NUREG-1465

Accident Source Terms for

Light-Water Nuclear Power Plants

Final Report

U.S. Nuclear Regulatory Commission

Offlce of Nuclear Regulatory Research

L Soffer, S. B. Burson, C. M. Ferrell,

R. Y. Lee, J. N. Ridgely

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NUREG-1465

Accident Source Terms for

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

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

9 A I8,4
. ... ...

Abstract

In 1962 The U.S. Atomic Energy Commission published

TLD-14844, "Calculation of Distance Factors for Power

and Test Reactors" which specified a release of fission

products from the core to the reactor containment in

the event of a postulated accident involving a

"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 evaluate

other important plant performance requirements.

During the past 30 years substantial additional

information on fission product releases has been

developed based on significant 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 Reactors

(LWRs). Current LWR licensees may voluntarily

propose applications based upon it. These will be

reviewed by the NRC staff.

iii NUREG-1465
CONTENTS

Page

Abstract .............................................................................. iii

Preface......................................................................................... vii

1 Introduction And Background ...................................... . I

1.1 Regulatory Use of Source Terms ..I........................................................ 1

1.2 Research Insights Since TID-14844 ....................................................... 2

2 Objectives And Scope ...................................... . 3

2.1 General .............................................................................. 3

2.2 Accidents Considered . ........................................................... 3

2.3 Limitations ............................................................................ 4

3 Accident Source Terms ...................................... .5

3.1 Accident Sequences Reviewed ........................................................... 5

3.2 Onset of Fission Product Release ......................................................... 5

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

3.4 Fission Product Composition and Magnitude .......................... . 9

3.5 Chemical Form ........................................................................ 10

3.6 Proposed Accident Source Terms ...................... 12

3.7 Nonradioactive Aerosols ...................... 14

4 Margins And Uncertainties ...................... 15

4.1 Accident Severity and Type ......... .. 15

4.2 Onset of Fission Product Release ......... . . 15

4.3 Release Phase Durations ......... .. 15

4.4 Composition and Magnitude of Releases .................................................. 16

4.5 Iodine Chemical Form .......................................................... 17

5 In-Containment Removal Mechanisms .......................................................... 17

5.1 Containment Sprays .......................................................... 18

5.2 BWR Suppression Pools .......................................................... 18

5.3 Filtration Systems ............................ .............................. 19

5.4 Water Overlying Core Debris ........................................................... 20

5.5 Aerosol Deposition ............................ .............................. 20

6 References ........................................................... 21

TABLES

1.1 Release Phases of a Severe Accident ............... ....................................... 2

3.1 BWR Source Term Contributing Sequences ........................ 5.......................

3.2 PWR Source Tberm 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 STCP Radionuclide Groups ............ ............................................ 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) ....................... 11

v NUREG-1465
CONTENTS (Cont'd)

Page

3.10 Mean Values of Radionuclides Into Containment for BWRs,

Low RCS Pressure, High Zirconium Oxidation ............ ................................. 11

3.11 Mean Values of Radionuclide Releases Into Containment for PWRs, Low RCS Pressure,

High Zirconium Oxidation ....................... ............................. 11

3.12 BWR Releases Into Containment .................................................... 13

3.13 PWR Releases Into Containment .................................................... 13

4.1 Measures of Low Volatile In-Vessel Release Fractions ............... ....................... 16

APPENDICES

A Uncertainty Distributions ...................................................... 24

B STCP Bounding Value Releases .......................... ............................ 28

NUREG-1465 vi
Preface

In 1962, the Atomic Energy Commission issued

Technical Information Document (IMD) 14844,

"Calculation of Distance Factors for Power and Test

Reactors." In this document, a release of fission

products from the core of a light-water reactor (LWR)

into the containment atmosphere ("source term") was

postulated for the purpose of calculating off-site doses

in accordance with 10 CFR Part 100, "Reactor Site

Criteria." The source term postulated an accident that

resulted in substantial meltdown of the core, and the

fission products assumed released into the containment

were based on an understanding at that time of fission

product behavior. In addition to site suitability, the

regulatory applications of this source term (in

conjunction with the dose calculation methodology)

affect the design of a wide range of plant systems.

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

provide a postulated fission product source term

released into containment that is based on current

understanding of LWR accidents and fission product

behavior. The information contained in this document

is applicable to LWR designs and is intended to form

the basis for the development of regulatory guidance,

primarily for future LWRs. This report will serve as a

basis for possible changes to regulatory requirements.

However, acceptance of any proposed changes will be

on a case-by-case basis.

Source terms for future reactors may differ from those

presented in this report which are based upon insights

derived from current generation light-water reactors.

An applicant may propose changes in source term

parameters (timing, release magnitude, and chemical

form) from those contained in this report, based upon

and justified by design specific features.

NUREG-1465
1 INTRODUCTION AND BACKGROUND

1.1 Regulatory Use of Source Terms

The use of postulated accidental releases of radioactive

materials is deeply embedded in the regulatory policy

and practices of the U.S. Nuclear Regulatory

Commission (NRC). For over 30 years, the NRC's

reactor site criteria in 10 CFR Part 100 (Ref. 1) have

required, for licensing purposes, that an accidental

fission product release resulting from "substantial

meltdown" of the core into the containment be

postulated to occur and that its potential radiological

consequences be evaluated assuming that the containment remains intact but leaks at its maximum allowable

leak rate. Radioactive material escaping from the

containment is often referred to as the "radiological

release to the environment." The radiological release

is obtained from the containment leak rate and a

knowledge of the airborne radioactive inventory in the

containment atmosphere. The radioactive inventory

within containment is referred to as the

"in-containment accident source term."

The expression "in-containment accident source term,"

as used in this document, denotes the radioactive

material composition and magnitude, as well as the

chemical and physical properties of the material within

the containment that are available for leakage from the

reactor to the environment. The "in-containment

accident source term" will normally be a function of

time and will involve consideration of fission products

being released from the core into the containment as

well as removal of fission products by plant features

intended to do so (e.g., spray systems) or by natural

removal processes.

For currently licensed plants, the characteristics of the

fission product release from the core into the

containment are set forth in Regulatory Guides 1.3 and

1.4 (Refs. 2,3) and have been derived from the 1962

report, TID-14844 (Ref. 4). This release consists of

100% of the core inventory of noble gases and 50% of

the iodines (half of which are assumed to deposit on

interior surfaces very rapidly). These values were based

largely on experiments performed in the late 1950s

involving heated irradiated U0 2 pellets. TID-14844

also included 1% of the remaining solid fission

products, but these were dropped from consideration in Regulatory Guides 1.3 and 1.4. The 1% of

the solid fission products are considered in certain

areas such as equipment qualification.

Regulatory Guides 1.3 and 1.4 (Refs. 2 and 3) specify

that the source term within containment is assumed to

be instantaneously available for release and that the

iodine chemical form is assumed to be predominantly

(91%) in elemental (02) form, with 5% assumed to be

particulate iodine and 4% assumed to be in organic

form. These assumptions have significantly affected the

design of engineered safety features. Containment

isolation valve closure times have also been affected by

these assumptions.

Use of the TID-14844 release has not been confined to

an evaluation of site suitability and plant mitigation

features such as sprays and filtration systems. The

regulatory applications of this release are wide,

including the basis for (1) the post-accident radiation

environment for which safety-related equipment should

be qualified, (2) post-accident habitability requirements

for the control room, and (3) post-accident sampling

systems and accessibility.

In contrast to the TID-14844 sourge term and

containment leakage release used for design basis

accidents, severe accident releases to the environment

first arose in probabilistic risk assessments (e.g.,

Reactor Safety Study, WASH-1400 (Ref. 5)) in

examining accident sequences that involved core melt

and containments that could fail. Severe accident

releases represent mechanistically determined best

estimate releases to the environment, including

estimates of failures of containment integrity. This is

very different from the combination of the nonmechanistic release to containment postulated by

TID-14844 coupled with the assumption of very limited

containment leakage used for Part 100 siting calculations for design basis accidents. The worst severe

accident releases resulting from containment failure or

containment bypass can lead to consequences that are

much greater than those associated with a TID-14844

source term released into containment where the

containment is assumed to be leaking at its maximum

leak rate for its design conditions. Indeed, some of the

most severe releases arise from some containment

bypass events, such as rupture of multiple steam

generator tubes.

Although severe accident source terms have not been

used in individual plant licensing safety evaluations,

they have had significant regulatory applications.

Source terms from severe accidents (beyond-designbasis accidents) came into regulatory consideration and

usage shortly after the issuance of WASH-1400 in 1975,

and their application was accelerated after the Three

Mile Island accident in March 1979. Current

applications rely to a large extent on the results of

WASH-1400 and include (1) part of the basis for the

sizes of emergency planning zones for all plants, (2) the

basis for staff assessments of severe accident risk in

I NUREG-1465
plant environmental impact statements, and (3) part of

the basis for staff prioritization and resolution of

generic safety issues, unresolved safety issues, and

other regulatory analyses. Source term assessments

based on WASH-1400 methodology appear in many

probabilistic risk assessment studies performed to date.

1.2 Research Insights Since

TID-14844

Source term estimates under severe accident conditions

became of great interest shortly after the Three-Mile

Island (rMI) accident when it was observed that only

relatively small amounts of iodine were released to the

environment compared with the amount predicted to

be released in licensing calculations. This led a number

of observers to claim that severe accident releases were

much lower than previously estimated.

The NRC began a major research effort about 1981 to

obtain a better understanding of fission-product

transport and release mechanisms in LWRs under

severe accident conditions. This research effort has

included extensive NRC staff and contractor efforts

involving a number of national laboratories as well as

nuclear industry groups. These cooperative research

activities resulted in the development and application

of a group of computer codes known as the Source

Term Code Package (STCP) (Ref. 6) to examine

core-melt progression and fission product release and

transport in LWRs. The NRC staff has also sponsored

significant review efforts by peer reviewers, foreign

partners in NRC research programs, industry groups,

and the general public. The STCP methodology for

severe accident source terms has also been reflected in

NUREG-1150 (Ref. 7), which provides an updated risk

assessment for five U.S. nuclear power plants.

As a result of the NRC's research effort to obtain a

better understanding of fission product transport and

release mechanisms in LWRs under severe accident

conditions, the STCP emerged as an integral tool for

analysis of fission product transport in the reactor

coolant system (RCS) and containment. The STCP

models release from the fuel with CORSOR (Ref. 8)

and fission product retention and transport in the RCS

with TRAPMELT (Ref. 9). Releases from core-concrete

interactions are modeled using the VANESA and

CORCON (Ref. 10) codes. Depending upon the

containment type, SPARC or ICEDF (Refs. 11,12) are

used in conjunction with NAUA (Ref.13) to model the

transport and retention of fission product releases from

the RCS and from core-concrete interactions into the

containment, with subsequent release of fission

products to the environment consistent with the state

of the containment.

Improved modeling of severe accident phenomena,

including fission product transport, has been provided

by the recently developed MELCOR (Ref. 14) code. At

this time, however, an insufficient body of calculations

is available to provide detailed insights from this model.

Using analyses based on the STCP and MELCOR

codes and NUREG-1150, the NRC has sponsored

studies (Refs. 15-17) that analyzed the timing,

magnitude, and duration of fission product releases. In

addition, an examination and assessment of the

chemical form of iodine likely to be found within

containment as a result of a severe accident has also

been carried out (Ref. 18).

In contrast to the instantaneous releases that were

postulated in Regulatory Guides 1.3 and 1.4, analyses of

severe accident sequences have shown that, despite

differences in plant design and accident sequence, such

releases can be generally categorized in terms of

phenomenological phases associated with the degree of

fuel melting and relocation, reactor pressure vessel

integrity, and, as applicable, attack upon concrete below

the reactor cavity by molten core materials. The

general phases, or progression, of a severe LWR

accident are shown in Table 1.1.

Table 1.1 Release Phases of a Severe Accident

Release Phases

Coolant Activity Release

Gap Activity Release

Early In-Vessel Release

Ex-Vessel Release

Late In-Vessel Release

Initially there is a release of coolant activity associated

with a break or leak in the reactor coolant system.

Assuming that the coolant loss cannot be accommodated by the reactor coolant makeup systems or the

emergency core cooling systems, fuel cladding failure

would occur with a release of the activity located in the

gap between the fuel pellet and the fuel cladding.

As the accident progresses, fuel degradation begins,

resulting in a loss of fuel geometry accompanied by

gradual melting and slumping of core materials to the

bottom of the reactor pressure vessel. During this

period, the early in-vessel release phase, virtually all

the noble gases and significant fractions of the volatile

nuclides such as iodine and cesium are released into

containment. The amounts of volatile nuclides released

into containment during the early in-vessel phase are

strongly influenced by the residence time of the

radioactive material within the RCS during core

degradation. High pressure sequences result in long

NUREG-1465 2
residence times and significant retention and plateout

of volatile nuclides within the RCS, while low pressure

sequences result in relatively short residence times and

little retention within the RCS and consequently higher

releases into containment.

If failure of the bottom head of the reactor pressure

vessel occurs, two additional release phases may occur.

Molten core debris released from the reactor pressure

vessel into the containment will interact with the

concrete structural materials of the cavity below the

reactor (ex-vessel release phase). As a result of these

interactions, quantities of the less volatile nuclides may

be released into containment. Ex-vessel releases are

influenced somewhat by the type of concrete in the

reactor cavity. Limestone concrete decomposes to

produce greater quantities of CO and CO2 gases than

basaltic concrete. These gases may, in turn, sparge

some of the less volatile nuclides, such as barium and

strontium, and small fractions of the lanthanides into

the containment atmosphere. Large quantities of

non-radioactive aerosols may also be released as a

result of core-concrete interactions. The presence of

water in the reactor cavity overlying any core debris can

significantly reduce the ex-vessel releases (both

radioactive and non-radioactive) into the containment,

either by cooling the core debris, or at least by

scrubbing the releases and retaining a large fraction in

the water. The degree of scrubbing will depend, of

course, upon the depth and temperature of any water

overlying the core debris. Simultaneously, and

generally with a longer duration, late in-vessel releases

of some of the volatile nuclides, which had deposited in

the reactor coolant system during the in-vessel phase,

will also occur and be released into containment.

Two other phenomena that affect the release of fission

products into containment could also occur, as

discussed in Reference 7. The first of these is referred

to as "high pressure melt ejection" (HPME). If the

RCS is at high pressure at the time of failure of the

bottom head of the reactor pressure vessel, quantities

of molten core materials could be injected into the

containment at high velocities. In addition to a

potentially rapid rise in containment temperature, a

significant amount of radioactive material could also be

added to the containment atmosphere, primarily in the

form of aerosols. The occurrence of HPME is

precluded at low RCS pressures. A second

phenomenon that could affect the release of fission

products into containment is a possible steam explosion

as a result of interactions between molten core debris

and water. This could lead to fine fragmentation of

some portion of the molten core debris with an increase

in the amount of airborne fission products. While small

scale steam explosions are considered quite likely to

occur, they will not result in significant increases in the

airborne activity already within containment. Large

scale steam explosions, on the other hand, could result

in significant increases in airborne activity, but are

much less likely to occur. In any event, releases of

particulates or vapors during steam explosions will also

be accompanied by large amounts of water droplets,

which would tend to quickly sweep released material

from the atmosphere.

2 OBJECTIVES AND SCOPE

2.1 General

The primary objective of this report is to define a

revised accident source term for regulatory application

for future LWRs. The intent is to capture the major

relevant insights available from recent severe accident

research on the phenomenology of fission product

release and transport behavior. The revised source

term is expressed in terms of times and rates of

appearance of radioactive fission products into the

containment, the types and quantities of the species

released, and other important attributes such as the

chemical forms of iodine. This mechanistic approach

will therefore present, for regulatory purposes, a more

realistic portrayal of the amount of fission products

present in the containment from a postulated severe

accident.

2.2 Accidents Considered

In order to determine accident source terms for

regulatory purposes, a range of severe accidents that

have been analyzed for LWR plants was examined.

Evaluation of a range of severe accident sequences was

based upon work done in support of NUREG-1150

(Ref. 7). This work is documented in NUREG/CR-5747

(Ref. 17) and employed the integrated Source Term

Code Package (STCP) computer codes, together with

insights from the MELCOR code, which were used to

analyze specific accident sequences of interest to

provide the accident chronology as well as detailed

estimates of fission product behavior within the reactor

coolant system and the other pertinent parts of the

plant. The sequences studied progressed to a complete

core melt, involving failure of the reactor pressure

vessel and including core-concrete interactions, as well.

A key decision to be made in defining an accident

source term is the severity of the accident or group of

accidents to be considered. Footnote 1 to 10 CFR Part

100 (Ref. 1). in referring to the postulated fission

product release to be used for evaluating sites, notes

that "Such accidents have generally been assumed to

result in substantial meltdown of the core with

subsequent release of appreciable quantities of fission

products." Possible choices range from (1) slight fuel

damage accidents involving releases into containment

3 NUREG-1465
of a small fraction of the volatile nuclides such as the

noble gases, (2) severe core damage accidents involving

major fuel damage but without reactor vessel failure or

core-concrete interactions (similar in severity to the

TMI accident), or (3) complete core-melt events with

core-concrete interactions. These outcomes are not

equally probable. Since many reactor systems must fail

for core degradation with reactor vessel failure to occur

and core-concrete interactions to occur, one or more

systems may be returned to an operable status before

core melt commences. Hence, past operational and

accident experience together with information on

modern plant designs, together with a vigorous program

aimed at developing accident management procedures,

indicate that complete core-melt events resulting in

reactor pressure vessel failure are considerably less

likely to occur than those involving major fuel damage

without reactor pressure vessel failure, and that these,

in turn, are less likely to occur than those involving

slight fuel damage.

For completeness, this report displays the mean or

average release fractions for all the release phases

associated with a complete core melt. However, it is

concluded that any source term selected for a particular

regulatory application should appropriately reflect the

likelihood associated with its occurrence.

It is important to emphasize that the release fractions

for the source terms presented in this report are

intended to be representative or typical, rather than

conservative or bounding values, of those associated

with a low pressure core-melt accident, except for the

initial appearance of fission products from failed fuel,

which was chosen conservatively. The release fractions

are not intended to envelope all potential severe

accident sequences, nor to represent any single

sequence, since accident sequences yielding both higher

as well as lower release fractions were examined and

factored into the final report presented here.

Source terms for future reactors may differ from those

presented in this report which are based upon insights

derived from current generation light-water reactors.

An applicant may propose changes in source term

parameters (timing, release magnitude, and chemical

form) from those contained in this report, based upon

and justified by design specific features.

The NRC staff also intends to allow credit for removal

or reduction of fission products within containment via

engineered features provided for fission product

reduction such as sprays or filters, as well as by natural

processes such as aerosol deposition. These are

discussed in Section 5.

2.3 Limitations

The accident source terms defined in this report have

been derived from examination of a set of severe

accident sequences for LWRs of current design.

Because of general similarities in plant and core design

parameters, these results are also considered to be

applicable to evolutionary LWR designs such as

General Electric's Advanced Boiling Water Reactor

(ABWR) and Combustion Engineering's (CE) System

80+.

Currently, the NRC staff is reviewing reactor designs

for several smaller LWRs employing some passive

features for core cooling and containment heat

removal. While the "passive" plants are generally

similar to present LWRs, they are expected to have

somewhat lower core power densities than those of

current LWRs. Hence, an accident for the passive

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

tables given in Section 3.3 are probably shorter than

those applicable to the passive plants. The release

fractions shown may also be overestimated somewhat

for high pressure sequences associated with the passive

plants, since longer times for accident progression

would also allow for enhanced retention of fission

products in the primary coolant system during core

heatup and degradation. Despite the lack of specific

accident sequence information for these designs, the

in-containment accident source terms provided below

may be considered generally applicable to the "passive"

designs.

The accident source terms provided in this report are

not considered applicable to reactor designs that are

very different from LWRs, such as high-temperature

gas-cooled reactors or liquid-metal reactors.

Recent information has indicated that high burnup fuel,

that is, fuel irradiated at levels in excess of about 40

GWD/MTU, may be more prone to failure during

design basis reactivity insertion accidents (RIA) than

previously thought. Preliminary indications are that

high burnup fuel also may be in a highly fragmented or

powdered form, so that failure of the cladding could

result in a significant fraction of the fuel itself being

released. In contrast, the source term contained in this

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

of those fission product gases residing in the gap

between the fuel pellet and the cladding. Because of

this recent information regarding high burnup fuels, the

NRC staff cautions that, until further information

indicates otherwise, the source term in this report

(particularly gap activity) may not be applicable for fuel

NUREG-1465 4
--- - --

irradiated to high burnup levels (in excess of about 40

GWDIMTU).

3 ACCIDENT SOURCE TERMS

The expression "in-containment source terms," as used

in this report, denotes the fission product inventory

present in the containment at any given time during an

accident. lb evaluate the in-containment source term

during the course of an accident, the time-history of the

fission product release from the core into the

containment must be known, as well as the effect of

fission product removal mechanisms, both natural and

engineered, to remove radioactive materials from the

containment atmosphere. This section discusses the

time-history of the fission product releases into the

containment. Removal mechanisms are discussed in

Section 5.

3.1 Accident Sequences Reviewed

All the accident sequences identified in NUREG-1150

were reviewed and some additional Source lbrm Code

Package (STCP) and MELCOR calculations were

p&formed. The dominant sequences which are

considered to significantly impact the source term are

summarized in Thble 3.1 for BWRs and Table 3.2 for

PWRs.

3.2 Onset of Fission Product Release

This section discusses the assumptions used in selecting

the scenario appropriate for defining the early phases

of the source term (coolant activity and gap release

phases). It was considered appropriate to base these

early release phases on the design basis initiation that

could lead to earliest fuel failures.

A review of current plant final safety analysis reports

(FSARs) was made to identify all design basis accidents

in which the licensee had identified fuel failure. For all

accidents with the potential for release of radioactivity

into the environment, the class of accident that had the

shortest time until the first fuel rod failed was the

design basis LOCA. As might be expected, the time

until cladding failure is very sensitive to the design of

the reactor, the type of accident assumed, and the fuel

rod design. In particular, the maximum linear heat

generation rate, the internal fuel rod pressure, and the

stored energy in the fuel rod are significant

considerations.

Table 3.1 BWR Source Term Contributing Sequences

Plant Sequence Description

Peach Bottom TC1 ATWS with reactor depressurized

TC2 ATWS with reactor pressurized

TC3 TC2 with wetwell venting

TB1 SBO with battery depletion

TB2 TB1 with containment failure at vessel failure

S2E1 LOCA (2"), no ECCS and no ADS

S2E2 S2E1 with basaltic concrete

V RHR pipe failure outside containment

TBUX SBO with loss of all DC power

LaSalle TB SBO with late containment failure

Grand Gulf TC ATWS early containment failure fails ECCS

TBI SBO with battery depletion

TB2 TB1 with H2 burn fails containment

TBS SBO, no ECCS but reactor depressurized

TBR TBS with AC recovery after vessel failure

SBO Station Blackout LOCA Loss of Coolant Accident

RCP Reactor Coolant Pump RHR Residual Heat Removal

ADS Automatic Depressurization System ATWS Anticipated Transient Without Scram

5 NUREG-1465
Table 3.2 PIVR Source Term Contributing Sequences

Plant Sequence Description

Surry AG LOCA (hot leg), no containment heat removal systems

TMLB' LOOP, no PCS and no AFWS

V Interfacing system LOCA

M3B SBO with RCP seal LOCA

S2D-8 SBLOCA, no ECCS and H2 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-H2 burn or DCH fails containment

S2DCF2 S2DCF1 except late H2 or overpressure failure of

containment

TMLU Transient, no PCS, no ECCS, no AFWS-DCH fails

containment

Oconee 3 TMLB' SBO, no active ESF systems

S1DCF 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

M3B LOCA (3") with SBO

TBA SBO induces hot leg LOCA-hydrogen burn fails

containment

ACD LOCA (hot leg), no ECCS no CS

M3B1 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 Spray CSRS CS Recirculation System

ATWS Anticipated Transient Without Scram LOOP Loss of Offsite Power

The details of the specific accident sequences are documented in NUREG/CR-5747, Estimate of

Radionuclide Release Characteristics into Containment Under Severe Accident Conditions (Ref. 17).

To determine whether a design basis LOCA was a

reasonable scenario upon which to base the timing of

initial fission product release into the containment,

various PRAs were reviewed to determine the

contribution to core damage frequency (CDF) resulting

from LOCAs. This information is shown in Table 3.3.

As can be seen from this table, LOCAs are a small

contributor to CDF for BWRs, but can be a substantial

contributor for PWRs. Therefore, for PWRs a large

LOCA is considered a reasonable initiator to assume

for modeling the earliest appearance of the gap activity

if the plant has not been approved for leak before

break (LBB) operation. For plants that have received

LBB approval, a small LOCA (6" line break) would

more appropriately model the timing. For BWRs, large

LOCAs may not be an appropriate scenario for gap

activity timing. However, since the time to initial fuel

rod failure is long for BWRs, even for large LOCAs,

NUREG-1465 6
use of the large LOCA scenario should not unduly

penalize BWRs and will maintain consistency with the

assumptions for the PWR. As with the PWR, for an

LBB approved plant, the timing associated with a small

LOCA (6" line break) would be more appropriate.

In order to provide a realistic estimate of the shortest

time for fuel rod failure for the LOCA, calculations

were performed using the FRAPCON2,

SCDAP/RELAP5 MOD 3.0, and FRAPT6 computer

codes for two plants. The two plants were a Babcock

and Wilcox (B&W) plant with a 15 by 15 fuel rod array

and a Westinghouse 4-loop (M!) plant with a 17 by 17

fuel rod array. For each plant, a sensitivity study was

performed to identify the size of the LOCA that

resulted in the shortest fuel rod failure time (Ref. 15).

In both cases, the accident was a double-ended

guillotine rupture of the cold leg pipe. The minimum

time from the time of accident initiation until first fuel

rod failure was calculated to be 13 and 24.6 seconds for

the B&W and _ plants, respectively. A sensitivity

study was performed to determine the effect of tripping

or not tripping the reactor coolant pumps. The results

indicated that tripping of the reactor coolant pumps

had no appreciable impact on timing. For a 6-inch line

break, the time until first fuel rod failure is expected to

be greater than 6.5 and 10 minutes, respectively.

Table 33 Contribution of LOCAs to Core Damage Frequency (CDF)-Internal Events

Percent or CDF

Percent of CDF caused by large LOCAs

Boiling Water Reactors caused by LOCAs (>6" line break)

Peach Bottom (NUREG-1150) 3.5 1.0

Grand Gulf (NUREG-1150) 0.1 0.03

Millstone 1 (Utility) 23 13

Pressurized W'ater Reactors

Surry (NUREG-1150) 15 4.3

Sequoyah (NUREG-1150) 63 4.6

Zion (NUREG-1150) 87 1.4

Calvert Cliffs (IREP) 21 < 1

Oconee-3 (EPRI/NSAC) 43 3.0

A comparison calculation was done using the TRACPF1 MOD I code, version 14.3U5Q.LG on the W

plant. This analysis indicated that the first fuel rod

failure would occur 34.9 seconds after pipe rupture, in

contrast to the value of 24.6 seconds calculated using

SCDAPIRELAP. The reasons for the difference

between the SCDAP/RELAPS MOD 3.0 and

TRAC-PF1 MOD 1 are discussed in Reference 15.

The review of the FSARs for BWRs indicates that fuel

failures may occur significantly later, on the order of

several minutes or more. No calculations have been

performed using the aforementioned suite of codes.

For determining the time of appearance of gap activity

in the containment (i.e., initial fuel failure), which

corresponds to the duration of the coolant activity

phase and the beginning of the gap activity phase, it

would be appropriate to perform a plant specific

calculation using the codes described above. However,

if no plant specific calculations are performed, the

minimum times discussed above may be used to provide

an estimate of the earliest time to fuel rod failure.

Source terms for future reactors may differ from those

presented in this report which are based upon insights

derived from current generation light-water reactors.

An applicant may propose changes in source term

parameters (timing, release magnitude, and chemical

form) from those contained in this report, based upon

and justified by design specific features.

3.3 Duration of Release Phases

Section 1.2 provided a qualitative discussion of the

release phases of an accident. This section provides

estimated durations for these release phases.

The coolant activity phase begins with a postulated pipe

rupture and ends when the first fuel rod has been

estimated to fail. During this phase, the activity

released to the containment atmosphere is that

associated with very small amounts of radioactivity

dissolved in the coolant itself. As discussed in Section

3.2 above, this phase is estimated to last about 25

seconds for Westinghouse PWRs, and about 13 seconds

for B&W PWRs, assuming a large break LOCA. For a

smaller LOCA (e.g., a 6-inch line break), such as would

7 NUREG-1465
be considered for a plant that has received LBB

approval, the coolant activity phase duration would be

expected to be at least 10 minutes. Although not

specifically evaluated at this time, Combustion

Engineering (CE) PWRs would be expected to have

coolant activity durations similar to Westinghouse

plants. For BWRs, the coolant activity phase would be

expected to last longer; however, unless plant specific

calculations are made, the durations discussed above

are considered applicable.

The gap activity release phase begins when fuel

cladding failure commences. This phase involves the

release of that radioactivity that has collected in the

gap between the fuel pellet and cladding. This process

releases to containment a few percent of the total

inventory of the more volatile radionuclides,

particularly noble gases, iodine, and cesium. During this

phase, the bulk of the fission products continue to be

retained in the fuel itself. The gap activity phase ends

when the fuel pellet bulk temperature has been raised

sufficiently that significant amounts of fission products

can no longer be retained in the fuel. As noted in

Reference 16, a review of STCP calculated results for

six reference plants, PWRs as well as BWRs, indicated

that significant fission product releases from the bulk of

the fuel itself were estimated to commence no earlier

than about 30 minutes and 60 minutes for PWRs and

BWRs, respectively, after the onset of the accident.

However, more recent calculations (Ref. 19) for the

Peach Bottom plant using the MELCOR code

indicated that the durations of the gap release for three

BWR accident sequences were about 30 minutes, as

well. On this basis, the duration of the gap activity

release phase has been selected to be 0.5 hours, for

both PWRs and BWRs.

During the early in-vessel release phase, the fuel as

well as other structural materials in the core reach

sufficiently high temperatures that the reactor core

geometry is no longer maintained and fuel and other'

materials melt and relocate to the bottom of the

reactor pressure vessel. During this phase, significant

quantities of the volatile nuclides in the core inventory

as well as small fractions of the less volatile nuclides

are estimated to be released into containment. This

release phase ends when the bottom head of the

reactor pressure vessel fails, allowing molten core

debris to fall onto the concrete below the reactor

pressure vessel. Release durations for this phase vary

depending on both the reactor type and the accident

sequence. Tables 3.4 and 3.5, based on results from

Reference 16, show the estimated duration times for

PWRs and BWRs, respectively.

Table 3.4 In-Vessel Release Duration for PWR Sequences

Release Duration

Plant Accident Sequence (Min)

Surry TMLB' (H) 41

Surry S3B (H) 36

Surry AG (L) 215

Surry V (L) 104

Zion TMLU (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

Oconee SPDCF (L) 84

*(H or L) Denotes whether the accident occurs at high or low pressure.

Based on the information in these tables, the staff

concludes that the in-vessel release phase is somewhat

longer for BWR plants than for PWR plants. This is

largely due to the lower core power density in BWR

plants that extends the time for complete core melt.

Representative times for the duration of the in-vessel

release phase have been selected to be 1.3 hours and

1.5 hours, for PWR and BWR plants respectively, as

recommended by Reference 17.

The ex-vessel release phase begins when molten core

debris exits the reactor pressure vessel and ends when

NUREG-1465 &
Table 3.5 In-Vessel Release Duration for BWR Sequences

Release Duration

Plant Accident Sequence* (Min)

Peach Bottom TC2 (H) 66

Peach Bottom TC3 68

Peach Bottom TC1 (L) 97

Peach Bottom TB1ITB2 (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/TBR (L) 96

*(H or L) denotes whether the accident occurs at high or low pressure.

the debris has cooled sufficiently that significant

quantities of fission products are no longer being

released. During this phase, significant quantities of the

volatile radionuclides not already released during the

early in-vessel phase as well as lesser quantities of

non-volatile radionuclides are released into

containment. Although releases from core-concrete

interactions are predicted to take place over a number

of hours after vessel breach, Reference 16 indicates

that the bulk of the fission products (about 90%), with

the exception of tellurium and ruthenium, are expected

to be released over a 2-hour period for PWRs and a

3-hour period for BWRs. For tellurium and ruthenium,

ex-vessel releases extend over 5 and 6 hours,

respectively, for PWRs and BWRs. The difference in

duration of the ex-vessel phase between PWRs and

BWRs is largely attributable to the larger amount of

zirconium in BWRs, which provides additional chemical

energy of oxidation. Based on Reference 17, the

ex-vessel release phase duration is taken to be 2 and 3

hours, respectively, for PWRs and BWRs.

The late in-vessel release phase commences at vessel

breach and proceeds simultaneously with the

occurrence of the ex-vessel phase. However, the

duration is not the same for both phases. During this

release phase, some of the volatile nuclides deposited

within the reactor coolant system earlier during core

degradation and melting may re-volatilize and be

released into containment. Reference 17, after a review

of the source term uncertainty methodology used in

NUREG-1150 (Ref. 7), estimates the late in-vessel

release phase to have a duration of 10 hours. This value

has been selected for this report.

A summary of the release phases and the selected

duration times for PWRs and BWRs is shown for

reference purposes in Table 3.6.

Table 3.6

Release Phase Durations for PWRs and BWRs

Duration, Duration,

PWRs BWRs

Release Phase (Hours) (Hours)

Coolant Activity 10 to 30 seconds 30 seconds'

Gap Activity 0.5 0.5

Early In-Vessel 1.3 1.5

Ex-Vessel 2 3

Late In-Vessel 10 10

Without approval for leak-before-break. Coolant

activity phase duration is assumed to be 10 minutes

with leak-before-break approval.

3.4 Fission Product Composition and

Magnitude

In considering severe accidents in which the containment might fail, WASH-1400 (Ref. 5) examined the

spectrum of fission products and grouped 54 radionuclides into 7 major groups on the basis of similarity in

chemical behavior. The effort associated with the STCP

9 NUREG-1465
further analyzed these groupings and expanded the 7

fission product groups into 9 groups. These are shown

in Table 3.7.

Table 3.7 STCP Radionuclide Groups

Group Elements

1 Xe, Kr

2 I, Br

3 Cs, Rb

4 Te, Sb, Se

5 Sr

6 Ru, Rh, Pd, Mo, 'T

7 La, Zr, Nd, Eu, Nb, Pm, Pr, Sm, Y

8 Ce, Pu, Np

9 Ba

Similarly, low pressure sequences cause aerosols

generated within the RCS to be swept out rapidly

without significant retention within the RCS, thereby

resulting in higher release fractions from the core into

containment.

Table 3.8 Revised Radionuclide Groups

Group Title Elements in Group

1 Noble gases Xe, Kr

2 Halogens I, Br

3 Alkali Metals Cs, Rb

4 Tellurium group 'T, Sb, Se

5 Barium, strontium Ba, Sr

6 Noble Metals Ru, Rh, Pd, Mo, Tc, Co

7 Lanthanides La, Zr, Nd, Eu, Nb, Pm,

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)

reported in NUREG/CR-5747 (Ref. 17) indicate only

minor differences between Ba and Sr releases. Hence,

a revised grouping of radionuclides has been developed

that groups Ba and Sr together. The relative

importance to offsite health and economic

consequences of the radioactive elements in a nuclear

reactor core has been examined and documented in

NUREG/CR-4467 (Ref. 20). In addition to the

elements already included in Thble 3.7, Reference 20

found that other elements such as Curium could be

important for radiological consequences if released in

sufficiently large quantities. For this reason, group 7

has been revised to include Curium (Cm) and

Americium (Am), while group 6 has been revised to

include Cobalt (Co). The revised radionuclide groups

used in this report including revised titles and the

elements comprising each group are shown in Table 3.8.

Source term releases into the containment were

evaluated by reactor type, i.e., BWR or PWR, from the

sequences in NUREG-1150 and the supplemental

STCP calculations discussed in Section 3.1.

Releases into containment during the early in-vessel

phase, prior to reactor pressure vessel failure, are

markedly affected by retention in the RCS, which is a

function of the residence time in the RCS during core

degradation. High pressure in the RCS during core

degradation allows for longer residence time of

aerosols released from the core. This, in turn, permits

increased retention of aerosols within the RCS and

lower releases from the core into the containment.

The relative frequency of occurrence of high vs. low

pressure sequences were examined for both BWRs and

PWRs. The results of this survey are shown in

Thble 3.9, and they indicate that a significant fraction of

the sequences examined, in terms of frequency,

occurred at low pressure. In addition, advanced PWR

designs are increasingly incorporating safety-grade

depressurization systems, primarily to minimize the

likelihood of high pressure melt ejection (HPME) with

its associated high containment atmosphere heat loads

and large amounts of atmospheric aerosols.

For these reasons, the composition and magnitude of

the source term has been chosen to be representative

of conditions associated with low pressure in the RCS

at the time of reactor core degradation and pressure

vessel failure. Reference 17 provides estimates of the

mean core fractions released into containment, as

estimated by NUREG-1150 (Ref. 7), for accident

sequences occurring under low RCS pressure and high

zirconium oxidation conditions. These are shown in

Tables 3.10 and 3.11.

3.5 Chemical Form

The chemical form of iodine and its subsequent

behavior after entering containment from the reactor

coolant system have been documented in

NUREG/CR-5732, Iodine Chemical Forms in LWR

Severe Accidents (Ref. 18) and in ORNLITM-12202,

"Models of Iodine Behavior in Reactor Containments,"

(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)

Low Pressure at

High Pressure at Intermed. press. Vessel Breach

Boiling Water Reactors Vessel Breach at vessel breach (<200 psi) No vessel breach

LaSalleexternal events only 0.27 N/A 0.67 0.06

LaSalleinternal 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 or Radionuclides Into Containment for BW'Rs,

Low RCS Pressure, High Zirconium Oxidation

Nuclide Early In-Vessel Ex-Vessel Late In-vessel

N.G. 1.0 0 0

I 0.27 0.37 0.07

Cs 0.2 0.45 0.03

le 0.11 0.38 0.01

Sr 0.03 0.24 0

Ba 0.03 0.21 0

Ru 0.007 0.004 0

La 0.002 0.01 0

Ce 0.009 0.01 0

Table 3.11 Mean Values of Radionuclide Releases Into Containment for PWRs,

Low 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 0.3 0.39 0.06

,lb 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

the containment is at least 95% CsI with the remaining

5% as I plus HI, with not less than 1% of each as I and

HI. Once the iodine enters containment, however,

additional reactions are likely to occur. In an aqueous

environment, as expected for LWRs, iodine is expected

to dissolve in water pools or plate out on wet surfaces

in ionic form as I-. Subsequently, iodine behavior

within containment depends on the time and pH of the

water solutions. Because of the presence of other

dissolved fission products, radiolysis is expected to

occur and lower the pH of the water pools. Without any

pH control, the results indicate that large fractions of

the dissolved iodine will be converted to elemental

iodine and be released to the containment atmosphere.

However, if the pH is controlled and maintained at a

value of 7 or greater, very little (less than 1%) of the

dissolved iodine will be converted to elemental iodine.

Some considerations in achieving pH control are

discussed in NUREG/CR-5950, "Iodine Evolution and

pH Control," (Ref. 22).

Organic compounds of iodine, such as methyl iodide,

CH31, can also be produced over time largely as a

result of elemental iodine reactions with organic

materials. Organic iodide formation as a result of

reactor accidents has been surveyed in WASH-1233,

"Review of Organic Iodide Formation Under Accident

Conditions in Water-Cooled Reactors," (Ref. 23), and

more recently in NUREG/CR-4327, "Organic Iodide

Formation Following Nuclear Reactor Accidents,"

(Ref.24). From an analysis of a number of containment

experiments, WASH-1233 concluded that, considering

both non-radiolytic as well as radiolytic means, no more

than 3.2 percent of the airborne iodine would be

converted to organic iodides during the first two hours

following a fission product release. The value of 3.2

percent was noted as a conservative upper limit and was

judged to be considerably less, since it did not account,

among other things, for decreased radiolytic formation

of organic iodide due to iodine removal mechanisms

within containment. Reference 24 also included results

involving irradiated fuel elements, and concluded that

the organic iodide concentration within containment

would be about 1 percent of the iodine release

concentration over a wide range of iodine

concentrations.

A conversion of 4 percent of the elemental iodine to

organic has been implicitly assumed by the NRC staff in

Regulatory Guides 1.3 and 1.4, based upon an upper

bound evaluation of the results in WASH-1233.

However, in view of the results of Ref. 23 that a

conversion of 3.2 percent is unduly conservative, a

value of 3 percent is considered more realistic and will

be used in this report. Where the pH is controlled at

values of 7 or greater within the containment,

elemental iodine can be taken as comprising no more

than 5 percent of the total iodine released, and iodine

in organic form may be taken as comprising no greater

than 0.15 percent (3 percent of 5 percent) of the total

iodine released.

Organic iodide formation in BWRs versus PWRs is not

notably different. Reference 18 examined not only

iodine entering containment as CsI; but also considered

other reactions that might lead to volatile forms of

iodine within containment, such as reactions of CsOH

with surfaces and revaporization of CsI from RCS

surfaces. Reference 18 indicates (Thble 2.4) that for the

Peach Bottom TC2 sequence, the estimated percentage

of iodine as HI was 3.2 percent, not notably less than

the PWR sequences examined. While organic iodide is

formed largely from reactions of elemental iodine, Ref.

22 clearly notes that reactions with HI may be

important.

Although organic iodine is not readily removed by

containment sprays or filter systems, it is unduly

conservative to assume that organic iodine is not

removed at all from the containment atmosphere, once

generated, since such an assumption can result in an

overestimate of long-term doses to the thyroid.

References 23 and 24 discuss the radiolytic destruction

of organic iodide, and Standard Review Plan Section

(S.R.P.) 6.5.2 notes the above reference and indicates

that removal of organic iodide may be considered on a

case-by-case basis. A rational model for organic iodine

behavior within containment would consider both its

formation as well as destruction in a time-dependent

fashion. Development of such a model, however, is

beyond the scope of the present report.

Clearly, where the pH is not controlled to values of 7

or greater, significantly larger fractions of elemental

iodine, as well as organic iodine may be expected within

containment.

All other fission products, except for the noble gases

and iodine, discussed above, are expected to be in

particulate form.

3.6 Proposed Accident Source Terms

The proposed accident source terms, including their

timing as well as duration, are listed in Thbles 3.12 for

BWRs and 3.13 for PWRs. The information for these

tables was derived from the simplification of the

NUREG-1150 (Ref. 7) source terms documented in

NUREG/CR-5747 (Ref. 17). It should also be noted

that the rate of release of fission products into the

containment is assumed to be constant during the

duration time shown.

NUREG-1465 12
Table 3.12 BWR Releases Into Containment*

Gap 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

Tellurium group 0 0.05 0.25 0.005

Barium, Strontium 0 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 Table 3.8 for a listing of the elements in each group

*** Gap release is 3 percent if long-term fuel cooling is maintained.

Table 3.13 PWR Releases Into Containment

Gap Release*** Early In-Vessel Ex-Vessel Late In-Vessel

Duration (Hours) 0.5 1.3 2.0 10.0

Noble Gases* 0.05 0.95 0 0

Halogens 0.05 0.35 0.25 0.1

Alkali Metals 0.05 0.25 0.35 0.1

Tellurium group 0 0.05 0.25 0.005

Barium, Strontium 0 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 Table 3.8 for a listing of the elements in each group

* Gap release is 3 percent if long-term fuel cooling is maintained.

It is emphasized that the release fractions for the

source terms presented in this report are intended to

be representative or typical, rather than conservative or

bounding values, of those associated with a low

pressure core-melt accident, except for the initial

appearance of fission products from failed fuel, which

was chosen conservatively. The release fractions are not

intended to envelope all potential severe accident

sequences, nor to represent any single sequence.

Tibles 3.12 and 3.13 in this, the final report, were

modified from the tables in the draft report which were

taken from Table 3.9 and Table 3.10, for BWRs and

PWRs, respectively. The changes and the reasons for

these was as follows:

1. BWR in-vessel release fractions for the volatile

nuclides (I and Cs) increased slightly while

ex-vessel release fractions for the same nuclides

was reduced as a result of comments received and

additional MELCOR calculations available after

issuance of the draft report. The total I and Cs

released into containment over all phases of the

accident remained the same.

2. Release fractions for Te, Ba and Sr were reduced

somewhat, both for in-vessel as well as ex-vessel

releases, in response to comments.

13 NUREG-1465
3. Release fractions for the non-volatile nuclides,

particularly during the early in-vessel phase were

reduced significantly based on additional research

results (Ref. 25) since issuance of NUREG-1150

which indicate that releases of low volatile

nuclides, both in-vessel as well as ex-vessel, have

been overestimated. A re-examination in response

to comments received showed that the supposed

"means" of the uncertainty distribution were in

excess of other measures of the distribution, such

as the 75th percentile. In this case, the 75th

percentile was selected as an appropriate measure

of the release fraction. For additional discussion

on this topic, see Section 4.4.

4. Gap activity release fractions were reduced from 5

percent to 3 percent for accidents not involving

degraded or molten core conditions, and where

long-term fuel cooling is maintained. See

additional discussion below.

Based on WASH-1400 (Ref. 5), the inventory of fission

products residing in the gap between the fuel and the

cladding is no greater than 3 percent except for cesium,

which was estimated to be about 5 percent.

NUREG/CR-4881 (Ref.16) reported a comparison of

more recently available estimations and observations

indicating that releases of the dominant fission product

groups were generally below the values reported in

Reference 5. However, the magnitude of fission

products released during the gap release phase can

vary, depending upon the type of accident. Accidents

where fuel failures occur may be grouped as follows:

1. Accidents where long-term fuel cooling is

maintained despite fuel failure. Examples include

the design basis LOCA where ECCS functions,

and a postulated spent fuel handling accident. For

this category, fuel failure is taken to result in an

immediate release, based upon References 5 and

16, of 3 percent of the volatile fission products

(noble gases, iodine, and cesium) which are in the

gap between the fuel pellet and the cladding. No

subsequent appreciable release from the fuel

pellet occurs, since the fuel does not experience

prolonged high temperatures.

2. Accidents where long-term fuel cooling or core

geometry are not maintained. Examples include

degraded core or core-melt accidents, including

the postulated limiting design basis fission product

release into containment used to show compliance

with 10 CFR Part 100. For this category, the gap

release phase may overlap to some degree with

the early in-vessel release phase. The release

magnitude has been taken as an initial release of 3

percent of the volatiles (as for category 1), plus an

additional release of 2 percent over the duration

of the gap release phase.

3. Accidents where fuel failure results from reactivity

insertion accidents (RIA), such as the postulated

rod ejection (PWR) or rod drop (BWR) accidents.

The accidents examined in this report do not

contain information on reactivity induced

accidents to permit a quantitative discussion of

fission product releases from them. Hence, the

gap release magnitude presented in Tables 3.12

and 3.13 may not be applicable to fission product

releases resulting from reactivity insertion

accidents.

Recent information has indicated that high burnup fuel,

that is, fuel irradiated at levels in excess of about 40

GWD/MTU, may be more prone to failure during

design basis reactivity insertion accidents than

previously thought. Preliminary indications are that

high bumup fuel also may be in a highly fragmented or

powdered form, so that failure of the cladding could

result in a significant fraction of the fuel itself being

released. In contrast, the source term contained in this

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

of those fission product gases residing in the gap

between the fuel pellet and the cladding. Because of

this recent information regarding high burnup fuels, the

NRC staff cautions that, until further information

indicates otherwise, the source term in Tables 3.12 and

3.13 (particularly gap activity) may not be applicable for

fuel irradiated to high burnup levels (in excess of about

40 GWD/MTU).

With regard to the ex-vessel releases associated with

core-concrete interactions, according to Reference 17,

there were only slight differences in the fission

products released into containment between limestone

vs. basaltic concrete. Hence, the table shows the

releases only for a limestone concrete. Further, the

releases shown for the ex-vessel phase are assumed to

be for a dry reactor cavity having no water overlying any

core debris. Where water covers the core debris,

aerosol scrubbing will take place and reduce the

quantity of aerosols entering the containment

atmosphere. See Section 5.4 for further information.

3.7 Nonradioactive Aerosols

In addition to the fission product releases into

containment shown in Tables 3.12 and 3.13, quantities

of nonradioactive or relatively low activity aerosols will

also be released into containment. These aerosols arise

from core structural and control rod materials released

during the in-vessel phase and from concrete decomposition products during the ex-vessel phase. A detailed

NUREG-1465 14
analysis of the quantity of nonfission product aerosols

released into containment was not undertaken. Precise

estimates of the masses of non-radioactive aerosols

released into containment are difficult to determine.

Reference 26 evaluated one PWR sequence (Sequoyah)

and one BWR (Peach Bottom) sequence and calculated

in-vessel non-radioactive aerosol masses of 350 and 780

kilograms, respectively, for the PWR and BWR

sequences. The same reference calculated that

ex-vessel aerosol masses (assuming a dry cavity) would

be higher, 3800 and 5600 kilograms, respectively, for

the PWR and BWR sequences investigated. However,

these values, particularly for the ex-vessel release

phase, may be excessive. NUREG/CR-4624 (Ref. 27)

examined several sequences for both PWRs and BWRs

and calculated ex-vessel releases to containment of

about 1000 and 4000 kilograms, respectively, for PWRs

and BWRs. NUREG/CR-5942 (Ref.19), making use of

the MELCOR code, calculated significantly lower

releases during the ex-vessel phase of about 1000

kilograms for the Peach Bottom plant.

In view of the wide diversity of calculated results, the

NRC staff concludes that precise estimates of the

release of non-radioactive aerosols are not available at

this time. Because nonradioactive aerosol masses could

have an effect upon the operation of certain plant

equipment, such as filter loadings or sump performance, during and following an accident, however, the

NRC staff concludes that the release of non-radioactive

aerosols should be considered by the designer using

methods considered applicable for his design, and the

potential impact upon the plant evaluated.

4 MARGINS AND UNCERTAINTIES

This section discusses some of the more significant

conservatisms and margins in the proposed accident

source term given in Section 3. Briefly, the proposed

release fractions have been developed from a complete

core-melt accident, that is, assuming core melt with

reactor pressure vessel failure and with the assumption

of core-concrete interactions. The timing aspects were

selected to be typical of a low pressure core-melt

scenario, except that the onset of the release of gap

activity was based upon the earliest calculated time of

fuel rod failure under accident conditions. The

magnitude of the fission products released into

containment was intended to be representative and,

except for the low volatile nuclides, as discussed in

section 4.4, was estimated from the mean values for a

typical low-pressure core-melt scenario.

4.1 Accident Severity and lype

As noted earlier in Section 2.2, this report discusses

mean or average release fractions for all the release

phases associated with a complete core-melt accident,

including reactor pressure vessel failure. The accident

selected is one in which core melt occurs at low

pressure conditions. A low pressure core melt scenario

results in a relatively low level of fission product

retention within the reactor coolant system, and a

consequently high level of release of fission products

from the core into containment during the early

in-vessel release phase. Since the bulk of the fission

products entering containment do so during the early

in-vessel release phase, selection of a low pressure core

melt scenario provides a high estimate of the total

quantity of fission products released into containment,

as well as that during the early in-vessel release phase.

4.2 Onset of Fission Product Release

The onset, or earliest time of appearance of fission

products within containment, has been selected on the

basis of the earliest time to failure of a fuel rod, given a

design basis LOCA. This is estimated to be from about

13 to 25 seconds for plants that do not have leakbefore-break approval for their reactor coolant system

piping, and it is expected to vary depending on the

reactor as well as the fuel rod design. This value, while

representing some relaxation from the assumption of

instantaneous appearance, is nevertheless conservative.

As noted in Reference 15, these estimates are valid for

a double-ended rupture of the largest pipe, assume that

the fuel rod is being operated at the maximum peaking

factor permitted by the plant Technical Specifications

and at the highest burnup levels anticipated, and

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

to fuel rod failure by factors of two or more. Nevertheless, the use of conservative assumptions in estimating

fuel rod failure times is considered appropriate since

such failure times are likely to be used primarily in

consideration of the necessary closure time for certain

containment isolation valves. Since it is important that

closure of such valves be ensured before the release of

significant radioactivity to the environment, a conservative estimate of fuel failure time and consequent onset

of fission product appearance is deemed appropriate.

For plants with leak-before-break approval for their

reactor coolant system piping, a longer duration before

fuel clad failure is expected. However, other constraints

may become the limiting factor on containment

isolation valve closure time.

4.3 Release Phase Durations

The durations of the various release phases have been

selected primarily by examination of the values

15 NUREG-1465
available for the group of severe accident scenarios

considered in Section 3. The durations of the early

in-vessel and ex-vessel release phases differs for BWRs

versus PWRs and reflect the differing core heatup rates

as well as the differing amounts of zirconium available

to supply chemical energy after core-melt. While the

selected durations of the release phases are realistic,

some conservatisms should be noted. The duration of

the early in-vessel release phase for BWRs and PWRs

is short and does not represent a probabilistically

weighted average or mean value for the accident

sequences considered. This will introduce a given

quantity of fission products into containment in a

shorter time than might be expected for a typical

sequence.

Similarly, the duration of the ex-vessel release phase,

while considered realistic for the bulk of the fission

products being released, is short for releases of

tellurium and ruthenium since, as noted in Section 3.3,

release of these nuclides occurs over a longer time.

The selected release duration times have been chosen

primarily on the basis of simplicity, since an accurate

determination of the duration of the release phases

depends not only on the reactor type but also on the

applicable accident sequence, which varies for each

reactor design.

4.4 Composition and Magnitude of

Releases

The composition of the fission products was initially

based on the grouping developed with the STCP, but

has been modified as discussed in Section 3.4.

The magnitudes of the fission products released into

containment for the accident source term were selected

in the draft version of this report to be the mean

values, using NUREG-1150 methodology, for BWR

and PWR low-pressure scenarios involving high

estimates of zirconium oxidation. The uncertainty

distributions for the in-vessel release and total release

into containment are displayed graphically in Appendix A. Bounding estimates for the releases into

containment taken from Reference 17, using the STCP

methodology, are shown in Appendix B.

The release magnitudes for the low volatile fission

products were reduced significantly in the final report.

This reduction was based upon recent experimental

research results (Ref. 25) since completion of

NUREG-1150, as well as a re-examination of the

uncertainty distribution, in response to comments on

the draft report. Research on in-vessel phenomena

includes the in-pile Severe Fuel Damage (SFD)

experiments in the Power Burst Facility (PBF), further

examination of the Three Mile Island (TMI) accident,

and the SASCHA out-of-pile tests. Ex-vessel insights

derive primarily from large scale tests performed as

part of the internationally sponsored ACE Program.

Reference 25 notes that, based on the SFD experiments

as well as the TMI accident, in-vessel release fractions

for cerium, for example, were about 104, compared to

the value of 10-2 cited in the draft report. Based on

these results, the NRC staff concludes that the low

volatile release fractions cited in draft NUREG-1465

are too high.

The uncertainty distributions were also examined to

obtain additional insight. As can be seen from the

uncertainty distributions in Appendix A, the range of

release estimates for the volatile nuclides, such as the

noble gases, iodine, cesium, and to some extent

tellurium, spans about one order of magnitude. For this

group of nuclides, use of the mean value is a

reasonable estimate of the release fraction. In contrast,

the range for the low volatile nuclides, such as barium,

strontium, cerium and lanthanum, spans about 4 to 6

orders of magnitude. For the latter group of nuclides,

the mean value can be misleading, since it may be well

in excess of other measures of the distribution. This is

illustrated in TIable 4.1 which tabulates the mean,

median, and 75th percentile values for several low

volatile nuclides released during the early in-vessel

phase.

Table 4.1 Measures of Low Volatile In-Vessel Release

Fractions

Nuclide Mean Median 75th percentile

Sr 0.03 0.001 0.006

Ba 0.04 0.003 0.009

La 0.002 0.00003 0.0003

Ce 0.01 0.00006 0.0006

As can be seen from Thble 4.1, the mean value for this

group of nuclides is one to two orders of magnitude

greater than the median value, and is about 5 times

greater than the 75th percentile of the distribution. For

this group of nuclides, the mean is controlled by the

upper tail of the distribution, and the details of the

whole distribution may be more indicative of the

uncertainty than the "bottom line" results, such as a

mean value. Because of this, the final version of this

report has chosen not to use the mean value in

estimating releases for the non-volatile nuclides. While

the median value might be selected as an alternate, it

fails to provide an appreciation of the range of values

lying above it. Since this report is intended for

regulatory applications, the intent is to avoid

under-estimation of potential releases or offsite doses,

without undue conservatism. Hence, for the final

NUREG-1465 16
report, the 75th percentile value has been selected for

the low volatile nuclides on the basis that it bounds

most of the range of values, without undue influence by

the upper tail of the distribution.

Uncertainties, particularly in understanding and

modeling core melt progression phenomena, can affect

the duration of the early in-vessel release phase,

including the timing of reactor pressure vessel failure.

An increase in duration of the early in-vessel phase can

lead to increased releases of volatile fission products

during the early in-vessel phase and a concomitant

reduction during the ex-vessel phase. An increase in

duration of the early in-vessel phase, however, also

provides additional time for fission product removal

within containment by natural processes or fission

product cleanup systems.

Upper bound estimates, tabulated in Appendix B,

indicate that virtually all the iodine and cesium could

enter the containment. Similarly, for tellurium, upper

bound estimates indicate that as much as about

two-thirds of the core inventory of tellurium could be

released into containment. Hence, for this important

group of radionuclides (iodine, cesium, and tellurium),

the upper bound estimates of total release into

containment are approximately 1.5 times the mean

value estimates.

For the lower volatility radionuclides such as barium

and strontium, upper bound estimates range from

about 50 to 70% of the core inventory released into

containment. Almost all of this is estimated to be

released as a result of core-concrete interactions. In

contrast, mean value estimates range from 15 to 25%.

Hence, in this case, the upper bound estimates are

about two to three times the mean values.

Finally, for the refractory nuclides such as lanthanum

and cerium, the upper bound estimates indicate that

about 5% of the inventory of these nuclides could

appear within containment, whereas the mean value

estimate indicates only about 1% released.

PRAs have indicated that, considering the magnitudes

of the radioactive species estimated to be released to

the environment for severe reactor accidents, the

radionuclides having the greatest impact on risk are

typically the volatile nuclides such as iodine and

cesium, with tellurium to a somewhat lesser degree.

The uncertainty distributions for this group of

radionuclides is also the smallest, as shown in the

graphical tabulations of Appendix A. Hence, our ability

to predict the behavior and releases for this group of

nuclides is significantly better than for other fission

product groupings.

Mean value estimates selected for the in-containment

accident source term provide reasonable estimates for

the important nuclides consisting of iodine, cesium, and

tellurium. These estimates show a relatively low degree

of uncertainty and are unlikely to be exceeded by more

than 50%. Uncertainty increases in estimating releases

for the remaining nuclides.

4.5 Iodine Chemical Form

The chemical form of iodine entering containment was

investigated in Reference 18. On the basis of this work,

the NRC staff concludes that iodine entering

containment from the reactor coolant system is

composed of at least 95% cesium iodide (CsI), with no

more than 5% 1 plus HI. Once within containment,

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

elemental iodine will potentially be an important

mechanism. If the pH is controlled to a level of 7 or

greater, such conversion to elemental iodine will be

minimal. If the pH is not controlled, however, a

relatively large fraction (greater for PWRs than BWRs)

of the iodine dissolved in containment pools in ionic

form will be converted to elemental iodine.

5 IN-CONTAINMENT REMOVAL

MECHANISMS

Since radioactive fission products within containment

are in the form of gases and finely divided airborne

particulates (aerosols), the principal mechanism by

which fission products find their way from the reactor

to the environment with an intact containment is via

leakage from the containment atmosphere. The specific

fission product inventory present in the containment

atmosphere at any time depends on two factors: (1) the

source, i.e., the rate at which fission products are being

introduced into the containment atmosphere, and

(2) the sink, the rate at which they are being removed.

Aspects of the release and transport of fission products

from the core into the containment atmosphere were

presented in Section 3.

Mechanisms that remove fission products from the

atmosphere with consequent mitigation of the

in-containment source term fall into two classes:

(1) engineered safety features (ESFs) and (2) natural

processes. ESFs to remove or reduce fission products

within the containment are presently required

(Criterion 41 in Appendix A of 10 CFR Part 50) and

include such systems as containment atmosphere

sprays, BWR suppression pools, and filtration systems

utilizing both particulate filters and charcoal adsorption

beds for the removal of iodine, particularly in elemental form. Natural removal includes such processes as

17 NUREG-1465
aerosol deposition and the sorption of vapors on

equipment and structural surfaces.

The draft version of this report contained a discussion

of some of the more important fission product removal

mechanisms, including some quantitative results. These

numerical results were intended to be illustrative of the

phenomena involved and were not intended to be

applied rigorously, however. It was recognized that the

data and illustrations used in the draft might not be

applicable to all situations.

In recognition of this, the NRC staff undertook to

examine, with contractor assistance, improved

understanding of fission product removal mechanisms.

At this time, this effort is still underway. Rather than

provide numerical values that may be inapplicable, this

report will provide references, where available, so that

the reader may utilize improved methodologies to

obtain results that apply to the situation at hand.

5.1 Containment Sprays

Containment sprays, covered in Standard Review Plan

(SRP) Section 6.5.2 (Ref. 28), are used in many PWR

designs to provide post-accident containment cooling as

well as to remove released radioactive aerosols. Sprays

are effective in reducing the airborne concentration of

elemental and particulate iodines as well as other

particulates, such as cesium, but are not effective in

removing noble gases or organic forms of iodine. The

reduction in airborne radioactivity within containment

by a spray system as a function of time is expressed as

an exponential reduction process, where the spray

removal coefficient, lambda, is taken to be constant

over a large part of the regime. Typical PWR

containment spray systems are capable of rapidly

reducing the concentration of airborne activity (by

about 2 orders of magnitude within about 30 minutes,

where both spray trains are operable). Once the bulk of

the activity has been removed, however, the spray

becomes significantly less effective in reducing the

remaining fission products. This is usually accounted

for by either employing a spray cut-off, wherein the

spray removal becomes zero after some reduction has

been achieved, or changing to a much smaller value of

lambda to reflect the decreased removal effectiveness

of the spray when airborne concentrations are low.

SRP Section 6.5.2 (Ref. 28) provides expressions for

calculating spray lambdas, depending on plant

parameters as well as the type of species removed. In

addition, SRP 6.5.2 currently suggests that the

containment sump solution be maintained at values at

or above pH levels of 7, commencing with spray

recirculation, to minimize revolatilization of iodine in

the sump water. Current guidance states that

containment spray systems be initiated automatically,

because of the instantaneous appearance of the source

term within containment, and that the spray duration

not be less than 2 hours. In contrast, the revised source

term information given in Section 3 suggests that spray

system actuation might be somewhat delayed for

radiological purposes, but that the spray system

duration should be for a longer period of about 10 or

more hours. Because sprays are effective in rapidly

removing particulates from the containment

atmosphere, intermittent operation over a prolonged

period may also provide satisfactory mitigation.

The spray removal coefficient for particulates appears

particularly important in view of the information

presented in Section 3, which indicates that most fission

products are expected to be in particulate form. The

spray removal coefficient (X) is derived from the

following equation from Standard Review Plan

Section 6.5.2

x =3hFE

2VD

h = Fall height of spray drops V = Containment building volume

F = Spray flow

E/D = the ratio of a dimensionless collection

efficiency E to the average spray drop Diameter D.

EJD is conservatively assumed to be equal to

10/meter for spray drops 1 mm in diameter changing

to 1/meter when the aerosol mass has been

depleted by a factor of 50.

Using values typical for PWRs, the formulation given in

SRP 6.5.2 estimates particulate removal rates to be on

the order of 5 per hour. Nourbakhsh (Ref. 29) examined the effectiveness of containment sprays, as

evaluated in NUREG-1150 (Ref. 7), in decontaminating both in-vessel and ex-vessel releases. Powers and

Burson (Ref. 30) have developed a more realistic, yet

simplified, model with regard to evaluating the

effectiveness of aerosol removal by containment sprays

5.2 BWR Suppression Pools

BWRs use pressure suppression pools to condense

steam resulting from a loss-of-coolant accident. Prior to

the release to the reactor building, these pools also

scrub radioactive fission products that accompany the

steam. Regulatory Guide 13 (Ref. 2) suggests not

allowing credit for fission product scrubbing by BWR

suppression pools, but SRP Section 6.5.5 (Ref. 31) was

revised to suggest allowing such credit. The pool water

will retain soluble, gaseous, and solid fission products

such as iodines and cesium but provide no attenuation

of the noble gases released from the core. The Reactor

Safety Study (WASH-1400, Ref. 5) assumed a

decontamination factor (DF) of 100 for subcooled

NUREG-1465 18
suppression pools and 1.0 for steam saturated pools.

Since 1975 when WASH-1400 was published, several

detailed models have been developed for the removal

of radioactive aerosols during steam flow through

suppression pools.

Calculations for a BWR with a Mark I containment

(Ref. 27) used in NUREG-1150 (Ref. 7) indicate that

DFs ranged from 1.2 to about 4000 with a median value

of about 80. The suppression pool has been shown to be

effective in scrubbing some of the most important

radionuclides such as iodine, cesium, and tellurium, as

these are released in the early in-vessel phase. The

NRC staff is also presently reviewing fission product

scrubbing by suppression pools to develop simplified

models.

If not bypassed, the suppression pool will also be

effective in scrubbing ex-vessel releases. Suppression

pool bypass is an important aspect that places an upper

limit on the overall performance of the suppression

pool in scrubbing fission products. For example, if as

little as 1% of the fission products bypass the

suppression pool, the effective DF, taking bypass into

account, will be less than 100, regardless of the pool's

ability to scrub fission products.

Although decontamination factors for the suppression

pool are significant, the potential for iodine

re-evolution can be important. Re-evolution of iodine

was judged to be important in accident sequences

where the containment had failed and the suppression

pool was boiling. There is presently no requirement for

pH control in BWR suppression pools. Hence, it is

possible that suppression pools would scrub substantial

amounts of iodine in the early phases of an accident,

only to re-evolve it later as elemental iodine. It may

well be that additional materials likely to be in the

suppression pool as a result of a severe accident, such

as cesium borate or cesium hydroxide and core-concrete

decomposition products, would counteract any

reduction in pH from radiolysis and would ensure that

the pH level was sufficiently high to preclude

re-evolution of elemental iodine. Therefore, if credit is

to be given for long-term retention of iodine in the

suppression pool, maintenance of the pH at or above a

level of 7 must be demonstrated. It is important to

note, however, that this is not a matter of concern for

present plants since all BWRs employ safety-related

filtration systems (see Section 5.3) designed to cope

with large quantities of elemental iodine. Hence, even

if the suppression pool were to re-evolve significant

amounts of elemental iodine, it would be retained by

the existing downstream filtration system.

5.3 Filtration Systems

ESF filtration systems are discussed in Regulatory

Guide 1.52 (Ref. 32) and are used to reduce the

radioactive aerosols and iodine released during

postulated accident conditions.

A typical ESF filtration system consists of redundant

trains that each have demisters to remove steam and

water droplets from the air entering the filter bank,

heaters to reduce the relative humidity of the air, high

efficiency particulate air (HEPA) filters to remove

particulates, charcoal adsorbers to remove iodine in

elemental and organic form, followed finally by

additional HEPA filters to remove any charcoal fines

released.

Charcoal adsorber beds can be designed, as indicated in

Regulatory Guide 1.52, to remove from 90 to 99% of

the elemental iodine and from 30 to 99% of the organic

iodide, depending upon the specific filter train design.

Revised insights on accident source terms, given in

Section 3, may have several implications for ESF

filtration systems. Present ESF filtration systems are

not sized to handle the mass loadings of nonradioactive aerosols that might be released as a result

of the ex-vessel release phase, which could produce

releases of significant quantities of nonradioactive as

well as radioactive aerosols. However, if ESF filtration

systems are employed in conjunction with BWR

suppression pools or if significant quantities of water

are overlaying molten core debris (see Section 5.4),

large quantities of nonradioactive (as well as

radioactive) aerosols will be scrubbed and retained by

these water sources, thereby reducing the aerosol mass

loads upon the filter system.

A second implication of revised source term insights for

ESF filtration systems is the impact of revised

understanding of the chemical form of iodine within

containment. Present ESF filtration systems presume

that the chemical form of iodine is primarily elemental

iodine, and these systems include charcoal adsorber

beds to trap and retain elemental iodine. Assuming that

pH control is maintained within the containment, a key

question is whether charcoal beds are necessary. Two

questions appear to have a bearing on this issue and

must be addressed, even assuming pH control. These

are (1) to what degree will Csl retained on particulate

filters decompose to evolve elemental iodine? and (2)

what effect would hydrogen bums have on the chemical

form of the iodine within containment? Based on

preliminary information, Csl retained on particulate

filters as an aerosol appears to be chemically stable

provided that it is not exposed to moisture. Exposure to

moisture, however, would lead to CsI decomposition

and production of iodine in ionic form (1), which in

turn would lead to re-evolution of elemental iodine.

Although ESF filtration systems are equipped with

demisters and heaters to remove significant moisture

before it reaches the charcoal adsorber bed, an

19 NUREG-1465
additional concern is that the demisters themselves may

trap some CsI aerosol.

In conclusion, present ESF filtration systems, while

optimized to remove iodine, particularly in elemental

form, have HEPA filters that are effective in the -

removal of particulates as well. Although such filtration

systems are not designed to handle the large mass

loadings expected as a result of ex-vessel releases, when

they are used in conjunction with large water sources

such as BWR suppression pools or significant water

depths overlaying core debris, the water sources will

reduce the aerosol mass loading on the filter system

significantly, making such filter systems effective in

mitigation of a large spectrum of accident sequences.

5.4 Water Overlying Core Debris

Experimental measurements (Ref. 33) have shown that

significant depths of water overlying any molten core

debris after reactor pressure vessel failure will scrub

and retain particulate fission products. The question of

coolability of the molten debris as a result of water

overlying it is still under investigation. A major factor

that may affect the degree of scrubbing is whether the

water layer in contact with the molten debris is boiling

or not.

Results from Ref. 33 indicate that both subcooled as

well as boiling water layers having a depth of about

3 meters had measured DFs of about 10. A recent study

(Ref. 34) performed for the NRC has provided a

simplified model to determine the degree of aerosol

scrubbing by a water pool overlying core debris

interacting with concrete.

5.5 Aerosol Deposition

Since the principal pathway for transport of fission

products is via airborne particulates, i.e., aerosols, this

subject is discussed in some detail. Aerosols are usually

thought of as solid particulates, but in general, the term

also includes finely divided liquid droplets such as

water, i.e., fog. The two major sources of aerosols are

condensation and entrainment. Condensation aerosols

form when a vapor originating from some hightemperature source moves into a cooler region where

the vapor falls below its saturation temperature and

nucleation begins. Entrainment aerosols form when gas

bubbles break through a liquid surface and drag

droplets of the liquid phase into the wake of the bubble

as it leaves the surface. In general, condensation

particles are smaller in size (submicron to a few

microns), while entrainment particles are usually larger

(1.0-100 microns). Once airborne, both types of

aerosols behave in a similar manner with respect to

both natural and engineered removal processes.

There are four natural processes that remove aerosols

from the containment atmosphere over a period of

time: (1) gravitational settling, (2) diffusiophoresis,

(3) thermophoresis, and (4) particle diffusion. (Particle

diffusion is less important than the first three processes

and will not be discussed further.) All particles fall

naturally under the force of gravity and collect on any

available surface that terminates the fall, e.g., the floor

or upper surfaces of equipment. Both diffusiophoresis

and thermophoresis cause the deposition of aerosol

particles on all surfaces regardless of their orientation,

i.e., walls and ceiling as well as the floor.

Diffusiophoresis is the process by which water vapor in

the atmosphere 'drags' aerosol particles with it as it

migrates (diffuses) toward a relatively cold surface on

which condensation is taking place. Thermophoresis

also causes aerosol particles to move toward and

deposit on colder surfaces but not as a result of mass

motion. Rather, the decreasing average velocity of the

surrounding gas molecules tends to drive the particle

down the temperature gradient until it traverses the

interface layer and comes into contact with the surface

where it sticks.

Aerosol agglomeration is another natural phenomenon

that has an influence on the rates at which the removal

processes described above will proceed. Agglomeration

results from the random inelastic collisions of particles

with each other. The process brings about a gradual

increase in average particle size resulting in more rapid

gravitational settling. Three phenomena contribute to

particle growth by agglomeration: (1) Brownian motion,

(2) gravitational fall, and (3) turbulence. Brownian

agglomeration is caused by particle collisions resulting

from random 'buffeting' by high-energy gas molecules.

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

particles on their way down. Finally, rapid variations in

gas velocity and flow direction in the atmosphere, Le.,

turbulence, tend to increase the rate at which particle

collisions occur and thus increase the average particle

size. It is to be expected that, as agglomeration

advances, the size of the particle will increase, and its

shape can be expected to change as well. These latter

factors have a strong influence on the removal

processes.

The agglomeration and aerosol removal processes all

depend critically upon the thermodynamic state and

thermal-hydraulic conditions of the containment

atmosphere. For example, the condensation onto and

evaporation of water from the aerosol particles

themselves have strong effects on all of the

agglomeration and removal processes. Water condensed

on aerosol particles increases their mass and makes

them more spherical; both of these effects tend to

increase the rate of gravitational settling. Some

NUREG-1465 20
aerosols, such as CsI and CsOH, are hygroscopic and

absorb water vapor even when the containment

atmosphere is below saturation. As with condensation,

hygroscopicity also increases the rate of deposition.

Because of its importance to fields such as weather and

atmosphere pollution, the behavior of aerosols has

been under study for many decades. A number of

computer codes have been developed to specifically

consider aerosol behavior as it relates to nuclear

accident conditions. The most complete mechanistic

treatment of aerosol behavior in the reactor

containment is found in CONTAIN, a computer code

developed at Sandia National Laboratories under NRC

sponsorship for the analysis of containment behavior

under severe accident conditions. The aerosol models

in the NAUA code are very similar to those used in

CONTAIN; NAUA was developed at the

Kernforschungszentrum, Karlsrhue, F.R.G., and was

used for aerosol treatment in the NRC STCP. There

are a number of other well-known aerosol behavior

computer codes, but these two are the most widely used

and accepted throughout the international nuclear

safety community.

The rate at which gravitational settling occurs depends

upon the degree of agglomeration at any particular

time (i.e., the average particle size) as well as the total

particle density m (mass per unit volume). Thus, as in

most cases where the decrement of a variable is

proportional to the variable itself, one can expect an

exponential behavior. The gravitational settling process

is quite complex and depends upon a large number of

physical quantities, e.g., collision shape factor, particle

settling shape factor, gas viscosity, effective settling

height, density correction factor, normalized Brownian

collision coefficient, gravitational acceleration, and

particle material density. The only variable in this list

that is independent of the plant, the accident scenario,

and the atmospheric thermal-hydraulic conditions is the

constant of gravitation. It follows that no single DF can

be ascribed to cover the entire range of plant designs,

accident scenarios, and source materials. An effort is

under way to establish a set of simplified algorithms

that can be used to provide a set of specific ranges of

atmosphere conditions. This effort is still underway at

this time.

6. REFERENCES

1. U.S. Nuclear Regulatory Commission; "Reactor

Site Criteria," Title 10, Code of Federal

Regulations (CFR), Part 100.

2. U.S. Nuclear Regulatory Commission;

"Assumptions Used for Evaluating the Potential

Radiological Consequences of a Loss of Coolant

Accident for Boiling Water Reactors," Regulatory

Guide 1.3, Revision 2, June 1974.

3. U.S. Nuclear Regulatory Commission;

"Assumptions Used for Evaluating the Potential

Radiological Consequences of a Loss of Coolant

Accident for Pressurized Water Reactors,"

Regulatory Guide 1.4, Revision 2, June 1974.

4. JJ. DiNunno et al., "Calculation of Distance

Factors for Power and Test Reactor Sites,"

Technical Information Document (11D}-14844,

U.S. Atomic Energy Commission, 1962.

5. U.S. Nuclear Regulatory Commission; "Reactor

Safety Study An Assessment of Accident Risks in

U.S. Commercial Nuclear Power Plants,"

WASH-1400 (NUREG-75/014), December 1975.

6. J. A. Gieseke et al., "Source Term Code Package:

A User's Guide," NUREG/CR-4587 (BMI-2138),

prepared for NRC by Battelle Memorial Institute,

July 1986.

7. U.S. Nuclear Regulatory Commission; "Severe

Accident Risks: An Assessment for Five U.S.

Nuclear Power Plants," NUREG-1150, December

1990.

8. M.R. Kuhlman, DJ. Lehmicke, and R.O. Meyer,

"CORSOR User's Manual," NUREG/CR-4173

(BMI-2122), prepared for NRC by Battelle

Memorial Laboratory, March 1985.

9. H. Jordan, and M.R. Kuhlman, "TRAP-MELT2

User's Manual," NUREG/CR-4205 (BMI-2124),

prepared for NRC by Battelle Memorial

Laboratory, May 1985.

10. D.A. Powers, J.E. Brockmann, and A.W. Shiver,

"VANESNA A Mechanistic Model of Radionuclide

Release and Aerosol Generation During Core

Debris Interactions with Concrete,"

NUREG/CR-4308 (SAND 85-1370), prepared for

NRC by Sandia National Laboratories, July 1986.

11. P.C. Owczarski, A.K. Postma, and R.I. Schreck,

'Technical Bases and User's Manual for the

Prototype of SPARC-A Suppression Pool Aerosol

Removal Code," NUREG/CR-3317 (PNL-4742),

prepared for NRC by Battelle Pacific Northwest

Laboratories, May 1985.

12. W.K. Winegardner, A.K. Postma, and M.W.

Jankowski, "Studies of Fission Product Scrubbing

within Ice Compartments," NUREG/CR-3248

(PNL-4691), prepared for NRC by Battelle Pacific

Northwest Laboratories, May 1983.

21 NUREG-1465
13. H. Bunz, M. Kayro, and W. Schock, "NAUA-Mod

4: A Code for Calculating Aerosol Behavior in

LWR Core Melt Accidents," KfK-3554,

Kernforschungszentrum Karlsruhe Germany,

1983.

14. R.M. Summers, et al., "MELCOR 1.8.0: A

Computer Code for Nuclear Reactor Severe

Accident Source Term and Risk Assessment

Analysis," NUREG/CR-5531 (SAND 90-0364),

prepared for NRC by Sandia National

Laboratories, January 1991.

15. K.R. Jones, et al, '"Tming Analysis of PWR Fuel

Pin Failures," NUREG/CR-5787 (EGG-2657),

prepared for NRC by Idaho National Engineering

Laboratory, September 1992.

16. H.P. Nourbakhsh, M. Khatib-Rahbar, and R.E.

Davis, "Fission Product Release Characteristics

into Containment Under Design Basis and Severe

Accident Conditions," NUREG/CR-4881

(BNL-NUREG-52059), prepared for NRC by

Brookhaven National Laboratory, March 1988.

17. H.P. Nourbakhsh,: "Estimates of Radionuclide

Release Characteristics into Containment Under

Severe Accidents," NUREG/CR-5747

(BNL-NUREG-52289), prepared for NRC by

Brookhaven National Laboratory, November 1993.

18. E.C. Beahm, C.F. Weber, and T.S. Kress, "Iodine

Chemical Forms in LWR Severe Accidents",

NUREG/CR-5732 (ORNLTM-11861), prepared

for NRC by Oak Ridge National Laboratory, April

1992.

19. JJ. Carbajo, "Severe Accident Source Term

Characteristics for Selected Peach Bottom

Sequences Predicted by the MELCOR Code,"

NUREG/CR-5942 (ORNLnTM-12229), prepared

for NRC by Oak Ridge National Laboratory,

September 1993.

20. DJ. Alpert, D.I. Chanin, and LT. Ritchie,

"Relative Importance of Individual Elements to

Reactor Accident Consequences Assuming Equal

Release Fractions."NUREG/CR-4467, prepared

for NRC by Sandia National Laboratories, 1986.

21. C.F. Weber, E.C. Beahm and T.S. Kress, "Models

of Iodine Behavior in Reactor Containments,"

ORNrITM-12202, Oak Ridge National

Laboratory, October 1992.

22. E.C. Beahm, R.A. Lorenz, and C.F. Weber,

"Iodine Evolution and pH Control,"

NUREG/CR-5950, (ORNLErM-12242), prepared

for NRC by Oak Ridge National Laboratory,

December 1992.

23. A.K. Postma, and R.W. Zavadowski, "Review of

Organic Iodide Formation Under Accident

. Conditions in Water-Cooled Reactors,"

WASH-1233, U.S. Atomic Energy Commission,

October 1972.

24. E.C. Beahm, W.E. Shockley, and O.L. Culberson,

"Organic Iodide Formation Following Nuclear

Reactor Accidents," NUREG/CR-4327,

(ORNLITM-9627), prepared for NRC by Oak

Ridge National Laboratory, December 1985.

25. D. J. Osetek, "Low Volatile Fission Product

Releases During Severe Reactor Accidents,"

DOE/ID-13177-2, prepared for U.S. Department

of Energy by Los Alamos Technical Associates,

October 1992.

26. M. Silberberg et al., "Reassessment of the

Technical Bases for Estimating Source Terms,"

NUREG-0956, July 1986.

27. R.S. Denning, et al., "Radionuclide Release

Calculations for Selected Severe Accident

Scenarios: BWR Mark I Design,"

NUREG/CR-4624, Vol. 1, prepared for NRC by

Battelle Memorial Institute, July 1986.

28. U.S. Nuclear Regulatory Commission:

"Containment Spray as a Fission Product Cleanup

System," Standard Review Plan, Section 6.5.2,

Revision 2, NUREG-0800, December 1988.

29. H.P. Nourbakhsh,: "In-Containment Removal

Mechanisms," Presentation to NRC staff January

3, 1992, Brookhaven National Laboratory, January

1992.

30. D.A. Powers and S.B. Burson, "A Simplified

Model of Aerosol Removal by Containment

Sprays," NUREG/CR-5966, (SAND92-2689),

prepared for NRC by Sandia National

Laboratories, June 1993.

31. U.S. Nuclear Regulatory Commission: "Pressure

Suppression Pool as a Fission Product Cleanup

System," Standard Review Plan, Section 6.5.5,

NUREG-0800, December 1988.

32. U.S. Nuclear Regulatory Commission: "Design,

Testing, and Maintenance Criteria for Postaccident

Engineered-Safety-Feature Atmosphere Cleanup

System Air Filtration and Adsorption Units of

Light-Water-Cooled Nuclear Power Plants,"

Regulatory Guide 1.52, Revision 2, March 1978.

NUREG-1465 22
33. J. Hakii et al., "Experimental Study on Aerosol

Removal Efficiency for Pool Scrubbing Under

High Temperature Steam Atmosphere,"

Proceedings of the 21st DOE/NRC Nuclear Air

Cleaning Conference, August 1990.

34. D.A. Powers and J.L Sprung, "A Simplified Model

of Aerosol Scrubbing by a Water Pool Overlying

Core Debris Interacting With Concrete,"

NUREG/CR-5901, (SAND92-1422), prepared for

NRC by Sandia National Laboratories, November

1993.

23 NUREG-1465
APPENDIX A

UNCERTAINTY DISTRIBUTIONS

NUREG-1465 24
a)

U)

a)

co

r.

0

to

a)V)

co

.2

P.0

i:

- r1

U =

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

Uncertainty Distributions for Total Rdeases Into Containment PWR,.Low RCS Pressure, Limestone Concrete, Dry Cavity, Two Openings After VB, FPART = 1.

25 NUREG-1465
en

0)

-Z

0

0)

-j

C.)

to

lk:

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

Uncertainty DistributIons for Total Releases Into Containment PWR. Low ICS Pressure, Basaltic

Concrete, Dr Cavity, Two Openings After YB, FPART = 1.

NUREG-1465 26
i0

inU 10-3 Clo -51 f

.2 10 sth

10~

107

l-B

10

10 -

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

10 -

-Lc .. r{s1 Th I

10

-Jh

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

Uncertainty Distributions fbor Total Releass late Contalatnent BWER, Low Pressure Fast Statlon

Blackout, LUmestone Concrete, Dry Pedestal, bou DrywvH Temperature, FPART = 1.

_0-27 NUREG-1465
APPENDIX B

STCP BOUNDING VALUE RELEASES

NUREG-1465 28
- Updated Bounding Value of Radionuclide Releases Into the Containment

Under Severe Accident Conditions for PWRs

n NVU

Hiah RCS Low RCS

Pressure Pressure

ST e

PIeh RCS

Pressurn

urme

Cont

NG 1.0 1.0 0. a

1 0.30 0.75 0.10 0.

Cs 0.30 0.75 .0.10 0.

To 0.20 0.50 0.05 0.

Sr-Ba 0.003 0.01 0.01 0.

Ru 0.003 0.01 0.05 o.C

La-Co 5 x105 1.5 x 10'4 0.01 0.

Release 40 minutes

Duration

(' All entries are fractions of Initial core Inventory.

(b) Assuming 100% of the core participate In CCI.

(') Except for To and Ru where the duration Is extended to five hours.

STr (b)

stone I

creto e

15

15

40

40

005

05

2 hourste'

3asaltic

oncrete

0.

0.15

0.15

0.30

0.15

0.005

0.05

STEV

Hlah RCS Low RCS

Pressure Pressure

0. 0.

0.05 0.02

0.02 0.02

0.02 0.01

10 hours -

C

m

0dI.-
z

IUpdated Bounding Value of Radionuclide Releases Into the Containment

Under Severe Accident Conditions for BWRs

MG

Cs

To

I r8-

Ru

La-Ca

Release

Duration

eTv (a

Hbh PCS Low RCS

Pressure Pressure")

1. 1.

0.50 0.75

0.50 0.75

0.10 0.15

0.003 0.01

0.003 0.01

5x10' 1.5x104

1.5 hours

Hith RCS

Pressure

0.

0.10

0.10

0.05

0.01

0.05

* 0.01

aTMXV!2

Llmestone Basaltic

Concrete Concrete

0. 0.

015 0.15

0.15 0.15

0.50 0.30

0.70 0.30

0.005 0.005

0.10 0.10

3 hours(d)

High RCS Low RCS

Pressure Pressure (b

0. 0.

0.10 0.02

0.05 0.01

0.02 0.02

10 hours

"' All entries are fractions of Initial core Inventory.

I' High pressure ATWS wre also considered In this category.

"' Assuming 100% of the core participate In CCI.

I' Except for To and Ru where the duration Is extended to six hours.
NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION . REPORT NUMBER 12 a9i (Al,~.d by NA C. Add Vo.I. $.g. Rjw.

32C013202. BIBLIOGRAPHIC DATA SHEET 'a umkfe I 

(SM inS fCtomS on the t.rain

2. TITLE AND SUBTITLE NUREG- 1465

Accident Source Terms for Lischt-Water Nuclear Power Plants

3. OATE REPORT PUBLISHED

MONTH Y YEAR

February 1995

4. FIN OR GRANT NUMBER

S. AUTHOR(S) 6. TYPE OF REPORT

L. Soffer, S. B. Burson, C. M. Ferrell, R. Y. Lee, J. N. Ridgelv 7.PERIOD COVEREO rneNjr"i

U.r s~Antlt r%,ul-Tnm, , N% %/fl inARFCCi'&fl 1%f no nif.Y_ At~h as'hf ax~ I, 'VM. .W U E. U'MPAl'. . MU ( qo yL m O. mbfg ,ujt nrw p v

-E d maia 4ad&oj

Division of Systems Technology

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington. DC 20555 -0001

9. SPONSORING ORGANIZATION -NAME AND ADDR ESS (If NRC. rp -Sw'v u egonrcMt, *t eofdoe NRC Ories on. Oftivo potion,. U.S. SIKOCMAeotjrorr Commni'o.,

an -wln odk

Same as above

10. SUPPLEMENTARY NOTES

11. ABSTRACT (Ie_ or ir

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 fission 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 evaluate other important plant performance requirements.

During the past 30 years substantial additional information on fission product releases has been developed

based on significant 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 reactors (LWRs). Current LWR licensees may voluntarily propose

applications based upon it.

12. KEY WORbSIOES/MPTORS (fL fet uIil.w AM"" 13. AVAILAE)ITY STATEMENT Unlimited

14. SECURIITY CLASSIFICATION

ITM, PWj

Severe Accident Source Term Unclassified

Core Meltdown 1Th rARpen

Design Basis Accident Unclassified

TID-14844 Replacement 15. NUMBER OF PAGES

Core Fission Product Releases

IS. PRICE

NPC FORM 33 m(2"
Federal Recycling Program