of the gap activity initial fission product release into the containment, if the plant has not been approved for leak before t
various PRAs were reviewed to determine the break (LBB) operation. For plants that have received contribution to core damage frequency (CDF) resulting LBB approval, a small WCA (6" line break) would from LOCAs. 'Ihis information is shown in Thble 3.3.
more appropriately model the timing. For BWRs, large As can be seen from this table. LOCAs are a small LOCAs may not be an appropriate scenario for gap contributor to CDF for BWRs, but can be a substantial activity timing. However, since the time to initial fuel i
contributor for PWRs. Therefore, for PWRs a large rod failure is long for BWRs, even for large LOCAs,
)
l
b a
i i
i use of the large LOCA scenario should not unduly performed to identify the size of the LOCA that
{
penalize BWRs and will maintain consiaency with the resulted in the shortest fuel rod failure time (Ref.15).
assumptions for the PWR. As with the PWR, for an in both cases, the accident was a double-ended LBB approved plant, the timing associated with a small guillotine rupture of the cold leg pipe.The minimum LOCA (6" line break) would be more appropriate.
time from the time of accident initiation until first fuel rod failure was calculated to be 13 and 24.6 seconds for
- In order to provide a realistic estimate of the shortest the B&W and E plants, respectively. A sensitivity time for fuel rod failure for the LOCA, calculations study was performed to determme the effect of tripping were performed using the FRAPCON2, or not trippmg the reactor coolant pumps.The results SCDAP/RELAPS MOD 3.0, and FRAFIE computer mdicated that tripping of the reactor coolant pumps codes for two plants. The two plants were a Babcock had no appreciable impact on timmg. For a 6-tnch lm, e l
and Wilcox (B&W) plant with a 15 by 15 fuel rod array break, the time until first fuel rod failure is expected to l
- and a Westinghouse 4-loop (E) plant with a 17 by 17 be greater than 6.5 and 10 minutes, respectively.
fuel rod array. For each plant, a sensitivity study was Table 3.3 Contribution of LOCAs to Core Damage Frequency (CDF)-Internal Events Percent of CDF Percent of CDF caused by large LOCAs Boiling Water Reactors caused by LOCAs
( > 6" line break) j Peach Bottom (NUREG-1150) 3.5 1.0 Grand Gulf (NUREG-1150) 0.1 0.03 Millstone 1 (Utility) 23 13 Pressurized Water Reactors Surry (NUREG-1150) 15 4.3 Sequoyah (NUREG-1150) 63 4.6 Zion (NUREG-1150) 87 1.4 i
Calvert Cliffs (IREP) 21
<1 Oconec-3 (EPRI/NSAC) 43 3.0 i
A comparison calculation was donc using the TRAC-Source terms for future reactors may differ from those PF1 MOD 1 code, version 14.3U50.LG on the E presented in this report which are based upon insights plant. His analysis indicated that the first fuel rod derived from current generation light-water reactors.
failure would occur 34.9 seconds after pipe rupture, in An applicant may propose changes in source term contrast to the value of 24.6 seconds calculated using parameters (timing, release magnitude, and chemical SCDAP/RELAP. The reasons for the difference form) from those contained in this report, based upon between the SCDAP/RELAP5 MOD 3.0 and and justified by design specific features.
TRAC-PF1 MOD 1 are discussed in Reference 15.
3.3 Duration of Release Phases He review of the FSARs for BWRs indicates that fuel i
Section 1.2 provided a qualitative discussion of the failures may occur significantly later, on the order of release phases of an accident. This section provides i
several mmutes or more. No calculations have been estimated durations for these release phases.
performed using the aforementioned suite of codes.
The coolant activity phase begins with a postulated pipe For determining the time of appearance of gap activity rupture and ends when the first fuel rod has been in the containment (i.e., initial fuel failure), which estimated to fail. During this phase, the activity corresponds to the duration of the coolant activity released to the containment atmosphere is that phase and the beginning of the gap activity phase, it associated with very small amounts of radioactivity would be appropriate to perform a plant specific dissolved in the coolant itself. As discussed in Section calculation using the codes described above. However, 3.2 above, this phase is estimated to last about 25 if no plant specific calculations are performed, the seconds for Westinghouse PWRs, and about 13 seconds minimum times discussed above may be used to provide for B&W PWRs, assuming a large break LOCA. For a an estimate of the earliest time to fuel rod failure.
smaller LOCA (e.g., a 6-inch line break), such as would 7
be considered for a plant that has received LBB than about 30 minutes and 60 minutes for PWRs and approval, the coolant activity phase duration would be BWRs, respectively, after the onset of the accident.
expected to be at least 10 minutes. Although not However, more recent calculations (Ref.19) for the specifically evaluated at this time, Combustion Peach Bottom plant using the MELCOR code Engineering (CE) PWRs would be expected to have indicated that the durations of the gap release for three coolant activity durations similar to Westinghouse BWR accident sequences were about 30 minutes, as plants. For BWRs, the coolant activity phase would be well. On this basis, the duration of the gap activity expected to last longer; however, unless plant specific release phase has been selected to be 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, for calculations are made, the durations discussed above both PWRs and BWRs.
l are considered applicable.
During the early in. vessel release phase, the fuel as The gap activity release phase begins when fuel well as other structural materials in the core reach cladding failure commences. This phase involves the sufficiently high temperatures that the reactor core release of that radioactivity that has collected in the geometry is no longer maintained and fuel and other gap between the fuel pellet and cladding. This process materials melt and relocate to the bottom of the releases to containment a few percent of the total reactor pressure vessel. During this phase, significant inventory of the more volatile radionuclides, quantities of the volatile nuclides in the core inventory particularly noble gases, iodine, and cesium. During this as well as small fractions of the less volatile nuclides phase, the bulk of the fission products continue to be are estimated to be released into containment.This retained in the fuel itself. The gap activity phase ends release phase ends when the bottom head of the when the fuel pellet bulk temperature has been raised reactor pressure vessel fails, allowing molten core sufficiently that significant amounts of fission products debris to fall onto the concrete below the reactor can no longer be retained in the fuel. As noted in pressure vessel. Release durations for this phase vary Reference 16, a review of STCP calculated results for depending on both the reactor type and the accident six reference plants, PWRs as well as BWRs, indicated sequence. Thbles 3.4 and 3.5, based on results from that significant fission product releases from the bulk of Reference 16, show the estimated duration times for the fuel itself were estimated to commence no earlier PWRs and BWRs, respectively.
Table 3.4 In. Vessel Release Duration for PWR Sequences Release Duration Plant Accident Sequence *
(Min)
Surry HiLB' (H) 41 Surry S3B (H) 36 Surry AG (L) 215 Surry V
(L) 104 Zion HILU (H) 41 Zion S2DCR/S2DCF (H) 39 Sequoyah S3HF/S3B (H) 46 Sequoyah S3B1 (H) 75 Sequoyah TMLB' (H) 37 Sequoyah TBA (L) 195 Sequoyah ACD (L) 73 Oconee TMLB' (H) 35 Oconce SIDCF (L) 84
- (H or L) Denotes whether the accident occurs at high or low pressure.
Based on the information in these tables, the staff release phase have 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 concludes that the in-vessel release phase is somewhat 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, for PWR and BWR plants respectively, as longer for BWR plants than for PWR plants. His is recommended by Reference 17.
l largely due to the lower core power density in BWR plants that extends the time for complete core melt.
The ex-vessel release phase begins when molten core Representative times for the duration of the in-vessel debris exits the reactor pressure vessel and ends when NUREG-1465 8
s 5
i Table 3.5 In. Vessel Release Duration for BWR Sequences
- Plant..
Accident Sequence *.
Release Duration (Min).
'I Peach Bottom TC2 (H) 66
' Peach Bottom TC3 68 Peach Bottom TC1 (L).
97 Peach Bottom TB1/rB2 (H) 91 Peach Bottom '
V (L) 69
' Peach Bottom S2E (H) 81 Peach Bottom TBUX (H) 67 LaSalle TB (H) 81 Grand Gulf TB (H) 122
)
Grand Gulf TC1 (L) 130 Grand Gulf TBS /rBR (L)
.y
- (H or L) denotes whether the accident occurs at high or low pressure.
the debris has coole<:1 sufficiently that significant release 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 />. This value I
quantities of fission products are no longer being has been selected for this report.
released. During this phase, significant quantities of the l
volatile radionuclides not already released during the A summary of the release phases and the selected early in-vessel phase as well as lesser quantities of duration times for PWRs and BWRs is shown for non-volatile radionuclides are released into reference purposes in 'Ihble 3.6.
containment. Although releases from core-concrete interactions are predicted to take place over a number Table 3.6 of hours after vessel breach, Reference 16 indicates Release Phase Durations for PWRs and BWRs that the bulk of the fission products (about 90%), with the exception of tellurium and ruthenium, are expected
- Duration,
- Duration, to be released over a 2-hour period for PWRs and a PWRs BWRs 3-hour period for BWRs. For tellurium and ruthenium, Release Phase (Hours)
(Hours) t ex-vessel 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, for PWRs and BWRs. The difference in Coolant Activity 10 to 30 seconds
- 30 seconds
- duration of the ex-vessel phase between PWRs and Gap Activity 0.5 0.5 BWRs is largely attributable to the larger amount of zirconium in BWRs, which provides additional chemical Early In-Vessel 1.3 1.5 energy of oxidation. Based on Reference 17, the Ex-Vessel 2
3 ex-vessel release phase duration is taken to be 2 and 3 Late In-Vessel 10 10 t
hours, respectively, for PWRs and BWRs.
- Without approval for leak-before-break. Coolant activity phase duration is assumed to be 10 minutes The late in-vessel release phase commences at vessel with leak-before-break approval, breach and proceeds simultaneously with the occurrence of the ex-vessel phase. However, the 3.4 Fission Product Cornposition and duration is not the same for both phases. During this 1 Magnitude release phase, some of the volatile nuclides deposited within the reactor coolant system earlier during core In considering severe accidents in which the contain-degradation and melting may re-volatilize and be ment might fail, WASH-1400 (Ref. 5) examined the released into containment. Reference 17, after a review spectrum of fission products and grouped 54 radionu-of the source term uncertainty methodology used in clides into 7 major groups on the basis of similarity in I
NUREG-1150 (Ref. 7), estimates the late in-vessel chemical behavior. Tite effort associated with the S'ICP 9
l further analyzed these groupings and expanded the 7 Similarly, low pressure sequences cause aerosols fission product groups into 9 groups. Dese are shown Senerated within the RCS to be swept out rapidly in 1hble 3.7.
without significant retention within the RCS, thereby resulting in higher release fractions from the core into Table 3.7 STCP Radionuclide Groups Group Elements
. Table 3.8 Revised Radionuclide Groups 1
Xe,Kr Group Title Elements in Group 2
I, Br 1
Noble gases Xc,Kr 3
Cs, Rb 2
Halogens I, Br 4
Tb,Sb,Se 3
Alkali Metals Cs,Rb 5
Sr 4
lbilurium group lb,Sb,Se 6
Ru, Rh, Pd, Mo Tb 5
La, Zr, Nd, Eu, Nb, Pm, Pr, Sm, Y 6
Noble Metals Ru, Rh, Pd, Mo,1b, Co 8
Ce,Pu,Np 7
Lanthanides La, Zr, Nd, Eu, Nb, Pm, 9
Ba Pr, Sm, Y, Cm, Am 8
Cerium group Ce,Pu,Np Both the results of the STCP analyses and the uncertainty analysis (using the results of the NUREG-1150 source term expert panel elicitation)
The relative frequency of occurrence of high vs. Iow reported in NUREG/CR-5747 (Ref.17) indicate only pressure sequences were examined for both BWRs and minor differences between Ba and Sr releases. Hence, PWRs. The results of this survey are shown in a revised grouping of radionuclides has been developed Thble 3.9, and they indicate that a significant fraction of that groups Ba and Sr together. The relative the sequences examined, in terms of frequency, importance to offsite health and economic occurred at low pressure. In addition, advanced PWR consequences of the radioactive elements in a nuclear designs are increasingly incorporating safety-grade reactor core has been examined and documented m depressurization systems, primarily to minimize the NUREG/CR-4467 (Ref. 20). In addition to the likelihood of high pressure melt ejection (HPME) with elements already included in 1hble 3.7, Reference 20 its associated high containment atmosphere heat kiads found that other elements such as Curium could be and large amounts of atmospheric aerosols.
important for radiological consequences if released in
+
sufficiently large quantities. For this reason, group 7 has been revised to include Curium (Cm) and For these reasons, the composition and magnitude of Americium (Am), while group 6 has been revised to the source term has been chosen to be representative include Cobalt (Co). The revised radionuclide groups of conditions associated with low pressure in the RCS used in this report including revised titles and the at the time of reactor core degradation and pressure elements comprising each group are shown in Thble 3.8.
vessel failure. Reference 17 provides estimates of the mean core fractions released into < antainment, as estimated by NUREG-1150 (Ref. 7), for accident Source term releases into the containment were sequences occurring under low RCS pressure and high evaluated by reactor type, i.e., BWR or PWR, from the zirconium oxidation conditions. Hese are shown in sequences in NUREG-1150 and the supplemental Thbles 3.10 and 3.11.
STCP calculations discussed in Section 3.1.
Releases into containment during the early in-vessel 3.5 Chemical Form phase, prior to reactor pressure vessel failure, are markedly affected by retention in the RCS, which is a The chemical form of iodine and its subsequent function of the residence time in the RCS during core behavior after entering containment from the reactor degradation. High pressure in the RCS during core coolant system have been documented in degradation allows for longer residence time of NUREG/CR-5732, Iodine Chemical Forms in LWR terosols released from the core. This, in turn, permits Severe Accidents (Ref.18) and in ORNL/FM-12202, increased retention of aerosols within the RCS and "Models of Iodine Behavior in Reactor Containments,"
lower releases from the core into the containment.
(Ref. 21).
NUREG-1465 10
Table 3.9 Fraction of mean core damage frequency with high, intermediate, and low pressure sequences (internal events only unless otherwise noted)
Im Pressure at High Pressure at Intermed. press.
Vessel Breach Bolling Water Reactors Vessel Breach at vessel breach
(<200 psi)
No vessel breach LaSalle-external events only 0.27 N/A 0.67 0.06 LaSalle-internal events only 0.19 N/A 0.62 0.19.
Grand Gulf 0.28 N/A 0.51 0.21 Peach Bottom 0.51 N/A 0.41 0.08 Pressurized Water Reactors Surry 0.06 0.07 0.37 0.50 Sequoyah 0.14 0.21 0.24 0.41 Zion 0.03 0.15 0.72 0.10 Table 3.10 Mean Values of Radionuclides Into Containment for BWRs, im RCS Pressure, High Zirconium Oxidation I
Nuclide Early In Vessel Ex. Vessel Late In vessel l
N. G.
1.0 0
0 4
I 0.27 037 0.07 Cs 0.2 0.45 0.03 lb 0.11 038 0.01 Sr 0.03 0.24 0
Ba 0.03 0.21 0
Ru 0.007 0.0M 0
La 0.002 0.01 0
Ce 0.009 0.01 0
1 Table 3.11 Mean Values of Radionuclide Releases Into Containment for PWRs, Iany RCS Pressure, High Zirconium Oxidation Nuclide Early In Vessel Ex Vessel Late In vessel N.G.
1.0 0
0 I
0.4 0.29 0.07 Cs 03 039 0.06 Te 0.15 0.29 0.025 Sr 0.03 0.12 0
Ba 0.04 0.1 0
Ru 0.008 0.004 0
La 0.002 0.015 0
Ce 0.01 0.02 0
11 NUREG-1465
)
The results from Ref.18 indicate that iodine entering values of 7 or greater within the containment, the containment is at least 95% Csl with the remaining elemental iodine can be taken as comprising no more 5% as I plus III, with not less than 1% of each as I and than 5 percent of the total iodine released, and iodine 111. Once the iodine enters containment, however, in organic form may be taken as comprising no greater additional reactions are likely to occur. In an aqueous than 0.15 percent (3 percent of 5 percent) of the total environment, as expected for LWRs, iodine is expected iodine released.
to dissolve in water pools or plate out on wet surfaces in ionic form as I. Subsequently, iodine behavior Organic iodide formation in llWRs versus PWRs is not within containment depends on the time and pH of the notably different. Reference 18 examined not only water solutions. Because of the presence of other iodine entering containment as CsI: but also considered dissolved fission products, radiolysis is expected to other reactions that might lead to volatile forms of occur and lower the plI of the water pools. Without any iodine within containment, such as reactions of CsOH pli control, the results indicate that large fmetions of with surfaces and revaporization of Csl from RCS the dissolved iodine will be converted to elemental surfaces. Reference 18 indicates (Ihble 2.4) that for the iodine and be released to the containment atmosphere.
Peach Bottom TC2 sequence, the estimated percentage liowever, if the pil is controlled and maintained at a of iodine as HI was 3.2 percent, not notably less than value of 7 or greater, very little (less than 1%) of the the PWR sequences examined. While organic iodide is dissolved iode will be converted to elemental iodine, formed largely from reactions of elemental iodine, Ref.
Some considerations in achieving pil control are 22 clearly notes that reactions with 111 may be discussed in NUREG/CR-5950, " Iodine Evolution and important.
pH Control," (Ref. 22).
Although organic iodine is not readily removed by Organic compounds of iodine, such as methyl iodide, c ntainment sprays or filter systems, it is unduly CH 1, can also be produced over time largely as a conservative to assume that orgame todme is not 3
result of elemental iodine reactions with organic rem ved at all from the contamment atmosphere, one:
materials. Organic iodide formation as a result of generated, since such an assumption can result in an reactor accidents has been surveyed in WASH-1233, verestimate of long-term doses to the thyroid.
" Review of Organic lodide Formation Under Accident References 23 and 24 discuss the radiolytic destruction Conditions in Water-Cooled Reactors," (Ref. 23), and f rg nic iodide, and Standard Review Plan Section more recently in NUREG/CR-4327, " Organic Iodide (S.R.P.) 6.5.2 notes the above reference and indicates Formation Following Nuclear Reactor Accidents,"
that removal of orgame iodide may be considered on a (Ref.24). From an analysis of a number of containment case-by-case basis. A rational model for organic iodine e avmr within contamment would consider both its experiments, WASH-1233 concluded that, considering, both non-radiolytic as well as radiolytic means, no more f rmation as well as destruction in a time. dependent than 3.2 percent of the airborne iodine would be fashion. Development of such a model, however, is converted to organic iodides during the first two hours beyond the scope of the present report.
following a fission product release. 'Ite value of 3.2 Clearly, where the pil is not controlled to values of 7 percent was noted as a conservative upper limit and was judged to be considerably less, smce it did not account, or greater, significantly larger fractions of elemental among other things, for decreased radiolytic formation iodine, as well as organic iodine may be expected within containment.
of orgame iodide due to iodme removal mechamsms within containment. Reference 24 also included results All other fission products, except for the noble gases involving irradiated fuel elements, and concluded that and iodine, discussed above, are expected to be in the orgame iodide concentration within containment particulate form.
would be about 1 percent of the iodine release concentration over a wide range of iodine 3.6 Proposed Acc. dent Source'Ibrms concentrations.
i
'Ihe proposed accident source terms, including Geir A conversion of 4 percent of the elemental iodine to timing as well as duration, are listed in Thbles 3.12 for organic has been implicitly assumed by the NRC staff in llWRs and 3.13 for PWRs. 'Ihe infonnation for these Regulatory Guides 1.3 and 1.4, based upon an upper tables was derived from the simplification of the bound evaluation of the results in WASH-1233.
NUREG-1150 (Ref. 7) source terms documented in However, in view of the results of Ref. 23 that a NUREG/CR-5747 (Ref.17). It should also be noted conversion of 3.2 percent is unduly conservative, a that the rate of release of fission products into the value of 3 percent is considered more realistic and will containment is assumed to be constant during the be used in this report. Where the pH is controlled at duration time shown.
NUREG-1465 12
i l
l l
l' Table 3.12 BWR Releases Into Containment
- E l
Cap Release ***
Early In. Vessel Ex. Vessel Late in. Vessel Duration (Hours) 0.5 1.5 3.0
_10.0 Noble Gases **
0.05 0.95 0
0 Halogens 0.05 0.25 0.30 0.01 Alkali Metals 0.05 0.20 0.35 0.01 Tbilurium group 0
0.05 0.25 0.005 t
0.02 0.1 0
Noble Metals 0
0.0025 0.0025 0
Cerium group 0
0.0005 0.005 0
Lanthanides 0
0.0002 0.005 0
- Values shown are fractions of core inventory.
" See Thble 3.8 for a listing of the c!cments in each group
- " Gap release is 3 percent if long-term fuel cooling is mamtained.
i Table 3.13 PWR Releases Into Containment
- Gap Release ***
Early In. Vessel Ex. Vessel Late In. Vessel Juration (Hours) 0.5 1.3 2.0 10.0 I
Noble Gases" 0.05 0.95 0
0 Halogens 0.05 0.35 0.25 0.1 i
Alkali Metals 0.05 0.25 0.35 0.1 Tellurium group 0
0.05 0.25 0.005 l
0.02 0.1 0
Noble Metals 0
0.0025 0.0025 0
Cerium group 0
0.0005 0.005 0
Lanthanides 0
0.0002 0.005 0
- Values shown are fractions of core inventory.
i
" See lhble 3.8 for a listing of the elements in each group r
"* Gap release is 3 percent if long-term fuel cooling is maintained.
l It is emphasized that the release fractions for the PWRs, respectively. The changes and the reasons for source terms presented in this report are intended to these was as follows:
be representative or typical, rather than conservative or i
bounding values, of those associated with a low 1.
BWR in-vessel release fractions for the volatile pressure core-melt accident, except for the initial nuclides (I and Cs) increased slightly while appearance of fission products from failed fuel, which ex-vessel release fractions for the same nuclides was chosen conservatively.The release fractions are not was reduced as a result of comments received and intended to envelope all potential severe accident additional MELCOR calculations available after sequences, nor to represent any single sequence.
issuance of the drrit report. The total I and Cs released into containment over all phases of the accident remained the same.
'Ihbles 3.12 and 3.13 in this, the final report, were 2.
Release fractions for ib, Ba and Sr were reduced modified from the tables in the draft report which were somewhat, both for in-vessel as well as ex-vessel tal.en from 'Ihble 3.9 and 'lhble 3.10, for BWRs and releases, in response to comments.
i 13 NUREG
3.
Release fractions for the non-volatile nuclides, additional release of 2 percent over the duration particularly during the early in-vessel phase were of the gap release phase.
reduced significantly based on additional research results (Ref. 25) since issuance of NUREG-1150 3.
Accidents where fuel failure results from reactivity which indicate that releases of low volatile insertion accidents (RIA), such as the postulated nuclides, both in-vessel as well as ex-vessel, have rod ejection (PWR) or rod drop (BWR) accidents.
been overestimated. A re-examination in response The accidents examined in this report do not to comments received showed that the supposed contain information on reactivity induced "means" of the uncertainty distribution were in accidents to permit a quantitative discussion of excess of other measures of the distribution, such fission product releases from them. Hence, the as the 75th percentile. In this case, the 75th gop release magnitude presented in ihbles 3.12 percentile was selected as an appropriate mc asure and 3.13 may not be applicable to fission product of the release fraction. For additional discussion releases resulting from reactivity insertion on this topic, see Section 4.4.
accidents.
'tecent information has ir.dicated that high burnup fuel, 4.
Gap activity release fractions were reduced from 5 that is, fuel irradiated at levels m excess of about 40 percent to 3 percent for accidents not involving GWDMIU, may be more prone to failure during degraded or molten core conditions and where long-term fuel cooling is maintained. See design basis reactivity insertion accidents than previously thought. Prelunmary mdict.tions are that additional discussion below, high burnup fuel also may be in a highly fragmented or p wdered form, so that failure of the cladding could Based on WASH-1400 (Ref. 5) the inventory of fission result m a significant fraction of the fuelitself being products residing in the gap between the fuel and the released. In contrast, the source term contamed in th,s i
cladding is no greater than 3 percent except for cesium, rep rt is based upon fuel behavior results obtamed at which was estimated to be about 5 percent.
l wer burnup levels where the fuel pellet remams NUREG/CR-4881 (Ref.16) repcrted a comparison of intact upon cladding failure, resulting in a release only more recently available estimations and observations of those fission product gases residing in the gap l
indicating that releases of the dominant fission product between the fuel pellet and the cladding. Because of groups were generally below the values reported in this recent information regarding high burnup fuels, the Reference 5. However, the magnitude of fission NRC staff cautions that, until further information products released during the gap release nhase can ndicates otherwise, the source term in Thbles 3.12 and vary, dependmg upon the type of accident. Accidents 3.13 (particularly gap activity) may not be applicable for where fuel failures occur may be grouped as follows:
fuel irradiated to high burnup levels (in excess of about 1.
Accidents where long term fuel cooling is maintained despite fuel failure. Examples include With regard to the ex-vessel releases associated with the design basis LOCA where ECCS functions, core-concrete interactions, according to Reference 17, and a postulated spent fuel handling accident. For there were only slight differences in the fission this category, fuel failure is taken to result in an products released into containment between limestone immediate release, based upon References 5 and vs. basaltic concrete. Hence, the table shows the 16, of 3 percent of the volatile fission products releases only for a limestone concrete. Further, the (noble gases, iodine, and cesium) which are in the releases shown for the ex-vessel phase are assumed to gap between the fuel pellet and the cladding. No be for a dry reactor cavity having no water overlying any subsequent appreciable release from the fuel core debris. Where water covers the core debris, pellet occurs, since the fuel does not experience aerosol scrubbing will take place and reduce the prolonged high temperatures-quantity of aerosols entering the containment atmosphere. See Section 5.4 for further information.
2 Accidents where long-term fuel cooling or core geometry are not maintained. Examples include 3.7 Nonradioactive Aerosols degraded cx>re or core-melt accidents, meluding the postulated limiting design basis fission product in addition to the fission product releases into release into containment used to show compliance containment shown in Thbles 3.12 and 3.13, quantities with 10 CFR Part 100. For this category, the gap of nonradioactive or relatively low aethity aerosols will release phase may overlap to some degree with also be released into containment. Rese aerosols arise the early in-vessel release phase. The release froni core structural and control rod materials released magnitude has been taken as an initial release of 3 during the in-vessel phase and from concrete decompo-percent of the volatiles (as for category 1), plus an sition products during the ex-vessel phase. A detailed NUREG-1465 14
. =.
t analysis of the quantity of nonfission product acrosols 4.1 Accident Severity and Type i
released into containment was not undertaken. Precise i
estimates of the masses of non-radioactive acrosols As noted earlier in Section 2.2, this report discusses released into containment are difficult to determine.
mean or average release fractions for all the release phases associated with a complete core-melt accident, i
including reactor pressure vessel failure. He accident selected is one in which core melt occurs at low Reference 26 evaluated one PWR sequence (Sequoyah) pressure conditions. A low pressure core melt scenario and one llWR (Peach Bottom) sequence and calculated in-vessel non-radioactive acrosol masses of 350 and 780 mults in a yelatively low level of fission product retention within the reactor coolant system, and a kilograms, respectively, for the PWR and BWR omsequently high level of release of fission products sequences. He same reference calculated that fr m the core into contamment durmg the early ex-vessel aerosol masses (assuming a dry cavity) would in-vessel release phase. Since the bulk of the fission be higher,3800 and $600 kilograms, respectively, for pr ducts entering contamment do so during the early the PWR and BWR sequences investigated. However, m-vessel release phase, selection of a low pressure core these values, particularly for the ex-vessel release melt scenano provides a high estimate of the total phase, may be excessive. NUREG/CR-4624 (Ref. 27) quantity of ftssion products released into containment, examined several sequences for both PWRs and BWRs as well as that dunng the early m-vessel release phase.
and calculated ex-vessel releases to containment of about 1000 and 4000 kilograms, respectively, for PWRs and 13WRs. NUREG/CR-5942 (Ref.19), making use of 4.2 Onset of Fission Product Release the MELCOR code, calculated significantly lower ne onset, or earliest time of appearance of fission releases during the ex-vessel phase of about 1000 products within containment, has been selected on the kilograms for the Peach Bottom plant.
basis of the earliest time to failure of a fuel rod, given a design basis LOCA. His is estimated to be from about 13 to 25 seconds for plants that do not have leak-In view of the wide diversity of calculated results, the before-break approval for their reactor coolant system NRC staff concludes that precise estimates of the piping, and it is expected to vary depending on the release of non-radioactive acrosols are not available at reactor as well as the fuel rod design. His value, while this time. Because nonradioactive aerosol masses could representing some relaxation from the assumption of have an effect upon the operation of certain plant instantaneous appearance, is nevertheless conservative.
i equipment, such as filter loadings or sump perfor-As noted in Reference 15, these estimates are valid for mance, during and following an accident, however, the a double-ended rupture of the largest pipe, assume that NRC staff concludes that the release of non-radioactive the fuel rod is being operated at the maximum peaking aerosols should be considered by the designer using factor permitted by the plant Technical Specifications enethods considered applicable for his design, and the and at the highest burnup levels anticipated, and potentialimpact upon the plant evaluated.
assume that the emergency core cooling system (ECCS) is not operating. Use of more realistic assumptions for any of these parameters would increase estimated times t fuel r d f ilure by facton; of two or more. Neverthe-4 MARGINS AND UNCERTAINTIES less, the use of conservative assumptions in estimating fuel rod failure times is considered appropriate since His section discusses some of the more significant such failure times are likely to be used primarily in conservatisms and margins in the proposed accident consideration of the necessary closure time for certain source term given in Section 3. Briefly, the proposed contamment isolation valves. Smce it is important that closure of such valves be ensured before the release of release fractions have been developed from a complete core-melt accident, that is, assuming core melt with signifirmt radioactivity to the environment, a conserva-reactor pressure vessel failure and with the assumption tive esumate of fuel failure time and consequent onset i
of core-concrete interactions. De timing aspects were of fission product appearance is deemed appropn, ate.
selected to be typical of a low pressure core-melt For plants with leak-before-break approval for their scenario, except that the onset of the release of gap reactor coolant system pipmg, a longer du'ation before activity was based upon the earliest calculated time of fuel clad failure is expected. However, other constraints fuel rod failure under accident conditions. He rn y become the limiting factor on containment is lation valve closure time.
magnitude of the fission products released into containment was intended to be representative and, 4.3 Release Phase Durations except for the low volatile nuclides, as discussed m
(
section 4.4, was estimated from tht nean values for a ne durations of the various release phases have been typical low-pressure core-melt scenario.
selected primarily by examinatbn of the values 15 NUREG-1465 l
available for the group of severe accident scenarios examination of the Three Mile Island (TMI) accident, considered in Section 3.The durations of the early and the SASCHA out-of-pile tests. Ex-vessel insights in-vessel and ex-vessel release phases differs for BWRs derive primarily from large scale tests performed as versus PWRs and reflect the differing core heatup rates part of the internationally sponsored ACE Program.
as well as the differing amounts of zirconium available Reference 25 notes that, based on the SFD experiments to supply chemical energy after core-melt. While the as well as the TMI accident, in-vessel release fractions 4
selected durations of the release phases are realistic, for cerium, for example, were about 10, compared to the value of 10-2 ited in the draft report. Based on some conservatisms should be noted. The duration of c
the early in-vessel release phase for BWRs and PWRs these results, the NRC staff concludes that the low is short and does n;. represent a probabilistically volatile release fractions cited in draft NUREG-1465 weighted average or mean value for the accident are too high.
sequences considered. His will introduce a given quantity of fission products into containment in a The uncertainty distributions were also examined to shorter time than might be expected for a typical obtain additional insight. As can be seen from the sequence.
uncertainty distributions in Appendix A, the range of release estimates for the volatile nuclides, such as the Similarly, the duration of the ex-vessel release phase, noble gases, iodine, cesium, and to some extent while considered realistic for the bulk of the fission tellurium, spans about one order of magnitude. For this products being released, is short for releases of group of nuclides, use of the mean value is a tellurium and ruthenium since, as noted in Section 3.3, reasonable estimate of the release fraction. In contrast, release of these nuclides occurs over a longer time.
the range for the low volatile nuclides, such as barium, strontium, cerium and lanthanum, spans about 4 to 6 The selected release duration times have been chosen orders of magnitude. For the latter group of nuclides, primarily on the basis of simplicity, since an accurate the mean value can be misleading, since it may be well determination of the duration of the release phases in excess of other measures of the distribution. This is illustrated in 'Ihble 4.1 which tabulates the mean, depends not only on the reactor type but also on the applicable accident sequence, which varies for each median, and 75th percentile values for several low reactor design.
volatile nuclides released during the early in-vessel phase.
4.4 Coniposition and Magnitude of Table 4.1 Measures or tow voiatiie in.vessei Reiease Fractions 7
Releases The composition of the fission products was initially Nuclide Mean Median 75th percentile based on the grouping developed with the STCP, but has been modified as discussed in Section 3.4.
Sr 0.03 0.001 0.006 Ba 0.04 0.003 0.009 The magnitudes of the fission products released into La 0.002 0.00003 0.0003 containment for the accident source term were selected Ce 0.01 0.00006 0.0006 m the draft version of this report to be the mean values, using NUREG-1150 methodology, for UWR and PWR low-pressure scenarios involving high As can be seen from Thble 4.1, the mean value for this estimates of zirconium oxidation. He uncertainty group of nuclides is one to two orders of magnitude distributions for the in-vessel release and total release greater than the median value, and is about 5 times into containment are displayed graphically in Appen-greater than the 'i5th percentile of the distribution. For dix A. Bounding estimates for the releases into this group of nuclides, the mean is controlled by the l
containment taken from Reference 17, using the STCP upper tail of the distribution, and the details of the methodology, are shown in Appendix B.
whole distribution may be more indicative of the uncenainty than the " bottom line" results, such as a The release magnitudes for the low volatile fission mean value. Because of this, the final version of this i
products were reduced significantly in the final report, report has chosen not to use the mean value in His reduction was based upon recent experimental estimating releases for the non-volatile nuclides. While research results (Ref. 25) since completion of the median value might be selected as an alternate, it NUREG-1150, as well as a re-examination of the fails to provide an appreciation of the range of values uncertainty distribution, in response to comments on lying above it. Since this report is intended for the draft report. Research on in-vessel phenomena regulatory applications, the intent is to avoid includes the in-pile Severe Fuel Damage (SFD) under-esthnation of potential releases or offsite doses, experiments in the Iower Burst Facility (PBF), further without undue conservatism. Hence, for the final NUREG-1465 16
I i
i report, the 75th percentile value has been selected for Mean value estimates selected for the in-cx ntainment the low volatile nuclides on the basis that it bounds accident source term provide reasonable estimates for most of the range of values, without undue influence by the important nuclides consisting of iodine, cesium, and l
the upper tail of the distribution.
tellurium. Dese estimates show a relatively low degree of uncertainty and are unlikely to be exceeded by more Uncertainties, particularly in understanding and than 50%. Uncertainty increases in estimating releases modeling core melt progression phenomena, can affect for the remaining nuclides.
the duration of the early in-vessel release phase, l
including the timing of reactor pressure vessel failure.
4.5 Iodine Chemical Form An increase in duration of the early in-vessel phase can lead to increased releases of volatile fission products The chemical form of iodine entering containment was j
during the early in-vessel phase and a concomitant investigated in Reference 18. On the basis of this work, reduction during the ex-vessel phase. An increase in the NRC staff concludes that iodine entering l
duration of the early in-vessel phase, however, also containment from the reactor coolant system is provides additional time for fission product removal composed of at least 95% cesium iodide (CsI), with no within containment by natural processes or fission more than 5% I plus Hl. Once within containment, j
product cleanup systems.
highly soluble cesium iodide will readily dissolve in water pools and plate out on wet surfaces in ionic form.
Radiation-induced conversion of the ionic form to Upper bound estimates, tabulated in Appendix B, elemental iodine will potentially be an important mdicate that virtually all the todme and cesmm could mechanism. If the pH is controlled to a level of 7 or enter the containment. Similarly, for tellurium, upper greater, such conversion to elemental iodine will be bound estimates m, dicate that as much as about minimal. If the pH is not controlled, however, a two-thirds of the core inventory of tellurium could be relatively large fraction (greater for PWRs than BWRs) l released into containment. Hence, for this unportant of the iodine dissolved in containment pools in ionic group of radionuclides (iodine, cesium, and tellunum),
form will be converted to elemental iodine.
the upper bound estimates of total release m, to containment are approximately 1.5 times the mean value estimates.
5 IN-CONTAINMENT REMOVAL i
MECHANISMS For the lower volatility radionuclides such as barium and strontium, upper botmd estimates range from Since radioactive fission products within containment about 50 to 70% of the core inventory released into are in the form of gases and finely divided airborne containment. Almost all of this is estimated to be Particulates (aerosols), the principal mechanism by released as a result of core-concrete interactions. In which fission products find their way from the reactor I
contrast, mean value estimates range from 15 to 25%.
to the environment with an intact containment is via Hence, in this case, the upper bound estimates are leakage from the containment atmosphere. The specific about two to three times the mean values.
fission product inventory present in the containment atmosphere at any time depends on two factors: (1) the "E
8 "##
E Finall, for the refracto'Y nuclides such as lanthanum introduced into the containment atmosphere, and Y
and cerium, the upper bound estimates mdicate that (2) the sink, the rate at which they are being removed.
I about 5% of the inventory of these nuclides could Aspects of the release and transport of fission products t
appear withm containment, whereas the mean value from the core into the containment atmosphere were i
estimate mdicates only about 1% released.
presented in Section 3.
j PRAs have indicated that, considering the magnitudes Mechanisms that remove fission products from the
[
of the radioactive species estimated to be released to atmosphere with consequent mitigation of the the emironment for severe reactor accidents, the in-containment source term fall into two classes:
i radionuclides having the greatest impact on risk are (1) engineered safety features (ESFs) and (2) natural typically the volatile nuclides such as iodine and processes. ESFs to remove or reduce fission products i
cesium, with tellurium to a somewhat lesser degree.
within the containment are presently required ne uncertainty distributions for this group of (Criterion 41 in Appendix A of 10 CFR Part 50) and radionuclides is also the smallest, as shown in the include such systems as containment atmosphere graphical tabulations of Appendix A. Hence, our ability sprays, BWR suppression pools, and filtration systems to predict the behavior and releases for this group of utilizing both particulate filters and charcoal adsorption l
nuclides is significantly better than for other fission beds for the removal of iodine, particularly in elemen-s product groupings.
tal form. Natural removal includes such processes as 17 NUREG-1465
aerosol deposition and the sorption of vapors on containment spray systems be initiated automatically, equipment and structural surfaces.
because rf the instantaneous appearance of the source term within containment, and that the spray duration ne draft version of this report contained a discussion not be less than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. In contrast, the revised source of some of the more important fission product removal term information given in Section 3 suggests that spray mechanisms, including some quantitative results. Dese system actuation might be somewhat delayed for numerical results were intended to be illustrative of the radiological purposes, but that the spray system phenomena involved and were not intended to be duration should be for a longer period of about 10 or applied rigorously, however. It was recognized that the more hours. Because sprays are effective in rapidly data and illustrations used in the draft might not be removing particulates from the containment applicable to all situations.
atmosphere, intermittent operation over a prolonged period may also provide satisfactory mitigation.
In recognition of this, the NRC staff undertook to g
gg gg cxamine, with contractor ecsistance, improved understanding of fission product removal mechanisms.
particularly important in view of the information presented in Section 3, which indicates that most fission I
At this time, this effort is still underway. Rather than products are expected to be m. particulate form. The provide numerical values that may be inapplicable, this report will provide references, where available, so that spray removal coefficient (h) is derived from the the reader may utilize improved methodologies to following equation from Standard Review Plan Section 6.5.2 obtain results that apply to the situation at hand.
l I
" 3hFE 5.1 Containment Sprays 2vo h 8
'P Containment sprays, covered in Standard Review Plan h"
ent b hing!o ume ta (SRP) Section 6.5.2 (Ref. 28), are used in many PWR F
- Spray flow
}
designs to provide post-accident containment cooling as E/D = the ratio of a dimensionless collection l
well as to remove released radioactive aerosols. Sprays efficiency E to the average spray drop Diameter D.
j are effective in reducing the airborne concentration of E/D is conservatively assumed to be equal to elemental and particulate iodines as well as other 10/ meter for spray drops 1 mm in diameter changing
[
particulates, such as cesium, but are not effective in to 1/ meter when the aerosol mass has been removing noble gases or organic forms of iodine. He depleted by a factor of 50.
reduction in airborne radioactivity within containment by a spray system as a function of time is expressed as Using values typical for PWRs, the formulation given in i
an exponential reduction process, where the spray SRP 6.5.2 estimates particulate removal rates to be on removal coefficient, lambda, is taken to be constant the order of 5 per hour. Nourbakhsh (Ref. 29) exa-over a large part of the regime. 'Iypical PWR mined the effectiveness of containment sprays, as i
containment spray systems are capable of rapidly evaluated in NUREG-1150 (Ref. 7), in decontamin-f reducing the concentration of airborne activity (by ating both in-vessel and ex-vessel releases. Powers and about 2 orders of magnitude within about 30 minutes, Burson (Ref. 30) have developed a more realistic, yet f
where both spray trains are operable). Once the bulk of simplified, model with regard to evaluating the the activity has been removed, however, the spray effectiveness of acrosol removal by containment sprays becomes significantly less effective in reducing the i
remaining fission products. This is usually accounted 5.2 BWR Suppression Pools for by either employing a spray cut-off, wherem the spray removal becomes rero after some reduction has BWRs use pressure suppression pools to condense
[
been achieved, or changing to a much smaller value of steam resulting from a loss-of-coolant accident. Prior to t
lambda to reflect the decreased removal effectiveness the release to the reactor building, these pools also of the spray when airborne concentrations are low.
scrub radioactive fission products that accompany the steam. Regulatory Guide 1.3 (Ref. 2) suggests not SRP Section 6.5.2 (Ref. 28) provides expressions for allowing credit for fission product scrubbing by BWR calculating spray tambdas, depending on plant suppression pools, but SRP Section 6.5.5 (Ref. 31) was i
parameters as well as the type of species removed. In revised to suggest allowing such credit. De pool water t
addition, SRP 6.5.2 currently suggests that the will retain soluble, gaseous, and solid fission products containment sump solution be maintained at values at such as iodines and cesium but provide no attenuation i
or above pH levels of 7, commencing with spray of the noble gases released from the core.The Reactor recirculation, to minimize revolatilization of iodine in Safety Study (WASH-1400, Ref. 5) assumed a the sump water. Current guidance states that decontamination factor (DF) of 100 for subcooled i
NUREG-1465 18
(
,.-m m.. - -
m I
t i
suppression pools and 1.0 for steam saturated pools.
radioactive aerosols and iodine released during Since 1975 when WASH-1400 was published, several postulated accident conditions.
detailed models have been developed for the removal of radioactive aerosols during steam flow through A typical ESF filtration system consists of redundant suppression pools.
trains that each have demisters to remove steam and 1
water droplets from the air entering the filter bank, Calculations for a BWR with a Mark I containment heaters to reduce the relative humidity of the air, high (Ref. 27) used in NUREG-1150 (Ref. 7) indicate that efficiency particulate air (HEPA) filters to remove
)
DFs ranged from 1.2 to about 4000 with a median value particulates, charcoal adsorbers to remove iodine in of about 80. The suppression pool has been shown to be elemental and organic form, followed finally by effective in scrubbing some of the most important additional HEPA filters to remove any charcoal fines radionuclides such as iodine, cesium, and IcIlurium, as released.
these are released in the early in-vessel phase. The NRC staff is also presently reviewing fission product Charcoal adsorber beds can be designed, as indicated in scrubbing by suppression pools to develop simplified Regulatory Guide 1.52, to remove from 90 to 99% of models.
the elemental iodine and from 30 to 99% of the organic if not bypassed, the suppression pool will also be i dide, depending upon the specific filter train design.
effective in scrubbing ex-vessel releases. Suppression pool bypass is an important aspect that places an upper Revised insights on accident source terms, given in limit on the overall performance of the suppression Section 3, may have several implications for ESF pool m scrubbing fission products. For example, if as filtration systems. Present ESF filtration systems are little as 1% of the fission products bypass the not sized to handle the mass loadings of non-suppression pool, the effective DF, taking bypass int radioactive aerosols that might be released as a result i
account, will be less than 100, regardless of the pool's of the ex-vessel release phase, which could produce ability to scrub fission products.
releases of significant quantities of nonradioactive as well as radioactive aerosols. However, if ESF filtration i
Although decontamination factors for the suppression systems are employed in conjunction with BWR pool are significant, the potential for iodine suppression pools or if significant quantities of water l
re-evolution can be important. Re-evolution of iodine are overlaying molten core debris (see Section 5.4),
was judged to be important in accident sequences large quantities of nonradioactive (as well as where the containment had failed and the suppression radioactive) aerosols will be scrubbed and retained by pool was boiling. There is presently no requirement for these water sources, thereby reducing the acrosol mass pli control in BWR suppression pools. Hence, it is loads upon the filter system.
possible that suppression pools would scrub substantial amounts of iodine in the early phases of an accident.
A second implication of revised source term insights for only to re-evolve it later as elemental iodine. It may ESF filtration systems is the impact of revised well be that additional materials likely to be in the understanding of the chemical form of iodine within suppression pool as a result of a severe accident, such containment. Present ESF filtration systems presume as cesium borate or cesium hydroxide and core-concrete that the chemical form of iodine is primarily elemental decomposition products, would counteract any iodine, and these systems include charcoal adsorber reduction in pH from radiolysis and would ensure that beds to trap and retain elemental iodine. Assuming that i
the pH level was sufficiently high to preclude PH control is maintained within the containment, a key I
re-evolution of elemental iodine. Therefore, if credit is question is whether charcoal beds are necessary. 'Bvo to be given for long-term retention of iodine in the questions appear to have a bearing on this issue and suppression pool, maintenance of the pH at or above a must be addressed, even assuming pH control. These level of 7 must be demonstrated. It is important to are (1) to what degree will Csl retained on particulate note, however, that this is not a matter of concern for filters decompose to evolve elemental iodine? and (2) present plants since all BWRs employ safety-related what effect would hydrogen burns have on the chemical filtration systems (see Section 5.3) designed to cope form of the iodine within containment? Based on with large quantities of elemental iodine. Hence, even preliminary information, Csl retained on particulate if the suppression pool were to re-evolve significant filters as an aerosol appears to be chemically stable amounts of elemental iodine, it would be retained by provided that it is not exposed to moisture. Exposure to the existing downstream filtration system.
moisture, however, would lead to Csl decomposition and production of iodine in ionic form (1-). which in 5.3 Filtration Systems turn would lead to re. evolution of elemental iodine.
Although ESF filtration systems are equipped with ESF filtration systems are discussed in Regulatory demisters and heaters to remove significant moisture Guide 1.52 (Ref. 32) and are used to reduce the before it reaches the charcoal adsorber bed, an 19 NUREG-1465
additional concern is that the demisters themselves may nere are four natural processes that remove acrosols trap some Csl aerosol.
from the containment atmosphere over a period of time: (1) gravitational settling, (2) diffusiophoresis, in conclusion, present ESF filtration systems, while (3) thermophoresis, and (4) particle diffusion. (Partic'e optimized to remove iodine, particularly in elemental diffusion is less important than the first three processes form, have HEPA filters that are effective in the and will not be discussed further.) All particles fall removal of particulates as well. Although such filtration naturally under the force of gravity and collect on any systems are not designed to handle the large mass available surface that terminates the fall, e.g., the floor loadings expected as a result of ex-vessel releases, when or upper surfaces of equipment. Both diffusiophoresis they are used in conjunction with large water sources and thermophoresis cause the deposition of aerosol such as IlWR suppression pools or significant water particles on all surfaces regardless of their orientation, depths overlaying core debris, the water sources will i.e., walls and ceiling as well as the floor, reduce the aerosol mass loading on the filter system Diffusiophoresis is the process by which water vapor in significantly, making such filter systems effective in the atmosphere ' drags' aerosol particles with it as it mitigation of a large spectrum of accident sequences.
migrates (diffuses) toward a relatively cold surface on which condensation is taking place. ncrmophoresis also causes acrosol particles to move toward and 1
5.4 Water Overlym.g Core Debris deposit on colder surfaces but not as a result of mass m tion. Rather, the decreasing average velocity of the Experimental measurements (Ref. 33) have shown that surrounding gas molecules tends to dnve the particle significant depths of water overlying any molten core down the temperature gradient until it traverses the debris after reactor pressure vessel failure will scrub interface layer and comes mto contact with the surface and retain particulate fission products. De question of where at sticks.
coolability of the molten debris as a result of water overlying it is still under investigation. A major factor Aerosol agglomeration is another natural phenomenon that may affect the degree of scrubbmg is whether the that has an influence on the rates at which the removal water layer m contact with the molten debn,s is boilmg processes described above will proceed. Agglomeration m n t.
results from the random inelastic collisions of particles p ces@gs a%a gram ca u
Results from Ref. 33 indicate that both subcooled as
"##E E
- E well as boilin8 water la7ers havin8 a depth of about gravitational settling. Three phenomena contribute to 3 meters had measured DFs of about 10. A recent study particle growth by agglomeration: (1) Brownian motion, (Ref. 34) performed for the NRC has provided a (2) gravitational fall, and (3) turbulence. Brownian simpitfied model to determine the degree of aerosol agglomeration is caused by particle collisions resulting scrubbing by a water pool overlying core debns from random ' buffeting' by high-energy gas molecules.
interacting with concrete.
Gravitational agglomeration results from the fact that some particles fall faster than others and therefore tend to collide with and stick to other slower falling,
5.5 Aerosol Deposition particles on their way down. Finally, rapid variations m Since the principal pathway for transport of fission gas velocity and flow direction in the atmosphere, i.e.,
products is via airborne particulates, i.e., aerosols, this turbulence, tend to increase the rate at which particle subject is discussed in some detail. Acrosols are usually collisions occur and thus increase the average particle thought of as solid particulates, but in general, the term size. It is to be expected that, as agglomeration also includes finely divided liquid droplets such as advances, the size of the particle will increase, and its water, i.e., fog. The two major sources of acrosols are shape can be expected to change as well.These latter condensation and entrainment. Condensation aerosols factors have a strong influence on the removal form when a vapor originating from some high-processes.
temperature source moves into a cooler region where the vapor falls below its saturation temperature and The agglomeration and acrosol removal processes all nucleation begins. Entrainment acrosols form when gas depend critically upon the thermodynamic state and l
bubbles break through a liquid surface and drag thermal-hydraulic conditions of the containment r
l droplets of the liquid phase into the wake of the bubble atmosphere. For example, the condensation onto and f
as it leaves the surface. In general, condensation evaporation of water from the aerosol particles particles are smaller in size (submicron to a few themselves have strong effects on all of the l
microns), while entrainment particles are usually larger agglomeration and removal processes. Water condensed (1.0-100 microns). Once airborne, both types of on acrosol particles increases their mass and makes aerosols behave in a similar manner with respect to them more spherical; both of these effects tend to both natural and engineered removal processes.
increase the rate of gravitational settling. Some NUREG-1465 20
aerosols, such as Csl and CsOH, are hygroscopic and Accident for Boiling Water Reactors," Regulatory absorb water vapor even when the containment Guide 1.3, Revision 2, June 1974.
ctmosphere is below saturation. As with condensation, hygroscopicity also increases the rate of deposition.
3.
U.S. Nuclear Regulatory Commission;
" Assumptions Used for Evaluating the Potential Because of its importance to fields such as weather and Radiological Consequences of a less of Coolant ctmosphere po!!ution, the behavior of acrosols has Accident for Pressurized Water Reactors,"
been under study for many decades. A number of Regulatory Guide 1.4, Revision 2, June 1974.
computer codes have been developed to specifically consider acrosol behavior as it relates to nuclear 4.
JJ. DiNunno et al., " Calculation of Distance accident conditions. The most complete mechanistic Factors for Power and 'lest Reactor Sites "
treatment of aerosol behavior in the reactor Technical Information Document (TID)-14844, containment is found in CONTAIN, a computer code U.S. Atomic Energy Commission,1%2.
developed at Sandia National Laboratories under NRC sponsorship for the analysis of containment behavior 5.
U.S. Nuclear Regulatory Commisska; " Reactor under severe accident conditions.The aerosol models Safety Study: An Assessment of Au dent Risksin in the NAUA code are very similar to those used in U.S. Commercial Nuclear Power Plants,"
CONTAIN; NAUA was developed at the WASH-1400 (NUREG-75/014), December 1975.
Kernforschungszentrum, Karlsrhue, ER.G., and was used for aerosol treatment in the NRC STCP. There 6.
J. A. Gieseke et al., " Source 'Ibrm Code Package:
are a number of other well-known aerosol behavior A User's Guide," NUREG/CR-4587 (BMI-213R computer codes, but these two are the most widely used prepared for NRC by Battelle Memorial Institute, end accepted throughout the international nuclear July 1986.
safety community.
7.
U.S. Nuclear Regulatory Commission; " Severe The rate at which gravitational settling occurs depends Accident Risks: An Assessment for Five U.S.
upon the degree of agglomeration at any particular Nuclear Power Plants," NUREG-1150, December time (i.e., the average particle size) as well as the total 1990.
particle density m (mass per unit volume). Thus, as in most cases where the decrement of a variable is 8.
M.R. Kuhlman, DJ. Lehmicke, and R.O. Meyer, proportional to the variable itself, one can expect an "CORSOR User's Manual," NUREG/CR-4173 cxponential behavior. 'Re gravitational settling process (BMI-2122), prepared for NRC by Battelle is quite complex and depends upon a large number of Memorial Laboratory, March 1985.
physical quantities, e.g., collision shape factor, particle settling shape factor, gas viscosity, effecGve settling 9.
H. Jordan, and M.R. Kuhlman, " TRAP-MELT 2 height, density correction factor, normalized Brownian User's Manual," NUREG/CR-4205 (BMI-2124),
collision coefficient, gravitational acceleration, and Prepared for NRC by Battelle Memorial particle material density. The only variable in this list Laboratory, May 1985.
that is independent of the plant, the accident scenario, and the atmospheric thermal-hydraulic conditions is the 10.
D.A. Powers, J.E. Brockmann, and A.W. Shiver, constant of gravitation. It follows that no single DF can "VANESA: A Mechanistic Model of Radionuclide be ascribed to cover the entire range of plant designs, Release and Aerosol Generation During Core accident scenarios, and source materials. An effort is Debris Interactions with Concrete,"
under way to establish a set of simplified algorithms NUREG/CR-4308 (SAND 85-1370), prepared for that can be used to provide a set of specific ranges of NRC by Sandia National Laboratories, July 1986.
atmosphere conditions. This effort is still underway at this time.
11.
P.C. Owczarski, A.K. Postma, and R.I. Schreck, "Ibchnical Bases and User's Manual for the Prototype of SPARC-A Suppression Pool Aerosol
- 6. REFERENCES Removal Code," NUREG/CR-3317 (PNLc4742),
prepared for NRC by Battelle Pacific Northwest 1.
U.S. Nuclear Regulatory Commission; " Reactor Laboratories, May 1985.
Site Criteria," Title 10, Code of Federal Regulations (CFR), Part 100.
12.
W.K. Winegardner, A.K. Postma, and M.W.
Jankowski," Studies of Fission Product Scrubbing 2.
U.S. Nuclear Regulatory Commission; within Ice Compartments," NUREG/CR-3248
" Assumptions Used for Evaluating the Potential (PNir4691), prepared for NRC by Battelle Pacific Radiological Consequences of a Loss of Coolant Northwest Laboratories, May 1983.
21 NUREG-1465
- 13. II. Bunz, M. Kayro, and W. Schock, "NAUA-Mod for NRC by Oak Ridge National Laboratory, 4: A Code for Calculating Aerosol Behavior in December 1992.
LWR Core Melt Accidents," KfK-3554, Kernforschungszentrum Karlsruhe Germany,
- 23. A.K. Postma, and R.W. Zavadowski, "Resiew of 1983.
Organic Iodide Formation Under Accident Conditions in Water-Cooled Reactors,"
- 14. R.M. Summers, et al., "MELCOR 1.8.0: A WASH-1233, U.S. Atomic Energy Commission, Computer Code for Nuclear Reactor Severe October 1972.
Accident Source'Ierm and Risk Assessment Analysis," NUREG/CR-5531 (SAND 90-0364),
- 24. E.C. Beahm, W.E. Shockley, and O.L Culberson, prepared for NRC by Sandia National
" Organic lodide Formation Following Nuclear Laboratories, January 1991.
Reactor Accidents," NUREG/CR-4327, (ORN11FM-9627), prepared for NRC by Oak 15.
K.R. Jones, et al,"'Iiming Analysis of PWR Fuel Ridge National Laboratory, December 1985.
Pin Failures," NUREG/CR-5787 (EGG-2657),
prepared for NRC by Idaho National Engineering
- 25. D. J. Osetek, " Low Volatile Fission Product Laboratory, September 1992.
Releases During Severe Reactor Accidents,"
DOE /ID-13177-2, prepared for U.S. Department
- 16. H.P. Nourbakhsh, M. Khatib-Rahbar, and R.E.
of Energy by Los AlamosTechnical Associates, Davis," Fission Product Release Characteristics October 1992.
into Containment Under Design Basis and Severe Accident Conditions," NUREG/CR-4881
- 26. M. Silberberg et al., " Reassessment of the (BNL-NUREG-52059), prepared for NRC by Tbchnical Bases for Estimating Source Thrms,"
Brookhaven National Laboratory, March 1988.
NUREG-0956, July 1986.
17.
H.P. Nourbakhsh.: " Estimates of Radionuclide 27.
R.S. Denning, et al., "Radionuclide Release Release Characteristics into Containment Under Calculations for Selected Severe Accident Severe Accidents," NUREG/CR-5747 Scenarios: BWR Mark I Design,"
(BNL-NUREG-52289), prepared for NRC by NUREG/CR-4624, Vol.1, prepared for NRC by Brookhaven National Laboratory, November 1993.
Battelle Memorial Institute, July 1986.
- 18. E.C. Beahm, C.E Weber, and T.S. Kress, " Iodine
- 28. U.S. Nuclear Regulatory Commission:
Chemical Forms in LWR Severe Accidents",
" Containment Spray as a Fission Product Cleanup NUREG/CR-5732 (ORN111N-11861), prepared System," Standard Review Plan, Section 6.5.2, for NRC by Oak Ridge National Laboratory, April Revision 2, NUREG-0800, December 1988.
1992.
- 29. H.P. Nourbakhsh,: "In-Containment Removal
- 19. JJ. Carbajo, " Severe Accident Source '1brm Mechanisms," Presentation to NRC staff January Characteristics for Selected Peach Bottom 3,1992, Brookhaven National Laboratory, January Sequences Predicted by the MELCOR Code,"
1992.
NUREG/CR-5942 (ORN11rM-12229), prepared for NRC by Oak Ridge National Laboratory, 30.
D.A. Powers and S.B. Burson, "A Simplified September 1993.
Model of Aerosol Removal by Containment Sprays," NUREG/CR-5966, (SAND 92-2689),
- 20. DJ. Alpert, D.I. Chanin, and LT. Ritchie, prepared for NRC by Sandia National
" Relative Importance of Individual Elements to Laboratories, June 1993.
Reactor Accident Consequences Assuming Equal Release Fractions."NUREG/CR-4467, prepared 31.
U.S. Nuclear Regulatory Commission: " Pressure for NRC by Sandia National Laboratories,1986.
Suppression Pool as a Fission Product Cleanup System," Standard Review Plan, Section 6.5.5,
- 21. C.E Weber, E.C. Beahm and T.S. Kress, "Models NUREG-0800, December 1988.
of Iodine Behavior in Reactor Containments "
OR!i11TM-12202. Oak Ridge National
- 32. U.S. Nuclear Regulatory Commission: " Design, Laboratory, October 1992.
Tbsting, and Maintenance Criteria for Postaccident Engineered-Safety-Feature Atmosphere Cleanup
- 22. E.C. Beahm, R.A. Lorenz, and C.E Weber, System Air Filtration and Adsorption Units of Ij ht-Water-Cooled Nuclear Power Plants,"
" Iodine Evolution and pH Control,"
g NUREG/CR-5950, (ORN111M-12242), prepared Regulatory Guide 1.52, Revision 2 March 1978.
NUREG-1455 22
(
i
- 33. ' J. Hakii et al., " Experimental Study on Aerosol
- 34. D.A. Powers and J.L Sprung, "A Simplified Model
, - Removal Efficiency for Pool Scrubbing Under of Aerosol Scrubbing by a Water Pbol Overlying
-l High'Ibmperature Steam Atmosphere,"
Core Debris Interacting With Concrete,"
Proceedings of the 21st DOE /NRC Nuclear Air NUREG/CR-5901, (SAND 92-1422), prepared for l
Cleaning Conference, August 1990.
NRC by Sandia National I.aboratories, November l
'1993.
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NUREG-1465 24
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Uncertainty Distributions for Total Releases into Containaient PWR,.14w RCS Pressure. Llane-stone Concrete Dry Cavity, Two Openings After VB, FPART = 1.
4 25 NUREG-1465 4
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Uncertainty Distribestions for Total Releases late Contalasnent PWR,14w RCS Pressere, Basakle Concrete, Dry Cavity, Two Opealags Aner VB, FFART = 1.
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27 NUREG-1465
APPENDIX B STCP BOUNDING VALUE RELEASES i
NUREG-1465 28
4 l
t i
Updated Bounding Value of Radionuclide Releases lato the Containment Under Severe Accident Conditions for PWRs E fl ILs IItxv" b
Hlah RCS Low RCS HI0h RCS Umestone Basaltic Hlah RCS Low RCS Pressure Pl. essure Pressure Concrete Concrete Pressure Pressure NG 1.0 1.0 0.
O.
O.
O.
O.
I 0.30 0.75 0.10 0.15 0.15 0.05 0.02 Cs 0.30 0.75 0.10 0.15 0.15 0.02 0.02 e
Te 0.20 0.50 0.05 0.40 0.30 0.02 0.01 Sr-Ba 0.003 0.01 0.01-0.40 0.15 Ru 0.003 0.01 0.05 0.005 0.005 1.a-Ce 5 x 10~5 1.5 x 10 0.01 0.05 0.05 4
Release 40 minutes 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />sM to hours Duration 4
(*3 All entries are fractions of Initial core Inventory.
M Assuming 100% of the core participate in CCI.
M Except for To and Ru where the duration is extended to five hours.
ZClc m
s:
twO
.L#
u Updated Bounding Value of Radionucilde Rolesses into the Contalnment Under Severe Accident Conditions for BWRs EIer AI.
M EIm Hinh RCS Low RCS Hioh RCS Limestone pasaltic Hiah RCS t.ow RCS Pressure Pressure *'
Pressure Concrete Concrete Pressure Pressure *'
NG 1.
1.
O.
O.
O.
O.
O.
I 0.50 0.75 0.10 0.15 0.15 0.10 0.02 Ca 0.50 0.75 0.10 0.15 0.15 0.05 0.01
.Te 0.10 0.15 0.05 0.50 0.30 0.02 0.02 8
Sree 0.003 0.01 0.01 0.70 0.30 Re 0.003 0.01 0.05 0.005 0.005 LarCe 5 x 10' 1.5 x 10 0.01 0.10 0.10 4
Release 1J hours 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />" 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> Duration
- 8 All entries are fractions of inittel core inventory.
I
" High pressure ATWS are also considered la this category.
"3 Assuming 100% of the core participate in CCI.
- Emeopt for Te and Ru where the duro6on is outended to six hours.
I
NRC Fome 336 U.S. NUCLE AR L EiULt. TORY COMMIS$lON I LEPOAYNUMBEN
- ck or.
1"*.'".'1"i = ;,1?lh,U*~ ""'
not uca BIBLIOGRAPHIC DATA SHEET
<sse,, men.cr== a * ""'"'
- 2. YlTLE AND sVSTITLE Accident Source Terms for Light-Water Nuclear Power Plants 3
DATE REPORT PUBLISHED mo r g
ama February 1995
- 6. TYPE OF REPORT
'"'D"*'*****'
L. Soffer, S. B. Burson, C. M. Ferrell. R. Y. Lee, J. N. Ridgely
'J. PEnFORMING ORGANIZATIGN - NAME AND ADOREls tse enc wr o.. sea. os+= er sapen. u.s murmer meesseis,y co-aumen.ewawaiar earnent seceawarme.e, ar asrae endf manfsar sesWatt Division of Systems Technology Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington. DC 20555 -0001
- 9. SPONSORING ORG ANIZ ATION - N AME AND ADDR ESS <se mac, era. s,aw m aae.e. er ear, armer.,,
m, one o oerne er messa, u s sucmar seyesere,y ca.-
c ew missiwie afsbesti Same as above
- 10. SUPPLEME NT ARY NOTES
- 11. ABSTRACT oco.orer er==>
In 1962 the U.S. Atomic Energy Commission published TID-14844. " Calculation of Distance Factors for Power and Test Reactors" which specified a release of Ossion products from the core to the reactor containment for a postulated accident involving " substantial meltdown of the core". This " source term", the basis for the NRC's Regulatory Guides 1.3 and 1.4, has been used to determine compliance with the NRC's reactor site criteria,10 CFR Part 100, and to evalude other important plant performance requirements.
During the past 30 years substantial additional information on fission product releases has been developed based on signincant severe accident research. This document utilizes this research by providing more realistic estimates of the " source term" release into containment, in terms of timing, nuclide types, quantities and chemical form, given a severe core-melt accident. This revised " source term" is to be applied to the design of future light water reackrs (LWRs). Current LWR licensees may voluntarily propose applications based upon it.
- 2. avaspagsTv sTATsastwT
- 12. Q'.E v vs RDs/DEsCA :Pf ors tin, e 4 er a, esse ener ed assee seameness a. air.es, sne, esser.s Unhmited n
- 14. StCuRITY CLA358F6CATich craa roes Severe Accident Source Term Unclassified Core Meltdown tra.
Unclassified Design Basis Accident is. NUMsER oF Pacts TID-14844 Replacement Core Fission Product Releases
- 16. PRICE seac eoans as casi
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